ORNL-1845: 1955-09 (19.9 MB PDF)

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ORNL-1845(Del0)
REACTORS -POWER
OPERATION OF THE AIRCRAFT REACTOR EXPERIMENT
MARTIN MARIETTA ENERGY SYSTEMS LIBRARIES
BY
W. B. Cottrell
H. E. Hungerford
J. K. Leslie
J. L. Meew
! 3 4456 0349822 1
September 6, 1955
Oak Ridge National Laboratory
Oak Ridge, Tennessee
UNITED STATES ATOMIC ENERGY COMMISSION
Technical Information Service
Date Declassified: February 1 2 , 1959.
Consolidation of this material into compact form to permit
economical, direct reproduction has resulted in multiple folios for some pages, e.g., 10-12, 27-29, etc.
E
LEGAL N O T I C E
This report was prepared a s an account of Government sponsored work. Neither the United
States, nor the Commission, nor any person acting on behalf of the Commission:
A. Makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use
of any information, apparatus, method, or process disclosed in this report may not infringe
privately owned rights; or
B. Assumes any liabilities with respect to the use of, or for damages resulting from the
use of any information, apparatus, method, or process disclosed in this report.
As used in the above, “person acting on behalf of the Commission” includes any employee or contractor of the Commission, or employee of such contractor, to the extent that
such employee or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides access to, any information pursuant to his employment or contract
with the Commission, or his employment with such contractor.
1.
This report has been reproduced directly from the best
available copy.
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Printed in USA. Price $3.50. Available from the Office of
Technical Services, Department of Commerce, Washington
25, D. C.
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Technical Information Service Extension
Oak Ndge. Tennessee
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ORNL-l845(DeI.)
BY
W. B. Cottrell
H. E. Hungerford
J. K. Leslie
J. L. Meem
September 6, 1955
Work performed under Contract No. W-7405-Eng-26
Oak Ridge National Laboratory
Oak Ridge, Tennessee
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MARTIN MARIETTA ENERGY SYSTEMS LIBRARIES
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ACKNOWLEDGMENTS
The experiment described herein was performed by the
ARE Operations Group under
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the immediate supervision of
other groups i n the
E. S. Bettis and J. L. Meem, with frequent support from
ANP Project.
In particular,
carrier and concentrate loading operation,
W. R. Grimes and his crew effected the
as w e l l as the fluoride sampling;
C. P.
Coughlen and his crew effected the sodium loading and sampling; a group from the
Analytical Chemistry Division, under the supervision of
J. C.
White,
performed a l l
analyses of fluoride and sodium samples; and
E. B. Johnson obtained a reactor power
calibration from analysis of a fuel sample.
Craft foremen
B. H. Webster and J. C.
Packard are both t o be commended for the excellent craftsmanship i n the construction
and maintenance of the system,which made the operation of the system such a success;
and a l l who were in the building during the operation are appreciative o f the concern o f
the Health Physics Group under
R. L.
Clark.
The bulk of t h i s report was prepared by the authors, but considerable assistance
was received from W.
K.
Ergen i n the analysis of the nuclear data.
The work of
numerous others i s referred to i n the various analyses which are presented i n the text
and i n the appendixes.
Cromer, and
In addition, the constructive criticism of
W. H. Jordan, S. J.
E. S. Bettis, each of whom reviewed t h i s report, i s gratefully acknowledged.
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FOREWORD
The operation of the Aircraft Reactor Experiment
endeavor at
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(ARE) culminated four years of
ORNL i n the f i e l d of nuclear propulsion of aircraft.
The f i n a l success
of the experiment i s a tribute t o the efforts of the 300-odd technical and s c i e n t i f i c
personnel who constitute the
ANP
Project at
ORNL, as well as those others who were
less prominently engaged i n t h e fabrication, assembly, and installation of the experiment.
Dr.
The project was capably directed toward t h i s goal during i t s formative years by
R. C. Briant (now deceased), and subsequently by W. H. Jordan and S. J. Cromer,
currently Director and Co-Director, respectively.
This i s the second i n a series of three reports which summarize the
and is concerned primarily w i t h the nuclear operation of the reactor.
ARE experience
The experiment
i s described, and the data obtained during the time from the start of the c r i t i c a l experii
ment u n t i l the reactor was shut down for the last time are analyzed.
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presented i n essentially chronological order, and frequent reference i s made t o appended
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material which includes important detailed data and information that may not be of
general interest.
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The material i s
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CONT EN TS
SUMMARY ...........................................................................................................................................................
..........................
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2. DESCRlPTlON OF T
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..................................
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Fuel System ..............
Sodium System ..........
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................................
Off-Gas System .......
....................
...........................
10
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Fuel Enrichment Sys
............................
..................................
Final Preparations
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Enrichment Procedure .....................................
Early Stages of the Experiment ...........
............................
..........................
......................................
...............
..........................
Subcr i t i c a l Mea sur
Approach to C r i t i c
Calibration of the Shim Rods ............
Measurement of th
5. LOW-POWER EXPERIMENTS ....
.........................................
....................................
.................
...................................
...............................
Radiation Surveys .......................................
..........................................
......................................
....................
................................
..................................
.....................
Calibration of Shiv Rods vs Regulating Rod ......................................
....................................
......................................
Effect of Fuel Flow on Reactivity ........................................
.............................................
Low-Power Measurement of the Temperature Coefficient ................................................
Adjustrrent of Chamber Position ............
........................................
6. HIGH-POWER EXPERIMENTS
...................................
....................
.......................
...............................
...............................................
..........................................
.................
............................................
Fuel and reactor temperature coefficients ..........
Sod iurn temperature coefficient ............
Measurement of the Xenon Poisoning ........
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Power Determination from Heat Extraction
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................................
..................................................
.................................
...................................
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Power c y c l i n g of reactor ....................................................................................................................
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40
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48
49
52
54
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57
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Reactor transients ..................................................................................................................................
Calculated power change resulting from a regulating rod movement ................................................
Reactor temperature differential as a function of helium blower speed ............................................
The phenomenon of the t i m e lag ............................................................................................................
Reactivity effects of transients in the sodium system ......................................................................
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Reactivity following a s c r a m . ................................................................................................................
..
. . . . . .. . . . .
Final Operation and Shutdown .................................................................
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7. RECOMMENDAT IONS .......................................................................................................................
APPENDIXES
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B. SUMMARY OF DESIGN AND OPERATIONAL DATA ..........................................................................
. .
Description ...............................................................................................................................................
Materials .............................................................................................
. . . .
.. .
.
.
Reactor Physics ..........................................................................................................................................
Shielding ...........................................................................................................................................
Reactor Control ...........................................................................................................................................
System Operating Conditions ....................................................................................................................
Miscellaneous ..............................................................................................................................................
, ,, , ,, ,
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C. CONTROL SYSTEM DESCRIPTION AND OPERATION ......................................................................
Control System Design ................................................................................................................................
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Instruments Description .............................................................................................................................
Controls Description ......................................................................
.......................................................
Console and Control Board Description .................................................................................................
Control Operations ................................................................................................
.............................
Reactor Operation ........................................................................................................................................
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125
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129
130
135
D. NUCLEAR OPERATING PROCEDURES ..............................................................................................
Addition of Fuel Concentrate ....................................................................................................................
..
Subcritical Experiments ..............................................................................................................................
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In i t i a I Cr .it i ca
I it y .....................................................................................................................................
Rod Calibration v s Fuel Addition ..........................................................................................................
Low-Power Experiments ...........................................................................................................................
Approach to Power ...................................................................................................................................
Experiments a t Power ...............................................................................................................................
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150
152
E. MATHEMATICAL ANALYSIS OF APPROACH TO CRITICALITY ......................................................
156
F. COLD, C L E A N CRITICAL MASS ....................................................................................
158
G. FLUX AND POWER DISTRIBUTIONS ....................................................................................................
Neutron Flux Distributions ......................................................................................................................
Fission-Neutron Flux Distributions ....................................................................................................
Power Distribution ..................................................................................................................................
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161
162
H. POWER DETERMINATION FROM FUEL ACTIVATION
Theory .......................................................................................................................................................
Experimental Procedure ..............................................................................................................................
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INHOUR FORMULA FOR A CIRCULATING-FUEL REACTOR WITH SLUG FLOW
168
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J. CALIBRATION OF THE SHIM RODS ..................................................................
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174
177
K. CORRELATION OF REACTOR AND LINE TEMPERATURES ......................................................
180
L. POWER DETERMINATION FROM HEAT EXTRACTION ....
184
...............
Calibration from Critical Experiment Data ......................................
..................................
......................................................
Calibration Against the Regulating Rod ............................
..........................................................................
Calibration by Using the Fission Chambers ......
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M. THERMODYNAMIC ANALYSES ........................................................
....................................
Insulation Losses, Heater Power Input, and Space Cooler Performance ................................
Experimental Values of Heat Transfer Coefficients ..............................................
..............
188
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N. COMPARISON OF REACTOR POWER DETERMINATIONS ..................................
190
0. ANALYSIS OF TEMPERATURE COEFFICIENT MEASUREMENTS ..................................................
..............................................................
Importance of the Fuel Temperature Coefficient
Effect of Geometry in the ARE
............................................................
........................
Time Lag Considerations ............................................................
.........................................
Subcritical Measurement of Temperature Coefficient .................................................................
Low-Power Measurements of Temperature Coefficie
High-Power Measurements of Temperature Coefficients of Reactivity ......................
192
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P. THEORETICAL XENON POISONING ....
200
....................................................................
Q. OPERATIONAL DIFFICULTIES ......
.................................................................
Enrichment System ..........................................................
...............................................................
Process Instrumentation .
...........................
Nuclear Instrumentation a
Annunciators ................................
..................................................................
..................................................
Heaters and Heater Controls ................................................
System Components ..........
Leaks .................................
.........................
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R. INTEGRATED POWER .._
Extracted Power _ . ....................................................
Nuclear Power ..............................
................................................................
..........................
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INTERPRETATION OF OBSERVED REACTOR PERIODS DURING TRANSIENTS ........................
212
S.
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T. NUCLEARLOG ..........................
....................................................................
U. THE ARE BUILDING ......
...............................................
BIBLIOGRAPHY ....................
...........................................
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SUMMARY
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The Aircraft Reactor Experiment (ARE) was operated successfully and without untoward d i f f i c u l t y
i n November 1954. The following statements summarize the notable information obtained from the
experi ment.
1. The reactor became c r i t i c a l with a mass of
32.8 Ib o f UZ3’, which gave a concentration of
23.9 Ib o f UZ3’ per cubic foot o f fluoride fuel.
For operation a t power, the U235 content of the
fuel mixture was increased t o 26.0 Ib/ft3, and thus
the final composition of the fuel mixture was 53.09
mole % NaF, 40.73 mole % ZrF,,
and 6.18 mole 5%
UF,.
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2. The maximum power level for sustained operation wos 2.5 Mw, with a temperature gradient of
355OF; the maximum fuel temperature a t this level
was 158OOF. Temperatures as high as 162OOF were
recorded during transients.
3. From the time the reactor f i r s t went c r i t i c a l
until the final shutdown, 221 hr hod elapsed, and
for the final 74 hr the power was i n the megawatt
range (0.1 to 2.5 Mw). The total integrated power
was about 96 Mw-hr.
4. While a t power the reactor exhibited excellent
stability and it was easily controlled because of
i t s high negative temperature coefficient of reactivity, which made the reactor a slave to the
load placed upon it. The fuel temperature coefficient was -9.8 x lo-’ ( A k / k ) / O F , and the overa l l coefficient for the reactor was 4.1x lo-’.
5. Practically a l l the gaseous fission products
and probably some of the other volatile fission
products were removed from the circulating fuel.
I n a 25-hr run a t 2.12 Mw the upper l i m i t of the
reactor poisoning due t o xenon was 0.01% A k / k .
No more than 5% o f the xenon stayed i n the molten
fluoride fuel.
6. The total time o f operation a t high temperature (1000 to 16OOOF) for the sodium circuit was
635 hr, und, for the fluoride fuel system, 462 hr.
During most o f the operating period the sodium
was circulated at 150 gpm and the fuel at 46 gpm.
7. The fabricability and compatibility of the
materials system, i.e., fluoride fuel, sodium coolant,
and lnconel structure, were demonstrated, a t least
for the operating times, temperatures, and flux
levels present.
8. All components and, with few exceptions, a l l
instrumentation performed according t o design
specifications. The performance of the pumps was
particularly gratifying, and the low incidence of
instrumentation failure was remarkable i n view o f
the quantity and complexity of the instruments
used.
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1. INTRODUCTION
T h e Aircraft Nuclear Propulsion (ANP) project
a t the Oak Ridge National Laboratory was formed
in the f a l l of 1949, a t the request o f the Atomic
Energy Commission, t o provide technical support
t o existing A i r Force endeavors in the field. The
effort gradually expanded and, following the
recommendation of the Technical Advisory Board
ORNL
i n the summer of
1950, was directed toward the
construction and operation o f an aircraft reactor
A complete description o f the ARE
experiment.
fa1Is naturally into three categories that correspond
to the three phases of the project: (1) design and
installation, (2) operation, and (3) postoperative
examination. Each of these phases i s covered by
a separate report, ORNL- 1844, ORNL-1845, and
ORNL- 1868, respectively. Much detailed information pertaining t o the selection of the reactor
type and t o the design, construction, and prenuclear operation o f the reactor experiment w i l l be
A s the t i t l e of t h i s
presented in ORNL-1844.
report (ORNL- 1845) indicates it i s concerned
primarily with the operation of the experiment, and
only insofar as they are necessary or useful t o the
understanding or evaluation of the nuclear operation are design and preliminary operational data
included herein.
The third report (ORNL-1868)
w i l l describe the aftermath o f the experiment, with
particular reference to corrosion, radiation effects,
effects that cannot be
and the decay of activity
evaluated at t h i s time because of the high level 04
the radioactivity of the equipment.
-
The
specific operating obiectives were to
attain a fuel temperature o f 15OO0F, with a 3500F
temperature r i s e across the reactor, and to operate
the system for approximately 100 Mw-hr. Other
objectives o f the experiment were t o obtain as
much experimental dota as possible on the reactor
operational characteristics. The, extent t o which
each of these objectives was f u l f i l l e d i s described
herein, anda measure of the success of the program
i s thus provided.
Although it was i n i t i a l l y planned t o use a
2
sodium-cooled,
sol id-fuel-element reactor, the
reactor design evolved f i r s t to that o f a sodiumcoo I ed, stati onary-l i qui d-fuel reactor and, finally,
to that of a circulating-fuel reactor employing
sodium as a reflector coolant. These evolutionary
processes l e f t their mark on the experiment, particularly i n that the reactor had t o incorporate a
moderator geometry that was originally specified
and ordered for the sodium-cooled reactor. The
adaptation o f t h i s moderator geometry to the
circulating-fuel reactor resulted i n a reactor in
which the fuel stream was divided into s i x parallel
circuits, each of which made numerous passes
through the core. These fuel passages were not
a condition which caused considerable
drainable
concern throughout the course o f the experiment.
Although it i s not the purpose o f t h i s report t o
give a detailed history of the design, construction,
or preliminary testing of the reactor system, the
final design and pertinent prenuclear operation are
briefly described. The bulk of the report concerns
the operation of the experiment from t h e time
uranium was added to the fuel system on October 30
u n t i l the evening o f November 12, when the reactor
In addition to
was shut down for the l a s t time.
the description of the various experiments and
analyses of t h e data which are presented in the
body o f the report, the entire nuclear operation,
as recorded i n the “Nuclear Log,” i s given i n
Appendix T. The report also includes a number
of recommendations based on the operating experience, and much detailed supporting data and
information not appropriate for inclusion i n the
body of the report are given in the other appendixes.
The various experiments that were performed
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on the ARE were designated as E, L, or H series,
depending upon whether they occurred during the
c r i t i c a l experiment, low-power operation, or highpower operation, respectively.
experiments, as listed below,
t h i s report.
Each o f these
i s discussed i n
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Experiment
T y p e of Experimenf
Series No.
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E-1
C r i t i c a l experiment
E- 2
Subcritical measurement of reactor temperature c o e f f i c i e n t
L- 1
Power determination a t
L-2
Regulating r o d c a l i b r a t i o n v s f u e l a d d i t i o n
L-3
F u e l system c h a r a c t e r i s t i c s
L-4
Power determination a t
L-5
R e g u l a t i n g rod c a l i b r a t i o n v s reactor period
L-6
C a l i b r a t i o n of shim r o d vs r e g u l a t i n g rod
L-7
E f f e c t of f u e l f l o w on r e a c t i v i t y
L-8
Low-power measurement of reactor temperature c o e f f i c i e n t
L-9
Adjustment of chamber p o s i t i o n
H- 1
Approach t o power:
H-2
T e s t o f off-gas system
H-3
Approach t o power:
H-4
High-power measurement of t h e f u e l temperature c o e f f i c i e n t
H-5
High-power measurement of t h e reoctor temperature c o e f f i c i e n t
H-6
Reactor startup an temperature c o e f f i c i e n t
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H-7
Sodium temperature c o e f f i c i e n t
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H-8
E f f e c t of a d o l l a r of r e a c t i v i t y
H-9
High-power measurement of reactor temperature c o e f f i c i e n t
H- IO
Moderator temperature c o e f f i c i e n t
H-11
Xenon run at full power
H- 12
R e a c t i v i t y e f f e c t s of sodium f l o w
H-13
Xenon b u i l d u p a t one-tenth f u l l power
H- 14
Operation at maximum power
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2. DESCRIPTION OF THE REACTOR EXPERIMENT
The ARE consisted of the circulating-fuel
reactor and the associated pumps, heat transfer
equipment, controls, and instrumentatiorr required
for i t s safe operation. A schematic arrangement of
The
the reactor system i s shown i n Fig. 2.1.
major functional parts of the system are discussed
briefly below. The physical plant i s described in
Appendix U. A detailed description of the reactor
and the associated system may be found i n the
design and installation report.' A summary of the
design and operational data, including a detailed
flow sheet of the experiment, i s given i n Appendix B.
REACTOR
The reactor assembly consisted of a 2-in.-thick
lnconel pressure shell i n which beryllium oxide
moderator and reflector blocks were stacked around
fuel tubes, reflector cooling tubes, and control
assemblies.
Elevation and plan sections of the
The
reactor are shown i n Figs. 2.2 and 2.3.
innermost region of the l a t t i c e assembly was the
core, which was a cylinder approximately 3 ft i n
diameter and 3 ft long. The beryllium oxide was
machined into small hexagonal blocks which were
s p l i t axially and stacked to effect the cylindrical
core and reflector. Each beryllium oxide block i n
the core had a 1.25-in. hole d r i l l e d a x i a l l y through
'Design and Installafion of the Aircraft Rearfor E x periment, ORNL-1844 (to be issued).
i t s center for the passage of the fuel tubes. The
outer 7.5 in. of beryllium oxide served a s the
reflector and was located between the pressure
shell and the cylindrical surface of the core. The
reflector consisted of hexagonal beryllium oxide
blocks, similar t o the moderator blocks, but w i t h
0.5- in. ho I es.
The fuel stream was divided into s i x p a r a l l e l
circuits at the i n l e t fuel header, which was located
above the top of the core and outside the pressure
shell. These c i r c u i t s each made 11 series p a s s i x
through the core, starting close to the core axis,
and progressing in serpentine fashion to the
periphery of the core, and f i n a l l y leaving the core
through the bottom of the reactor. T h e s i x c i r c u i t s
were connected t o the outlet header. Each tube
was of 1.235-in.-OD seamless lnconel tubing w i t h
a 60-mil wall.
The combination o f p a r a l l e l md
series fuel passes through the core was largely
the result of the need for assuring turbulent flow
i n a system i n which the f l u i d properties and tube
dimensions were fixed.
The reflector coolant, i.e., sodium, was admitted
into the pressure shell through the bottom. The
sodium then passed up through the reflector tubes,
bathed the inside walls of the pressure shell,
f i l l e d the moderotor interstices, and l e f t from the
plenum chamber at the top of t h e pressure shell.
The sodium, i n addition to cooling the reflector
and pressure shell, acted as a heat transfer medium
DWG 14562b
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BLOWER
Fig,
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2.1.
Schematic Diagram of the Aircraft Reactor Experiment.
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in the core by which moderator heat was readily
transmitted t o the fuel stream.
FUEL SYSTEM
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The fuel was a mixture o f the fluorides o f sodium
and zirconium, with sufficient uranium fluoride
added t o make the reactor c r i t i c a l . While the fuel
ultimately employed for the experiment was the
NaF-ZrF,-UF,
mixture with a composition o f
53.09-40.73-6.18 mole %, respectively, most preliminary
experimental
work
(i.e.,
pump tests,
corrosion tests) employed a fuel containing someT h e fuel was circulated around
what more UF,.
a closed loop from the pump t o the reactor, t o the
heat exchanger, and back t o the pump. An isometric
drawing of the fuel system i s given in Fig. 2.4.
The fuel pump was a centrifugal pump w i t h a
vertical shaft and a gas seal, as shown in Fig. 2.5.
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Fig. 2.2.
The Reactor (Elevation Section).
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Fig. 2.3. m e Reactor (Plan Section).
The fuel expansion volume around the impeller
cavity provided the only liquid-to-gas interface
i n the fuel system. While speeds up t o 2000 rpm
could be attained with the 15-hp d-c pump motor,
the desired fuel flow of 46 gpm was attained at
Although only one pump
a speed of 1080 rpm.
was used i n the experiment, a spare fuel pump,
isolated from the operating pump and i n parallel
with it, was provided.
From the pump the fuel flowed to the reactor,
where i t was heated, then t o two parallel fuel-tohelium heat exchangers, and back to the pump.
The cycle time was about 47 sec at f u l l flow, of
6
which approximately 8 sec was the time required
for the fuel to pass through the core. The two
fuel-to-helium heat exchangers were each coupled
to a helium-to-water heat exchanger.
The heat
extracted from the fuel was transferred via the
the ultimate
helium to water, and the water
was discharged. The helium flow rate
heat sink
i n the fuel-to-helium heat exchanger loop was
control led through a magnetic clutch that coupled
the blower to a 50-hp motor. Control of the helium
flow rate i n t h i s manner permitted smooth control
at any reactor power at which the heat generation
was great enough for the temperature coefficient
to be the controlling factor.
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ORNL-LR-DWG 6218
DRIVE SHEAVE
UPPER SEAL OIL RETAINER
OIL BLEED NIPPLE
PER SEAL LEAKAGE DRAIN)
--
CLAMP RING
BEARING HOUSING
TOP SHAFT SEAL A
SEAL FACE RING
INLET
-OIL
H BREATHER AND
T GLASS GAS EQUALIZER
INTERNAL SHAFT COOLING
LUBE OIL GUIDE TUBE - ~-
FUEL INJECTION SYSTEM NOZZLE
SEAL FACE RIN
OIL OVERFLOW
RESISTANCE
HEATER CONNECTION
OIL DRAIN HEADER CONNECTION
THERMOCOUPLE GLAN
3/8-16 NC HEX HD
P SCREW 1-1" LONG
STILLING WELL
I
t
'BAFFLE
PLATE
%D
-P
I
DISCHARGE GRAVITY
Fig. 2.5.
AS shown i n Fig. 2.4, the fuel system was
connected w i t h two f i l l tanks (only one o f which
was used) and one dump tank which had provisions
for removing the afterheat from the fuel.
The relatively high melting point o f the circulating fuel (about 1000°F for the NaF-ZrF,-UF,
required that a l l
fuel containing 6.18% UF,)
equipment within which it was circulated be
heated sufficiently t o permit loading, unloading,
and low-power operation.
T h i s heating was accomplished by means of electrical heaters attached
t o a l l components o f the fuel and sodium systems;
i.e., pressure shell, heat exchanger, pumps, and
tanks, as well as a l l fuel and sodium piping.
a
t .,
INLET
LEG
Fuel Pump.
In addition t o the heaters and insulation, a l l
fuel and sodium piping was surrounded by a 1- t o
1 '/,-in. annulus (inside the heaters) through which
helium was circulated.
T h e helium circulated
through the annuli was monitored a t various
stations around the system for evidence o f leaks
and was, in addition, a safety factor in that i t
could be expected t o keep hot any spots at which
heater failures occurred.
SODIUM S Y S T E M
The sodium circuit external t o the reactor,
shown in Fig. 2.6, was similar t o that of the fuel.
The sodium flowed from pump t o reactor, t o heat
0
Y
I
c
t
exchanger, t o pump.
The sodium pumps were
gas-sealed centrifugal pumps o f the same design
as the fuel pump, except that a smaller expansion
volume was provided. Again, there were two sodium
pumps, one o f which was a spare i n parallel with
the operating pump.
The sodium was circulated through the two
parallel sodium-to-helium heat exchangers after
being heated in the reactor. Again the heat was
transferred v i a the helium t o water i n two heliumto-water heat exchangers. The sodium system was
equipped with heaters, as was the fuel system,
and a l l piping was surrounded with the helium
annulus for leak monitoring and heat distribution.
PROCESS I N S T R U M E N T A T I O N
Since a basic purpose of the ARE was the
acquisition of experimental data, the importance
of complete and reliable instrumentation could
not be overemphasized.
Therefore there were
27 strip temperature recorders (mostly multipoint),
5 circular temperature recorders, 7 indicating flow
controllers, 9 temperature indicators (with up t o
96 points per instrument), about 50 spark-plug
level indicators, 20 ammeters, 40 pressure gages,
16 pressure regulators, and 20 pressure transmitters.
In addition there were numerous flow recorders,
indicators, alarms, voltmeters, tachometers, and
assorted miscellaneous instruments.
Most ARE process instrumentation was installed
t o permit observing and recording rotational speeds,
flow rates, temperatures, pressures, or liquid
levels. The operating values o f temperature, pressure, and flow at various stations around the ARE
f l u i d circuits are given i n Appendix B. Since most
commercially available instruments for measuring
flow and pressure are subject to temperature
limitations considerably lower than the minimum
operating temperature of the ARE and they employ
open lines i n which the ZrF, vapor could condense, they were not suitable for ARE applications.
Therefore a bellows type of device was adapted
for use as a pressure indicator, and a fluidimmersed inductance type o f instrument was
developed for measuring sodium and fuel l i q u i d
level and fuel flow.2 Sodium flow was measured
by an electromagnetic flowmeter.
Convent iona I
chromel-alumel thermocouples welded t o the pipe
walls were used t o determine temperatures. The
instrumentation flow sheet i s shown in Fig.
2.7.
Most of the indicating and recording instruments
were located in the control room. While numerous
important temperatures were recorded i n the control
room, about 75% o f the thermocouples were indicated (or recorded) only i n the basement.
NUCLEAR INSTRUMENTATION AND CONTROLS
Detailed information on the important nuclear
instrumentation and controls i s presented in ApBriefly, the nuclear instrumentation
pendix C.
(such as fission chambers and ion chambers), the
electronic components (such as preamplifiers and
power amplifiers), and the control features o f the
ARE were similar, and in many cases identical,
t o those of the MTR, the LITR, and other reactors
However, since the fission
now in operation.
chambers were located i n a high-temperature
region within the reflector, it was necessary t o
develop
special
high-temperature
chambers.
Helium was used t o cool them t o below 600OF.
The locations of the two fission chambers in the
reactor may be seen from examination of Fig. 2.3;
the locations of the remaining ion chambers are
Since the fission chambers
shown on Fig. 2.8.
moved through sleeves i n the reactor which were
parallel t o those for the regulating and shim rods,
it was convenient to group the drive mechanisms
with those for the rods i n an igloo outside the
shielding above the reactor, as shown i n Fig. 2.8.
A l l other ion chambers were external t o the reactor
pressure shell; they viewed the reactor a t midplane; and they were mounted so that they would
move horizontally.
These chambers included a
BF, counter (for the c r i t i c a l experiment), two
parallel circular plate (PCP) chambers, and two
compensated chambers.
There were two separate elements t o the control
the single regulating rod and the three
system
shim rods.
T h e locations o f these rods i n the
reactor may be seen from examingtion o f Fig. 2.3.
The regulating rod had a vertical movement o f
12 in. about the center o f the reactor and was
fabricated of stainless steel. Several such regulating rods, with varying amounts of metal, were
fabricated, and the one finally used in the experiment had a total value of 0.4% Ak/k for the 12-in.
movement and could be moved a t a rate o f 0.011%
(Ak/k.)/sec.
-
~
2For greater detail see A N P Quar. Prog. R e p . Sept.
10, 1952, ORNL-1375, p 23.
10
The three safety rods were located i n the core
on 120-deg points at a radius o f 7.5 in. from the
,
7
c
ORNL-LR-DWG
i
Fig.
2.8.
Isometric Drawing of Control Rods and Nuclear Instrumentation.
6224
center o f the reactor. Each rod was made up of
slugs o f a hot-pressed mixture of boron carbide
and iron; the slugs were canned in stainless steel.
The canned sections were slipped over a flexible
tube. The total rod travel was 36 in., and 13 min
was required for withdrawal of the rod. Thus, the
5.8% A k / k in each rod (total for the three was
approximately 17% A k / k ) could be withdrawn a t
a rate of 0.45% (Ak/k)/min per rod.
A small amount of the helium used t o cool the
fission chambers was directed through the shim
and regulating rod holes, but the helium flow was
so low that not much cooling was effected. T h i s
helium was circulated i n an independent circuit,
known as the rod cooling system, which had two
two-speed helium blowers i n parallel and three
helium-to-water heat exchangers.
T h e helium
circulated from the blowers through the rod sleeves
and fission chamber sleeves t o the heat exchangers and back to the blowers, and i t removed,
i n the process, up t o 35 kw of reactor heat. Part
of the helium returning from the reactor was
diverted through pipe annuli before reaching the
heat exchanger.
OFF-GAS SYSTEM
The off-gas system was designed to permit the
collection, holdup, and controlled dischqrge o f
the radioactive fission products that were evolved
as a consequence of reactor operation. Although
only the gases above the fuel system were expected t o have significant activity, the gas from
the sodium system was also discharged through
the off-gas system. A s shown i n Fig. 2.9, there
was, i n addition to the primary off-gas system, an
auxiliary system for discharging gases from the
p i t through nitrogen-cooled charcoal tanks.
In the primary system the o f f gases from both
the fuel and the sodium were directed t o a common
vent header. From t h i s header, helium plus the
v o l a t i l e fission gases passed through a NaK
scrubber, which removed bromine and iodine, and
then into two holdup tanks, which were connected
i n series, t o permit the decay of xenon and krypton.
From these the gases were released up the stack.
The release o f gases t o the stack was dependent
14
upon two conditions: (1) a wind velocity greater
than 5 mph and (2) radioactivity o f less than
0.8 pc/cm3. The activity was sensed by a monitor
which was located between the two holdup tanks.
F U E L ENRICHMENT SYSTEM
The fluoride fuel used i n the ARE was amenable
to a convenient enrichment technique i n which the
fuel system was f i r s t f i l l e d with a mixture o f the
fluorides of sodium and zirconium.
When t h i s
mixture, NaZrF,,
had been circulated for a sufficient time t o ascertain that the reactor was ready
t o be taken c r i t i c a l (no leaks, etc.), uranium in
the form o f molten Na,UF,
was added t o the
fluorides then i n the system.
The temperature
contour diagram of the NaF-ZrF,-UF,
system i s
shown i n Fig. 2.10. The melting point of Na,UF,
was 12OO0F, and there were no mixtures of higher
melting points on the join between Na,UF,
and
NaZrF,, which i s shown i n Fig. 2.11. The addition
o f Na,UF,
raised the melting point o f the mixture
only slightly above that of NaZrF, (955OF).
It was i n i t i a l l y intended that the enrichment
operation would consist of the remote addition o f
Na,UF,
to the system from a large tank which
contained a l l the concentrate.
From the large
tank, the concentrate was t o pass through an
intermediate transfer tank which would transfer
about 1 qt or less a t a time. T h e intermediate
tank was suspended from a load beam so that the
weight of the concentrate could be determined
before each addition o f enriched material. However,
t h i s system was discarded when preoperational
tests proved the temperature control of the system
t o be inadequate; i n addition, the accuracy of the
weight-measuring instrumentation on the transfer
tank was uncertain.
In l i e u o f the original enrichment system, a less
elaborate, more direct method of concentrate addition was employed. T h e enrichment procedure
actually used involved the successive connection
of numerous small concentrate containers t o an
intermediate transfer pot, which, i n turn, was
connected t o the fuel system by a l i n e which
injected the concentrate i n t o the pump tank above
the liquid level.
i
i
I
,
k
i
!
a
d
t
?
i
j
DWG. 24453A
?
ZrF,
NaF
d
745
040
625
Na3ZrF7 Na2ZrF6
Fig. 2.10.
515
NaZrF5
570
Na3Zr4F,9
The System NaF-ZrF,-UF,.
17
675
L
b
650
625
W
a 575
H
W
c
550
525
L
p.
i
500
NoZrF5
40
20
Fig, 2.11.
30
40
50
60
No2UF6 ( m o l e % )
70
00
90
NOZUF6
The Pseudo-Binary System NaZrF5-No2UF6.
..
1
18
r
3. PRENUCLEAR OPERATION
It i s not the purpose here to describe i n detail
the preliminary checks and shakedown tests which
preceded the nuclear operation of the system because t h i s information i s covered i n the design
However, some underand installation report.
standing of the scope and significance o f these
(8
preliminary" tests i s desirable, particularly since
at the conclusion o f t h i s phase of the operation
the fuel and sodium systems were l e f t f i l l e d with
clean carrier and sodium, respectively, and the
fuel system was ready for the fuel enrichment
operation which comprised the c r i t i c a l experiment.
The prel iminary tests accomplished many purposes, the more important being the cleaning and
leak-checking of the systems, performance and
functional testing of components, checking and
calibrating of instrumen.ts, and the practicing o f
operational techniques. The system, as an integral
unit, was sufficiently assembled by August 1 for
the operating crews to be placed on a three-shift
basis so that intensive tests on the var'ious systems, i n i t i a l l y with water as the circulating fluid,
could be started. There were, however, frequent
delays while changes and modifications that the
testing indicated as being desirable or necessary
were made.
The operation with sodium, i n particular, was responsible for major changes i n the
sodium vent system and the elimination of the
sodium purification system.
Tests with sodium
and carrier i n the systems were concluded on
October 30, at which time the system was ready
for the c r i t i c a l experiment.
'
8
i
I
J
4
CIRCULATION
d
1
f '
O F SODIUM
By the last week i n September, a l l temperature
tests preliminary to operation o f the sodium system
had been completed. It was necessary t o operate
the sodium system before operating the fuel system
because the sodium was used t o heat the reactor
t o above the fuel melting point. I n preparation for
f i l l i n g the sodium system, the temperature o f the
entire system was brought t o 600°F and sodium
was transferred from portable drums into the
The sodium used had been
system fill tanks.
carefully filtered to minimize the oxygen content,
which at the time of the f i l l i n g averaged about
0.025 wt 5%.
' D e s i g n and Installation of the Aircraft Reactor Experiment, ORNL-1844 (to be issued).
The system was f i l l e d w i t h sodium on the morning
of September 26. No particular d i f f i c u l t i e s were
encountered during the f i l l i n g or when the sodium
pumps were operated a t design speed, although
there was evidence of trapped gas i n the system.
The electromagnetic flowmeters, which were a t
f i r s t inoperative, functioned properly after sufficient time had elapsed for wetting o f the pipe
wall to occur. On the following afternoon a leak
developed i n a tube bend in the sodium purification
system, so the sodium was dumped.
Subsequent analyses revealed that the leak
occurred where a thermocouple pad had been
heliarc welded t o the tube. The excessive weld
penetration which caused the leak was attributed
t o a lack of instruction to the welder that a variation i n pipe wall thickness existed in t h i s section.
As a consequence of t h i s leak and d i f f i c u l t i e s
experienced during draining of the sodium, a
number o f changes were made i n the sodium system.
The purification systems were eliminated
and provisions were made for better heat control
for a l l vent lines and valves t o prevent freezing
o f the sodium and plugging of the lines. It was
concluded that the sodium purification system
could be removed without endangering the experiment because experience i n operating the system
had proved that the oxygen contamination was very
low; furthermore, there was no danger of the oxide
plugging the heat exchanger tubes because of the
low temperature differential in the system and the
large diameters of the tubes i n the heat exchangers.
It was felt, also, that the purification system
represented the weakest link i n the f l u i d c i r c u i t
because thinner walled tubing had been used there
than anywhere else i n the system, except for the
reactor tubes, which were assembled w i t h extreme
care. Even so, the thin-walled tubing i n the purification system would probably not have presented
a problem had the thermocouples been properly
welded.
On October 16 the sodium was recharged into
the system a t 6OO0F, circulated, and dumped four
times to thoroughly check the operability o f the
sodium system.
Analyses of the sodium taken
both at 600 and 925°F showed the oxygen content
to be satisfactorily low (of the order o f 0.026 wt %),
and therefore the i n i t i a l batch o f sodium was not
replaced. The sodium was circulated for several
days at 1300°F before fuel carrier, NaZrF5, was
4
19
?
i
I
i
added t o the fuel tank. During the period o f sodium
circulation, the typical system characteristics
were those given i n Figs. 3.1 and 3.2. The data
were comparable to those obtained during the water
tests.’ Such discrepancies as existed were within
experimental error and could easily have been due
t o changes i n the system that were made during
the time between the two tests, such as the removal of the purification systems and the removal
of the bypass around the reactor during the water
tests.
During and after the second sodium loading
operation the copper gauze in the helium stream
was monitored and no evidence o f sodium leakage
With helium i n the fuel system the
was found.
temperatures i n the fuel and sodium systems were
0 ‘
0
Fig.
40
20
3.1.
60
80
I
I
t
100 120 140
FLOW RATE (gprn)
160
180
I
200
Pump Speed vs Sodium F l o w Rate.
56
MAIN Na PUMP
48
-
40
L
n
$32
W
[r
3
rn
W
v,
24
DATA CORRECTED FOR
CALCULATED INSTRUMENT
ZERO SHIFT
~
(6
,
-~
-
I
0
0
Fig.
20
3.2,
/ *
~
’ . .-I
40
60
’
I
I
DATA
&ACTUAL
-1
8
20
.’
/
/
1I
I
j
l
80 100 120 I40
FLOW RATE (gprnl
~
1
I
160
180
2
Pressure H e a d vs Sodium F I o w Rate,
then gradually raised (about 10°F/hr) t o 13OOOF.
After the 13OOOF temperature was reached, the
helium pressure i n the fuel system was reduced
to 0.5 psig and sufficient krypton was added t o
bring the system pressure up t o 5.0 psig. The
helium i n the annuli was then sampled and analyzed
for krypton, and no evidence o f a leak was found.
CIRCULATION O F FUEL CARRIER
With the system isothermal a t 13OO0F, the fuel
carrier, NaZrF,,
was loaded into the fill tank a t
12006F, and the fuel system was evacuated t o
facilitate raising the fluoride into the system.
T h i s f i l l i n g operation, which occurred on October
25, required only a few minutes once the desired
vacuum was attained. During the f i l l i n g operation
it was possible to follow the progress o f the
fluoride through the system by watching the thermocouple indications along the pipe because o f the
100°F temperature difference between the i n i t i a l
fluoride temperature and the system temperature.
In contrast t o the sodium system, there did not
appear t o be any gas trapped in the fuel system
after loading; evidence o f t h i s was that there was
no l i q u i d level change when the pressure i n the
pump was changed, and the temperature differentials
across the six parallel fuel passages i n the reactor
changed simultaneously when a temperature perturbation was put into the system.
F i v e samples of the carrier were taken, and the
analytical results from each showed the concentrations of impurities and corrosion products given
i n Table 3.1.
There was twice as much carrier available as
was required t o fill the system, and i f chemical
analyses o f samples o f the carrier after it had
been circulated for about 50 hr had indicated too
high a corrosion-product buildup ( a s could have
resulted i f the system had not been adequately
cleaned), the carrier i n the system would have
been dumped and replaced with the extra carrier.
However, from the very favorable analyses of the
f i r s t two samples it was apparent that replacing
the carrier would not be necessary. (The chromium
content would have had to have been 500 ppm to
have necessitated the replacement. The corrosion
mechanism that defines the upper l i m i t o f the
chromium content i s described i n detail i n ORNL-
1844. ’)
During the time the carrier was being circulated
and before the enrichment operation commenced,
the system characteristics were determined. The
i
h
c
i
f
T A B L E 3.1. ANALYSES
Impurities and Corrosion Products Found (ppm)
Running T i m e
Ti me
Date
7
OF IMPURITIES AND CORROSION PRODUCTS I N CARRIER SAMPLES
(hr)
Cr
Fe
Ni
10/25
1650
9.7
90
<5
50
10/26
0214
19.1
81
<5
40
1911
36.1
81
<5
6
10/27
1919
60.2
90
18
12
10/29
0935
98.5
102
30
20
ORNL-LR-DWG 639.
4200
30
'
I
1 )I
MAIN FUEL PUMP
TES[C-3,0CTlZ5,4354
(000
-E 800
~
__
25
MAIN FUEL PUMP
TEST C-3, OCT 25, 4954
a
a 600
-
a
W
I
n
PRESSURE HEAD MEASURED FROM
+ 20 - PUMP SUCTION TO REACTOR INLET.
L
W
6393
ORNL-LR-DWG
1
-
--/
~
~
___
0
a
5
1
i5
a 400
3
ul
W
ul
(r
a
200
40
0
0
10
20
30
40
50
60
FLOW RATE (gpm)
Fig.
Rate.
3.3.
Pump Speed vs F u e l Carrier F l o w
5
.
n
0
10
20
30
40
50
60
FLOW RATE ( g p m )
data obtained i n terms of pump speed v s flow and
system head vs flow are given in Figs. 3.3 and
3.4, respectively. These data were considerably
different from those to be expected from an extrapolation of the water data. The discrepancies
are analyzed in more detail i n ORNL-1844,' but
the only completely satisfactory explanation i s
that the system head during the water test was too
high, possibly due to the temporary water Rotameter
installation or to one or more o f the reactor tubes
not being completely f i l l e d so that only partial
flow was obtained through the reactor during the
water test.
f
F I N A L PREPARATIONS FOR
NUCLEAR OPERATION
By t h i s time the neutron source had been placed
i n the reactor and the nuclear instrumentation
Fig.
Rate.
3.4.
Pressure Head vs F u e l Carrier F l o w
checked out. A BF, counter was installed in one
o f the instrument holes external to the reactor i n
order t o check the fission chambers during the
c r i t i c a l experiment.
Pulse height and voltage
curves were measured on a l l three fission chambers
(the two regulars plus a spare) i n order t o establish
the operational plateau for each chamber.
The mechanical operqtion tests o f the regulating
rod and of the three safety (shim) rods were made;
in all, over 25 rod drop tests were performed in
order t o ensure the r e l i a b i l i t y o f the rod insertions
The rod
a t high (1200 to 1300°F) temperatures.
drives were also continuously cycled i n order t o
check the performance o f the gear trains and
21
motors.
During t h i s time the rod magnet faces
were cleaned, and, i n order t o maintain more uniform drop currents, it was decided t o keep the
magnet faces engaged a t a l l times when the rods
were not i n use.
Before the fuel concentrate (Na,UF,)
was charged
t o the enrichment system, a number o f practice
operations were performed with carrier i n the enrichment system. The temperature control o f the
system was inadequate and the operation of the
weigh c e l l on the transfer tank was such that
reproducible weights could not be obtained. Although the enrichment system probably could have
been made t o work, considerable time would have
been required. Accordingly, the original enrichment system was abandoned, and a manually operated system was improvised.
The concentrate
was f i r s t batched down into a number o f containers.
22
The concentrate in each o f these containers was
then added t o the fuel system at the pump after
f i r s t passing through an intermediate transfer pot.
The amount o f concentrate added was determined
by weight measurements o f the concentrate containers before and after use. While t h i s system
was being set up, the operating crews were engaged
i n leak testing the various auxiliary systems,
determining operability o f heater circuits, checking
and reading thermocouples, establishing the performance o f the pump lubrication and cooling systems, checking the operation o f the annulus system
helium blowers, monitoring the annulus helium for
leaks, and checking the operability o f the various
off-gas systems, instrumentation, and the I ike.
By the morning o f October 30, these tests were
sufficiently well in hand for the c r i t i c a l experiment t o commence.
t
t
i
i
J
4. CRITICAL EXPERIMENT
On October 30 the sodium had been circulating
hr, and the fuel carrier about 110 hr. Both
the sodium system and the fuel system were at an
isothermal temperature o f 1300°F, and the fuel
system was ready for the addition o f concentrate
(Na,UF,)
to the carrier (NaZrF,).
The addition o f
the concentrate was necessarily time-consuming
because of the cautious manner i n which each
addition had t o be made during the approach t o
c r i t i c a l i t y ; however, t h i s operation would have
been completed more rapidly had it not been for
unforeseen d i f f i c u l t i e s . The reactor did not reach
c r i t i c a l i t y until 3:45 Phi, November 3, some four
days after the enrichment operation was started.
Unfortunately, much o f the four days was spent i n
removing plugs and in repairing leaks which occurred i n the enrichment line. Since these leaks
were largely the result of the improvised nature
o f the enrichment mechanism, they were not o f
serious con sequence.
The chronology of the c r i t i c a l experiment i s
shown i n Fig. 4.1, and a detailed description o f
the experiment, including a subcritical temperature
coefficient measurement, i s presented below. The
c r i t i c a l experiment, as well as a l l nuclear operations, followed, in general, the pattern prescribed
i n the “ARE Operating Procedures, Part I I , Nuclear
Operation,”’
which i s included i n t h i s report as
Appendix D.
275
i
i
ENRICHMENT PROCEDURE
I
,
-
By the afternoon of October 30 the chemists had
installed and checked out the injection apparatus,
and the systems were ready for the enrichment
operation t o commence. The loading station was
located directly above the main fuel pump. The
injection r i g consisted o f an oven i n t o which the
batch cans fitted, an intermediate, heated transfer
pot capable o f holding about 5.5 I b o f Concentrate,
a resistance-heated transfer l i n e connecting the
batch can with the transfer pot, and the electrically
heated injection l i n e running from the transfer pot
into the pump, as well as various auxiliary equipment, such as helium supply and vent lines,
vacuum pump, pressure control, and electrical
heating equipment. The resistance-heated transfer
l i n e was l e f t exposed t o the air so that blow
?
1
’J. L. Meem, A R E Operating Procedures, Part 11,
Nuclear Operation, ORNL CF-54-7-144
( J u l y 27, 1954).
torches could be applied i n case o f a plug. A flow
sketch of the injection system i s shown i n Fig.
4.2.
The fuel concentrate was batched i n t o small
cans prior t o loading. The three batch sizes are
given i n the following tabulation:
Can Size
Approxi mate
Concentrate Weight
(Ib)
A
30
B
10
C
0.5
The A and B size cans were used during the c r i t i cal experiment. The C size cans were held for
use i n calibrating the regulating rod. Table 4.1
l i s t s the batch cans by numbers in the order i n
which the cans were used in the c r i t i c a l experiment, the net amount o f concentrate and uranium
added, and the U235content.
Prior to an injection a batch can was selected,
set i n the oven, and heated t o 14OOOF. Meanwhile
the transfer lines were being heated. When proper
temperatures were reached, an injection was
attempted.
The intermediate transfer pot was
f i l l e d from the batch can by pressurizing the batch
can and venting the transfer pot. At the same time
the pump pressure was maintained above that o f
the transfer pot so that there could not be an inadvertent transfer into the pump.
When the spark
plug probes indicated that the transfer pot was
full, the pressure i n the batch can was released,
and the material i n the transfer pot was forced into
the pump by helium pressure.
T h i s method of
transfer was used so that only a measured amount
o f concentrate (not more than 5.5 Ib) could be
added at a given time and the system pressures
With low system pressures
could be kept low.
there was less need for venting and possibly
clogging the vent lines. Figure 4.3 shows a photo
o f the chemists preparing for an injection during
the c r i t i c a l experiment.
E A R L Y STAGES
OF THE EXPERIMENT
A t 1500 on October 30 the c r i t i c a l experiment
was started. The main fuel pump was f i r s t trimmed
t o i t s minimum prime level and the shaft speed
At this
was reduced t o 500 rpm (20-gpm flow).
23
4
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TIME OF DAY
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ORNL-LR-DWG
TO HELIUM SUPPLY
6395
VENT
TO HELIUM SUPPLY
TRANSFER LINE
\ELECTRICAL
CONNECl IONS
FOR RESISTANCE HEATING
I
ELECTRIC HEATERS
.
#
TRANSFER POT
4
FLOOR LEVEL
SAMPLING LINE
t
SUCTION
Fig.
T A B L E 4.1.
Date
J
Time
4.2.
Equipment for Addition of F u e l Concentrate t o F u e l System.
FUEL CONCENTRATE BATCHES ADDED DURING CRITICAL EXPERIMENT
Batch
Can
No.
Concentrate Added
g
Ib
Uranium Added
wt
%
Weight of U235 Added
9
g
Ib
10/30
1625
A-6
13,609
30.002
59.548
8104
7569
16.687
11/1
1415
A-7
11,510
25.375
59.513
6850
6398
14.105
1804
A-8
13,852
30.538
59.530
8246
7702
16.980
2203
A-9
13,806
30.437
59.587
8227
7684
16.940
0213
A-1 0
13,638
30.066
59.454
8108
7573
16.695
0610
A-1 2
13,241
29.191
59.53 1
7882
7362
16.230
083 1
A-5
8,22 1
18.124
59.637
4903
4579
10.095
0523
A-1 1
4,505
9.932
59.671
2688
2511
5.536
0941
8-22
6,432
14.180
59.702
3840
3587
7.908
1245
8-3 1
6,312
13.915
59.637
3764
3516
7.751
1536
8-20
4,567
10.068
59.529
2719
2540
5.600
I
11/2
11/3
1
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1
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exactly 2 min, and the volume of the fuel system,
a check on the rate of fuel flow was obtained:
time the volume of carrier in the fuel system was
calculated to be 4.82 ft3, and the weight was
927 Ib. At 1507, concentrate transfer was started
from batch can A-6 to the intermediate tank, and,
at 1539, transfer o f the f i r s t 5.5 Ib of fuel to the
system was accomplished. A t 1554, the second
5.5 I b o f fuel had been added. These fuel injections
f
4.82 (ft3) x 7.48 (gal/ft3)
(gpm) =
=
2 (min)
18.03
.
This checked fairly well with the value of about
read from the fuel flow recorder.
The remaining four injections from can A-6 were
accomplished smoothly. The pumps were speeded
up to an observed fuel flow rate of 46 gpm t o
obtain better mixing of the fuel.
A t 1720, fuel
sample 6 was removed for analysis. 2
At 1920 the f i r s t 5.5-lb iniection from can A-7
was started, but the transfer l i n e from the intermediate pot to the pump clogged due t o concentrate
freezing i n it. After several hours 04 unsuccessful
attempts to free the line, a gas leak occurred i n
the transter l i n e a t the intermediate tank pot, and
very noticeably affected the fission chamber recorders. Mixing of the concentrate with the carrier
did not occur rapidly, and therefore each time the
20 gpm
enriched slug entered the reactor i t produced a
multiplication that was observable on the fission
chamber count-rate recorder. Figure 4.4 i s a photograph of the trace o f fission chamber No. 1; the
pips that occurred during addition of the f i r s t and
second slugs of fuel are readily observoble. From
the time interval between pips, which wos almost
~~
2 T h e first f i v e samples taken prior to this time were
for carrier impurity analysis. These five analyses were
discussed i n chap. 3, “Prenuclear Operation.”
ORNL-LR-DWG
3853
-
,
TIME
Fig. 4.4.
Passage of Enriched Slugs Through Reactor.
29
i t was then decided t o i n s t a l l a new transfer l i n e
and transfer pot. At t h i s time it was reported that
no transfer o f concentrate had been made t o the
pump from can A-7. The f i r s t 5.5 Ib of can A-7
was l o s t t o the experiment through the leaks and
in the plugged line.
At 2000 on November 1, the new transfer l i n e ond
tank were installed and checked out ready for u5e.
At 2006, fuel sample 7 was drawn for analysis,
and, a t 2300, the c r i t i c a l experiment was resumed.
When the f i r s t 5.5-lb slug was injected, periodic
pips were again observed on the fission chamber
recorder. These pips had a period of 47 sec which,
together with a pump speed o f 1080 rpm, yielded
a calculated fuel flow o f 46 gpm, which corresponded to the flow observed on the fuel flowmeter.
At 2315, while attempting the next 5.5-lb transfer,
the l i n e again froze. The l i n e was disconnected
and found t o be plugged a t the injection f i t t i n g
through the pump flange. It was therefore decided
t o increase the current t o the resistance-heated
f i t t i n g and t o add a separate vent l i n e from the
pump so that the chemists could continue to “blow
through” the transfer l i n e and out the new vent
l i n e without raising the system pressure.
By 1320 on November 1 the new vent l i n e had
been installed and the transfer system was again
in operating condition.
The transfer of the remaining 20 I b o f concentrate in four batches from
can A-7 took place with no difficulty, and the
transfer was completed by 1415. At 1617, fuel
sample 8 was taken for analysis, and, a t 1707,
the f i r s t batch from can A-8 was transferred i n t o
the system. The remaining 5 batches were transferred smoothly, and the transfer from can A-8 was
completed by 1804. At 2028, fuel sample 9 was
removed for analysis. The next three additions o f
fuel from cans A-9, A-10, and A-12 were accomplished with l i t t l e d i f f i c u l t y .
At 0830 on November 2, while transferring the
f i r s t 5-lb batch from can A-5, a leak occurred i n
the injection l i n e just below the floor level o f the
loading station (see Fig. 4.3). Examination o f the
injection l i n e showed the leak to have occurred at
a Swagelok fitting. Approximately 0.2 I b o f concentrate was l o s t from the experiment as a result of
the leak. Since the l i n e had clogged, it was necessary to i n s t a l l a new section o f line. During the
time the repairs were being made to the injection
system, the original schedule was changed, and a
subcritical measurement o f the tempeFature coefficient of reactivity was made. I n addition, samples
10 and 11 were taken for analysis.
30
S U B C R I T I C A L M E A S U R E M E N T OF T H E
REACTOR TEMPERATURE COEFFICIENT
Approximately 100 Ib o f U235 (180 I b o f Na,UF,)
had been added at the time the leak occurred on
November 2, a t 0831, and it was estimated that
t h i s was about 80% of the total fuel needed. Although it had been planned to make a preliminary
measurement of the temperature coefficient of
reactivity with about 90% o f the fuel added (see
II
Nuclear Operating Procedures,”
Appendix D),
the plans were revised and the measurement was
made at t h i s time.
At the start of the experiment the reactor mean
temperature was 1306OF. The fuel heat exchanger
barrier doors were raised ( a t 1003); two minutes
later the fuel helium blower was started and i t s
speed increased to 275 rpm. The resultant cooling,
with time, as traced by the reactor fuel mean
temperature recorder, i s shown in Fig. 4.5, with
the counting rate from fission chamber No. 2 superimposed, The data for both fission chambers are
The counting rate of the
given i n Table 4.2.
fission chambers monitors the neutron flux.
The counting rate was observed t o r i s e w i t h
decreasing temperature and thus indicated a negat i v e temperature coefficient o f reactivity. Counts
were taken for 40 sec during every minute. At
1010, when the reactor temperature reached 1250°F,
the helium blower was stopped, and soon thereafter
the reactor temperature began to r i s e again, with
a corresponding decrease i n counting rate.
ORNL-LR-DWG
i
L-
t
L
1
‘ t ”
L
1
L
P
5
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6396
(400
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W
5
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1350
W
z
5 1300
bc
k
w
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J
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3
a $250
P
a
W
[i
(200
i
-
k
c
Fig. 4.5. Subcritical Measurement of the Reactor
Temperature Coefficient. (Plot of fission chamber
No. 2 counting rate superimposed on fuel mean
temperature chart.)
.
c
k
i
T A B L E 4.2. SUBCRITICAL MEASUREMENT O F T H E
REACTOR TEMPERATURE COEFFICIENT
?
i
Reactor Mean
Ti me
Tempe rat ure
(OF)
Fission
Fission
Chamber
Chamber
No. 1
No. 2
(counts/sec)
(counts/sec)
1004
1306
486.4
723
1005
1304
492.8
73 6
1006
1302
512.0
755
1007
1290
531.2
780.8
1008
1275
537.6
793.6
1009
1260
537.6
793.6
1010
1252
544.0
800.0
1011
1244
531.2
787.2
1012
1240
531.2
774.4
1013
1246
518.4
768.0
1014
1252
518.4
768.0
1015
1258
512
761.6
1154
1275
502
736.0
1319
1284
490.9
725.0
1422
1290
484
713.0
J
d.
I
ature minimum did not occur simultaneously. Alag of about 2.5 min i n the response o f the reactor
temperature was observed.
The reason for t h i s
time lag i s not well understood, but i t may have
some connection with the fact that the thermocouples were outside the reactor and hence did
not “see”
the changes immediately. D e t a i l s o f
t h i s and later measurements of the fuel and reactor
temperature coefficients are given i n Appendix 0.
I
APPROACH T O CRITICALITY
A t 0130 on November 3, the injection system was
again i n operation, and the fuel additions during
fhe remainder o f the c r i t i c a l experiment occurred
without mishap.
The remaining portions o f can
A-7 and a l l of cans 8-22, 8-31, and B-20 were
added during t h i s time.
Throughout the course o f the experiment, the
progress toward c r i t i c a l i t y was observed on the
With every fuel injection the
neutron detectors.
counting rates of the two fission chambers and the
BF, counter were simultaneously clocked and
recorded. The approach to c r i t i c a l i t y could readily
be seen by plotting the reactivity, k, against the
concentration of the fuel i n the system. The reactivity was obtained from the relationship
k = l - -
1
,
M
:
”
The reactor temperature coefficient, as estimated
from the data presented in Fig. 4.5, was o f the
(Ak/K)PF.
The data were
order o f -5 x
adequate to show that the reactor would be easy
t o control; therefore the experiment was allowed
to proceed.. It w i l l subsequently be observed that
the temperature data recorded i n the data room did
not exhibit the f u l l temperature drop across the
reactor in any run where heat was being abstracted’
from the system.
Consequently where the abstracted heat i s one of the parameters o f the
experiment, as i n the above temperature coefficient
measurement, it would be expected that some
temperature correction would be i n order.
The
existing data for the case i n question are, however, too meager to justify correction, although
based on correlations (app. K) with data taken
during the low- and the high-power runs, the control room temperature indications were lower than
the actual temperature, and therefore the estimated
temperature coefficient given above i s too high.
It may also be noted from examination o f Fig.
4.5 that the maximum counting rate and the temper-
where M i s the subcritical multiplication, which
i s determined fiom the expression
M = -
N
,
NO
where N i s the counting rate after a fuel injection
and N O i s the i n i t i a l counting rate. As c r i t i c a l i t y
i s approached, 1/M approaches zero, and therefore
at criticality, k = 1.
Figure 4.6 shows a plot o f k vs U2,’ concentration, where k i s determined from three neutron
detectors, i.e., two fission chambers and a BF,
counter.
The reactivity, as observed on the two
fission chambers, showed a rapid increase during
the early stages of the experiment and then leveled
off as the c r i t i c a l condition was near. The BF,
counter, on the other hand, showed a more uniform
approach to criticality.
The differences i n the
responses o f the two types o f detectors are discussed i n Appendix E. Table 4.3 presents the
data from which Fig. 4.6 was drawn.
A condensed, running uranium inventory during
the c r i t i c a l experiment i s given i n Table 4.4. The
I
31
ORNL-LR-DWG 3852
10
EXP. E-!,
I
NOV. 3
I
t
09
t
/I
I
I
I
I
I
18
20
22
&
C
oa
-I
1
0 98
-
07
-k
c
>- 0.96
i
t
t
F
u
? 06
c
0
W
Q
(r
n
W
R
0.94
05
I
+
i
04
0.92
I
16
U235 CONCENTRATION IN FUEL ( l b / f t 3 )
03
0 2
0
4
8
12
16
24
20
32
28
36
40
U235 CONCENTRATION IN F U E L (lb/ft3)
Fig,
4.6.
Approach t o Criticality:
f i g u r e s o f c o l u m n s 4, 6, a n d 1 1 were s u p p l i e d by
the c h e m i ~ t s . ~
At 1545 on N o v e m b e r 3 u p o n t h e a d d i t i o n o f t h e
s e c o n d b a t c h of f u e l from c a n B-20, a s u s t a i n i n g
-
c h a i n reaction was attained
the reactor was
c r i t i c a l . F i g u r e 4.7 s h o w s a p h o t o of t h e c o n t r o l
room j u s t a s t h e c r i t i c a l c o n d i t i o n w a s reached.
A t 1547 t h e r e a c t o r w a s g i v e n o v e r t o t h e s e r v o
m e c h a n i s m a t an e s t i m a t e d p o w e r o f 1 watt, a n d
h r a b r i e f r a d i a t i o n s u r v e y of
during the next
t h e r e a c t o r a n d h e a t e x c h a n g e r p i t s w a s made (cf.,
A t 1604 t h e
chap. 5, “ L o w - P o w e r E x p e r i m e n t s ” ) .
r e a c t o r w a s s h u t down, and t h u s t h e i n i t i a l p h a s e
o f t h e ARE o p e r a t i o n program w a s c o m p l e t e d .
\
3 J . P. B l a k e l y , Uranium in the ARE, O R N L CF-551-43 (Jan. 7, 1955).
Reactivity vs
Fuel C o n c e n t r a t i o n .
T h e u r a n i u m i n v e n t o r y ( T a b l e 4.4) s h o w e d t h a t
t h e c r i t i c a l c o n c e n t r a t i o n o f U235 w a s 23.94 Ib/ft3,
w h i c h c o r r e s p o n d e d to a t o t a l w e i g h t of 133.8 Ib
o f U235 i n t h e f u e l system. A t t h i s t i m e t h e t o t a l
w e i g h t o f t h e f u e l w a s 1156 Ib, a n d i t s v o l u m e w a s
5.587 ft3. T h e per c e n t b y w e i g h t o f U235 w a s
c a l c u l a t e d t o be 11.57.
T h e c r i t i c a l m a s s of U235 in t h e r e a c t o r w a s
c a l c u l a t e d From t h e r a t i o o f volumes:
M
= M
1
v+.
- ,
vs
where
M~
= c r i t i c a l mass o f
M~ = m a s s of
v
u~~~in r e a c t o r
core,
u~~~i n system,
= c r i t i c a l v o l u m e o f reactor,
32
t
L
TABLE 4.3. APPROACH TO CRITICALITY: REACTIVITY FROM VARIOUS
NEUTRON DETECTORS vs F U E L CONCENTRATION
Fission Chamber No. 1
i
Log
Run
u235
~
Counting
Rate, N
Multiplication,
Reactivity,
k
1
'
il
1
2
3
4
5
6
7
8
9
10
11
12
9.46*
26.7
47.44
86.96
147.2
251.9
450.9
648.2
859.3
1489
4028
0
3.39
6.16
9.37
12.45
15.36
18.09
19.74
20.63
21.88
23.08
23.94
1
2.82
5.02
9.19
15.6
26.6
47.6
68.5
90.8
157.4
425.6
Fission Chamber No.
__
0
0.645
0.801
0.891
0.936
0.962
0.979
0.985
0.989
0.994
0.998
__~-____
Counting
R
~
~
(counts/sec)
16.28*
42.5
71.54
128.5
217.2
367.6
649.9
977.4
1317
2316
6730
Multiplication,
~
, M
2
- ___-
Reoctivity,
Counting
R~,~,
BF, Counter
Multiplication,
M
(counts/sec)
(NIN,,)
1
2.61
4.39
7.90
13.34
22.6
39.9
60.0
80.9
142.3
413.5
0
0.617
0.772
0.873
0.925
0.956
0.975
0.983
0.988
0.993
0.998
(N/N~)
4.49*
7.04
9.19
13.43
19.73
29.5
49.0
70.5
93.7
161.3
442.0
Reactivity,
k
1
1.57
2.05
2.99
4.39
6.57
10.8
15.7
20.9
35.9
98.5
0
0.361
0.510
0.667
0.780
0.848
0.908
0.936
0.952
0.972
0.990
Critical
*Initial counting rate, No.
T A B L E 4.4.
URANIUM INVENTORY DURING CRlTlCAL EXPERIMENT
F u e l Concentrate Added
Date
10/3
10/31
11/1
I
11/2
i
11/3
11/4
1
I
L
Time
1425
1625
1740
2013
1415
1635
1804
2040
Run
No.
1
2
Weight
Volume
lib)
(ft3)
30.00
3
4
25.37
30.54
5
30.44
0213
0610
0831
1625
6
30.07
29.19
5.50
7
Sa
2025
0256
0523
8b
9
12.62
9.93
0941
1245
10
11
14.18
13.92
1536
1545
12
1310
1315
10.07
F u e l Mixture
~~~l
~ 2 3 5
Concentrate
Removed
P l u s Carrier
(Ib)
(Ib)
(Ib)
0.1046
Uranium-235
~~~~l
weigh+
Removed
2.447
2.531
2203
0910
0915
S a m d e s Removed
for A n a l y s i s
Total
Total
Density Weight
Volume
(Ib/ft3)
(ft3)
927.3
4.820
957.3
954.8
4.925
4.912
192.4
194.4
Weight
Added
i n System
(Ib)
(Ib)
16.687
952.3
4.900
14.105
0.0716
16.980
0.1261
1008
1005
198.3
2.670
4.988
4.976
5.082
5.069
196.0
2.286
977.7
977.5
47.613
47.487
16.940
16.695
16.230
3.058
0.2337
5.175
5.280
5.382
5.401
5.388
200.2
201.9
203.5
203.8
2.562
1036
1066
1095
1101
1098
64.427
81.122
97.352
100.410
100.176
2.716
0.2478
1095
1 io8
1118
1132
1146
1156
5.375
5.419
204.5
205.0
7.037
5.536
99.928
106.965
112.501
7.908
7.751
120.409
128.160
5.587
205.7
206.4
206.9
5.600
133.760
1153
1151
5.574
5.561
1148
1147
5.549
5.545
0.1065
0.1062
0.1049
0.1018
0.0192
0.0440
0.0346
0.0495
0.0486
0.0351
5.454
5.503
5.552
wt%
0
16.687
0.0427
0.0441
0.0885
Concentration
Ib/@
16.644
16.600
30.705
30.633
3.388
1.743
6.156
3.141
9.368
4.723
12.45
15.36
18.09
18.59
19.74
20.63
21.88
23.08
23.94
6.220
7.610
8.890
9.123
9.654
10.06
10.64
11.18
11.57
C r i t i c a l i t y reached
2.679
2.718
2.482
0.732
0.3100
0.3145
0.2872
0.0847
133.450
133.135
132.848
132.763
4
33
34
L
h
i
1
c
?
4
t
i
t
t
L
t
1.
h
J&
k
1
8,
4
i
I
c
c
L
b
P
e
ys
*
i
I
i
i
b
c
1.
;
i
i
P
t
Vs
= system volume.
Therefore
Mr =
133.8 I b
(i::i7f::$
=
32.8 Ib
.
information that i s available and, because o f the
general interest therein, typical axial and radial
flux and power distribution curves are given i n
Appendix G.
A N A L Y S E S O F F U E L SAMPLES
”
*
The calculated, cold, clean c r i t i c a l mass o f the
reactor, as obtained from the subsequent rod c a l i brations, was 32.75 Ib (cf., app. F).
The reactor was not instrumented t o permit the
measurement of the flux or power distributions
through the reactor, but measurements of these
distributions were made on a c r i t i c a l mockup of
the reactor at the ORNL C r i t i c a l Experiment Fac i l i t y about two years before the operation of the
ARE.4 These measurements represent the best
4D. C a l l i h a n and D. Scott, Preliminary Critical A s s e m b l y for the Aircraft Reactor Experiment, ORNL-1634
( O c t . 28, 1953).
TABLE 4.5.
In addition t o the fuel samples taken, as noted,
during the c r i t i c a l experiment, four more samples
were removed on November 4 after i n i t i a l c r i t i c a l i t y
was reached. A l i s t of a l l samples taken and the
results of the chemical analyses are presented i n
Table 4.5. Besides showing the analyses o f the
percentages of U and U235 by weight i n the fuel
system and a comparison with the U235 (wt %)
content obtained from the c r i t i c a l i t y data, the
table also l i s t s the results of the analyses for the
impurities and corrosion products Fe, Cr, and Ni
t o provide information on the purity of the fuel and
the corrosion rate of the fuel system.
CHEMICAL ANALYSES O F F U E L SAMPLES
u~~~from
Sample
Date
Impurities and Corrosion Products (ppm)
Time
Total
Uranium
No.
Cr
Fe
Ni
( w t %)
~ 2 3 5
Chemical
from C r i t i c a l
Analysis
Experiment D a t a
( w t %)
( w t %)
10/25
1650
1
90
200
50
10/26
1414
2
81
(5
40
1911
3
81
<5
6
10/27
1919
4
90
18
12
10/29
0935
5
102
30
20
10/30
1740
6
100
20
10
2.47 f 0.01
2.31
1.74
2013
7
150
15
10
1.84
k 0.04
1.72
1.74
1635
8
190
15
15
3.45 f 0.01
3.22
3J4
2040
9
200
10
17
5.43 Itr 0.01
5.07
4.72
1625
10
210
k 0.08
8.95
9.1 2
2025
11
205
11/1
11/2
11/4
11/5
0910
20
20
9.54 f 0.08
8.91
9.12
11.57
310
12.11
f 0.10
11.32
11.41
11.57
0915
13
300
12.21 f 0.12
1310
14
320
12.27 f 0.08
11.46
11.57
1315
15
310
12.24 5 0.12
11.43
11.57
1100
16
372
12.54 f 0.07
11.72
11.72
12.57 f 0.07
11.75
1 1.79
17
11/6
111’7
12
9.58
5
<5
0535
18
420
12.59 5 0.12
11.77
11.79
0423
19
445
13.59 f 0.08
12.70
12.38
35
The f i r s t f i v e samples were removed from the
system prior t o enrichment, and samples 16 through
19 were taken after the c r i t i c a l experiment and
during the low power runs a t the time o f the c a l i bration o f the regulating rods against fuel addition.
For the most part, the U235 content as given by
chemical analysis agreed t o within a few per cent
with that obtained from the running inventory.
When the sample for analysis was withdrawn from
the system too soon after a fuel addition, there
was a tendency for the analysis t o be low, undoubtedly, because o f inadequate mixing of the
additive with the bulk of the fuel. The chemical
analysis o f sample 1, however, was obviously i n
error, whereas that of sample 9 gave a uranium
content that was about 7% higher than the inventory
showed. The other sample analysis figures were
within about 2% of the percentages given by the
inventory. The analyses o f samples 12, 13, and
14, which were taken after the c r i t i c a l experiment,
agreed t o within about 1% with the inventory calcu1 ation.
The increase in the buildup o f chromium i n the
fuel system with time i s shown in Fig. 4.8 t o give
an indication of the corrosion rate of the system.
The i n i t i a l corrosion rate was quite small, but
after enriched fuel had been added t o the system,
the analyses showed a chromium content increase
o f abaut 50 ppm/day.
C A L I B R A T I O N O F T H E SHIM RODS
A s outlined i n the “Nuclear Operating Procedures,” Appendix D, during the f i r s t fuel additions the shim rods were withdrawn a l l the way
out o f the reactor t o obtain the multiplication.
After about 100 I b o f U235 had been added to the
system, the procedure was altered i n order to
obtain a shim rod calibration.
After each fuel
addition, a l l three rods were simultaneously withdrawn t o positions of 20, 25, 30, and 35 in. out;
total movement was 36 in. The counting rate o f
each of the neutron detectors was recorded for
each rod position. Figure 4.9 shows the reciprocal
multiplication and the reactivity as a function o f
U235 content of the fuel system for the various
rod positions. The data used for plotting Fig. 4.9
were obtained from the BF, counter and are presented in Table 4.6. A cross-plot o f reactivity v s
shim rod position determined the rod calibration
details, which are given i n Appendix J. The rods
36
ORNL-LR-DWG
6397
500
..
-
424
448
(4 7
400
I
IW
44 6
5
300
2
200
51
34
47
100
0
;
e
t
E
z
W
0
8
0
2
B
N3
0
Fig. 4.8. Increase in Chromium Concentration in
F u e l a s a Function of Time.
were found t o be nonlinear, with (Ak/k)/in. and
total Ak/k varying i n the manner shown i n Append i x D. From the calibration it was found that each
rod was worth a total of about 5.8% Ak/k.
A s a check on the general shape o f the reactivity
curves of the shim rods, a series of counts were
taken on the BF, counter and the fission chambers
for various rod positions for two different uranium
concentrations. These data are also discussed i n
Appendix J.
M E A S U R E M E N T OF T H E
REACTIVITY-MASS R A T I O
Prior t o the ARE operation it had been estimated
that when the c r i t i c a l mass (assumed t o be 30 16)
was reached, the value of the r a t i o (Ak/k)/(AM/M)
should be 0.232. T h i s value was obtained from a
calculation o f the r a t i o for various amounts of
U2,’, as shown i n Fig. D.l of Appendix D. From
the data taken during the c r i t i c a l experiment, it
was possible to establish an experimental curve
for the r a t i o and t o verify the value given above
for the ratio a t the c r i t i c a l mass.
In order t o find (Ak/k)/(AM/M)
experimentally,
use was made o f the curve o f Fig. 4.9 which gives
k i n terms o f the U235 content o f the fuel in the
system for various shim rod positions; i.e., for any
point,
Ak/k
-=
AM/M
&)(;
= R
which i s the reactivity-mass ratio.
,
The value
T A B L E 4.6.
REACTIVITY vs
U235 CONTENT O F FUEL SYSTEM FOR VARIOUS SHIM ROD POSITIONS
Experiment
"235
Shim Rod
E-1
Run No.
in System
Positions*
BF3 Counter
(Ib)
(in. out)
(countdsec)
8 A
100.41
20
25
30
35
32.0
38.4
46.9
54.3
0.140
0.1 17
0.0956
0.082
0.860
0.883
0.904
0.918
8 8
103.13
20
25
30
35
34.1
40.5
51.2
57.8
0.131
0.111
0.0877
0.0774
0.869
0.889
0.912
0.923
8 C
106.18
20
25
30
35
36.3
46.9
57.8
66.1
0.123
0.0954
0.0774
0.0676
0.877
0.905
0.923
0.932
8D
106.97
20
25
30
35
36.3
46.9
59.7
68.3
0.123
0.0954
0.0749
0.0655
0.877
0.905
0.925
0.935
9A
110.14
20
25
30
35
38.4
51.2
68.3
83.2
0.1 17
0.0874
0.0655
0.0538
0.883
0.9 13
0.934
0.946
9 6
112.50
20
25
30
35
40.5
55.5
76.8
93.9
0.1 11
0.0808
0.0582
0.0479
0,889
0.919
0,942
0.952
10 A
115.58
20
25
0.0998
0.0749
0.0538
0.0403
0.900
0.925
35
44.8
59.7
83.2
110.9
30
d
.
Counting Rate,
N o/N
k =1
- (NO/N)
0.946
0,960
10 B
118.63
20
25
30
35
42.7
68.3
93.3
136.5
0.105
0,0655
0.0479
0.0327
0.895
0.934
0.952
0.967
10 c
120.41
20
25
30
35
51.2
68.3
110.9
162.1
0.0873
0.0655
0,0403
0.0276
0,913
0,934
0.960
0.972
11 A
123.43
20
25
30
35
51.2
85.3
128.0
213.3
0.0873
0,0524
0.0349
0.0209
0.913
0.948
0.965
0.979
*Position of all three shim rods.
37
T A B L E 4.6 (continued)
Experiment
"235
Shim Rod
E-1
Run No.
i n System
Positions*
(Ib)
(in. out)
11 B
126.48
20
25
30
35
59.7
93.9
162.1
307.2
0.0749
0.0476
0.0275
0.0 145
0.925
0.952
0.973
0.985
11 c
128.16
20
25
30
35
59.7
93.9
196.3
443.7
0.0749
0.0476
0.0228
0.0101
0.925
0.952
0.977
0.990
12 A
131.08
20
25
30
35
59.8
110.9
256.0
955.7
0.0748
0.0403
0.0 175
0.00468
0.925
0.960
0.983
0.995
12 B
133.76
20
25
30
68.3
136.5
375.5
0.0655
0.0327
0.0119
0.935
0.967
0.988
35
Critical
Counting Rate,
BFg
k
N O/N
Counter
=1
- (N~/N)
(count s/sec)
1
b
* P o s i t i o n of a l l three shim rods.
88
-
8C 80
98
ORNL-LR-DWG
RUN NUMBER
4OA
106 1OC
9B
llA
118 11c
12A
6398
428
0 16
1
0140
0 98
012,,
0.96
\
>O
G
b-
;010
-t
0.94 >-
J
a
0.92
3
6
a
W
I
2
0
c
->
I1I
/
0.90
006"
0
n
0.04
"4
0.88
W
n
0.86
0 0 2 '1-
0402
104
406
108
110
112
114
u235
416
tl8
420
124
126
128
430
132
134
0.84
136
IN FUEL SYSTEM ( I b )
Fig. 4.9. R e a c t i v i t y and Reciprocal Multiplication vs
Rod Positions.
4
422
U235 Content of F u e l System for Various Shim
.
f
c
6399
ORNL-LR-DWG
0.35
curves the slopes approximated straight lines.
From the average values of M and k over these
increments of the curves and the measured slopes,
the r a t i o R was determined for each increment.
Figure 4.10 shows a plot o f R vs the U235 content
of the fuel i n the system.
0.30
\
(A.k/AM) i s the slope of the k vs M curves shown
For various small sections o f the
i n Fig. 4.9.
2
a
1
s
- 0.25
4
c
The value of R for the experimental c r i t i c a l mass
was calculated to be
(133.8 Ib)
Ak/k
0.236
-=
AM/M
0.20
25
26
27
28
29
30
3(
32
33
34
MASS U235 IN REACTOR ( I b )
4.10.
Fig.
R e a c t i v i t y M a s t R a t i o as a Function
of the U235 Content of the F u e l in the Reactor,
T A B L E 4.7.
for shim rod positions of 25, 30, and 35 in. Since
t h i s value (0.236) was i n excellent agreement with
the calculated value (0.232), it was used throughout the experiment.
The data are tabulated i n
Table 4.7.
REACTIVITY-MASS RATIO AS A FUNCTION OF T H E AMOUNT OF U235 IN
T H E REACTOR FOR VARIOUS SHIM ROD POSITIONS
Reactivity-Mass
Mass
Shim Rod
Range*
Pasition**
M , Mass of
U235 in Reactor
(1 b)
(in.)
(Ib)
100 to 106
25
30
35
25.24
25.24
25.24
A
M
Reactivity,
(Ib)
k
Ratio,
Ak
Ak/k
hw/M
1.47
1.47
1.47
0.891
0.9 13
0.924
0.0172
0.0 175
0.0 174
0.331
0.329
0.322
Average
114
to
120
25
28.66
30
35
28.66
28.66
1.47
1.47
1.47
0.929
0.951
0.963
0.0137
0.0136
0.0 135
0.287
0.279
0.273
Average
128 to 134
25
30
35
32.10
32.10
32.10
1.44
1.49
1.40
0.961
0.982
0.995
0.01 10
0.280
0.257
0.219
0,231
0.0100
0.0 100
Average
* T h e s e mass ranges refer to the curves of Fig.
0.327
0,236
4.9 o n which these calculations are based.
* * P o s i t i o n of a l l three rods.
39
5. LOW-POWER EXPERIMENTS
The reactor was operated, after i n i t i a l c r i t i c a l i t y
was attained, for about 20 min at a nominal power
o f 1 w and then shut down. The next morning a
series of low-power experiments was started. T h i s
series o f experiments lasted from the morning o f
November 4 u n t i l the morning of November 8. T h e
chronology of t h i s low-power operation i s given i n
Fig. 5.1.
Two o f the early experiments, L-1 and L-4, were
devoted to a determination o f the reactor power
from measurements o f the fuel activation.
These
runs were each o f 1 hr duration with the reactor a t
a nominal power of 1 and 10 w, respectively. After
each of these runs, fuel samples were taken and the
activity counts were made from which the reactor
power was determined. The method was inaccurate,
as subsequently evidenced by power calibrations
from the extracted power, but did indicate that the
nominal power estimates were low.
Radiation surveys of the entire experimental
system were made during both the nominal 1- and
10-w runs except that a survey o f the reactor p i t
was not attainable during the 10-w run because
the p i t was sealed at that time. Most of the equipment and many of the instruments were, however,
located i n the heat exchanger p i t s to which access
was maintained throughout the low-power operation.
Most of the time o f the low-power operation was
devoted to calibration o f the regulating and shim
rods.
The two essentially independent methods
used to calibrate the regulating rods were c a l i bration against fuel addition and calibration against
reactor periods by using the inhour equation (app. I).
The shim rods were then calibrated against the
regulating rod.
A l s o as an integral part o f the low-power operation the earlier measurement o f the temperature
coefficient was checked, and the fuel system performance characteristics were determined. Finally,
i n preparation for the high-power experiments, the
ion chambers, which had been positioned close to
the reactor during the c r i t i c a l and low-power experiments for greater sensitivity, were withdrawn
so that they would register sensibly a t the higher
powers.
POWER D E T E R M I N A T I O N F R O M
FUEL ACTIVATION
Reactor power measurements are usually made
by exposing gold or indium f o i l s to the neutron
40
flux i n a reactor soon after i n i t i a l c r i t i c a l i t y i s
attained. In the ARE i t was simpler to draw o f f a
sample o f the uranium-bearing fuel after operation
and measure i t s activity with an ion chamber. Comparison with a similar sample o f known activity
gave a determination o f the flux level and, hence,
the power level.
The procedure i s described i n
Appendix ti.
In run L-1, the reactor was operated for 1 hr a t
an estimated power level of 1 w, and a sample was
drawn o f f for measurement o f the activity.
The
specific activity was too low for a reliable determination and, accordingly, the experiment was
repeated i n run L-4 a t a nominal 10-w power level.
It was subsequently found from the heat balance
a t high power that the actual power was 2 7 w during
run L-4. However, the fuel sample a c t i v i t y was
only on the order o f one-half t o two-thirds the
a c t i v i t y to be expected from operation a t 27 w.
Apparently a considerable amount o f the gaseous
fission products was being given o f f from the fuel.
A comparison o f the data obtained from fuel a c t i vation with those obtained from heat balances i s
given i n Appendix N.
RADIATION SURVEYS
The primary purpose of experiments L-1 and L-4
was tocalibrate the estimated reactor power against
the power determined from the radioactivity o f fuel
samples, but, a t the same time, radiation dose
levels were measured a t various locations around
the reactor and the tank and heat exchanger pits.
These data w i l l subsequently be o f interest i n
evaluating the radiation damage t o various components of the system.
Most of the radiation measurements were made
w i t h a “Cutie-Pie” (a gamma-sensitive ionization
chamber), but a GM survey meter, a methane proportional counter, an electroscope, and a boroncoated electroscope were used for some measurements, i n particular, those made close to the
reactor. While the GM survey meter only provided
a check on the gamma dose as measured by the
“Cutie-Pie,”
the other instruments provided measurements of the fast- and thermal-neutron doses.
The proportional counter measured the fast-neutron
dose, and the thermal-neutron flux was calculated
from the difference of the two electroscope readings.
i
P
0
0
N
,
w
r
-0
X
m
z
z
C
r
rn
C
n
m
r
7
X
0
0
-
m
N
w
0
0
EXP. L-9
0
0
r
h)
0
-
TIME OF DAY
UI
0
-
I
3
0
m
z
0
0
-
Z
9
V
EXP. L-8
1
P
W
ul
-
xi
5m
0
z
<
m
0
w
W
w
P
;D
m
-U
X
m
;o
rn
z
0
0
m
w
r
7
I
-i
-rl
0
t
i
i
L
L
k
1
I
t
I
!
i
1
b
9
t
9
4
The radiation dose data from the two runs during
which the dose i n the p i t s was measured are presented i n Tables 5.1 and 5.2. Table 5.1 gives the
gamma and the fast- and thermal-neutron doses i n
the reactor p i t during the 2.7-w run; Table 5.2 gives
the gamma doses a t various stations i n the heat
exchanger p i t during both the 2.7- and 27-w runs.
Within the accuracy o f the measurements, t l y r e i s
excellent agreement between the doses per watt
from the two runs.
R E G U L A T I N G ROD CALIBRATION FROM
F U E L ADDITIONS
i
fuel was injected from the penguin. Upon completion o f the injection the reactor was again brought
c r i t i c a l to 1-watt power with the regulating rod.
The positions of the rods were again recorded.
The worth o f the rod was then obtained from the
relation
Ak
~
k
AM
=
0.236 M
and the known increment o f fuel added t o the system; the proportionality factor, 0.236, was obtained
experiment0 l l y
during the c r i t i c a l experiment
(cf., chap. 4).
<I
Twelve small cans, or
penguins” (so-called
because o f their shape), had been prepared for
adding small amounts o f fuel to the system i n order
t o calibrate the regulating rod (exp. L-2) after
i n i t i a l c r i t i c a l i t y had been reached. Each penguin
contained from 0.2 t o 0.5 I b of fuel concentrate
(0.1 to 0.3 Ib o f U235). The contents o f the penguins were injected directly into the pump rather
than through the intermediate transfer pot that was
used i n the c r i t i c a l experiments. Table 5.3 l i s t s
the penguins i n the chronological order i n which
they were used, together with the amount o f uranium
injected from each can.
The method used to calibrate the regulating rod
was t o note the difference between the position of
the rod a t c r i t i c a l i t y before and after each injection
of fuel, with the shim rods held constant. Before
each injection the reactor was brought from subc r i t i c a l to a nominal power o f 1 w, and the shim
and regulating rod positions were recorded. The
reactor was then brought subcritical on the regul a t i n g rod, with the shim rods i n position, and the
4
:
TABLE 5.1.
Position
Space cooler No.
s4
1
T o p center of reactor
(1 in.
i
While the rod was being changed (a delay o f
about 5 hr), the fuel system characteristics of
pressure head vs flow were obtained.
These
measurements are described i n a following section
o f t h i s chapter.
RADIATION LEVELS IN REACTOR PIT DURING 2.7-w RUN
Gamma
Fast-Neutron
Thermal-Neutron
Total
Dose Rate
Dose Rate
Dose Rate
Dose Rate*
(mr/hr per w )
(mrep/hr per w )
(mrep/hr per w )
(mr/hr)
48
67
280
280
35 (est.)
3200
440
180
22 (est.)
2350
12
10
8
760
above thermal shield)
Side of reactor a t mid-plane
,
The f i r s t run of experiment L-2 started a t 1651
on November 4.
After two injections had been
completed, i t was apparent that the worth o f the
12 in. o f vertical movement o f the regulating rod i n
terms o f A k / k was about 0.24%. It was desirable,
for reasons of safety and also for convenience i n
conducting hi gh-power transient experiments, to
have a rod worth about one dollar o f reactivity
(which for the ARE was 0.4% rather than 0.76% as
i n stationary-fuel reactors, because of delayedneutron l o s s in the circulating fuel). A number o f
spare rods of varying weights had been made up t o
take care o f such a contingency, and one was
selected and installed which had more nearly the
desired weight. T h e original rod weighed 19.2 g/cm
of length, and the new rod weighed 36 g/cm.
( a t surface of thermal
shield)
Manhole into p i t s
i
2
* T h e t o t a l dose rate was obtained from the weighted sum of the preceding columns by using an RBE of
neutrons and 5 for thermal neutrons.
130
10 for f a s t
I
43
TABLE 5.2.
GAMMA-RAY DOSE IN TANK AND H E A T EXCHANGER PITS
Doses per Watt
During 2.7-w
Run
During
27-w Run
(mr/hr)
(mr/hr)
0.3
0.1
0.2
0.1
0.07
0.3
0.15
0.15
0.07
0.07
0.07
0.07
-
Locations Surveyed i n Dump Tank P i t
Motor for space cooler No,
Vent valve
t
Doses per Watt
3
U-113 for tank No. 6
Motor for stack vent on hot fuel dump tank
U-112 for hot f u e l dump tank
U-100
Motor for space cooler No. 4
Air-operated valve for tank No. 5
Vent Valve
Helium i n l e t valve
.
B
t
Locations Surveyed i n Fuel-to-He1 ium Heat Exchanger P i t
Operator's p o s i t i o n at fuel sampler
Pressure transmitter PXT-6
Top plate of main f u e l pump
Top bearing of pump motor
V-belt o t pump
V-belt at pump motor
Line
120 under valve U-3
2 of l i n e 303
Bend No.
Motor for spoce cooler No.
6
West side o f heat exchanger No.
1
Bottom o f fuel flowmeter
Line
112 between valves U-63 and U-1
L u b r i c a t i o n pump for pump
Top of fuel storage tank
Helium analyzer dryers
6
22
150
45
7
7
110
105
10
135
95
22 0
65
4
0.3
1
West side o f heat exchanger No. 2
E a s t side o f heat exchanger No.
Sheet metal can around pump
4
120
80
6
4
12
45
120
105
200
90
1
230
200
185
Locarions Surveyed i n Sodium-to-Helium Heat Exchanger P i t s
Top o f stondby sodium pump
V-belt above standby pump
Line
310 under v a i v e B-141
7
Motor for space cooler No.
Top o f main sodium pump
V-belt above main pump
309 a t valve U-21
L i n e 313 a t bend No. 5
Line
Motor for space cooler No.
8
North side o f front heat exchanger
Kinney pump No.
1
Magnetic c l u t c h on
50-hp motor
North side of back heat exchanger
North end o f rod-coolant blower No.
EM flowmeter on l i n e 305
Standby pump l u b r i c a t i o n system
Main pump l u b r i c a t i o n system
44
1
1.5
1.5
13
4
1.5
0.7
2.2
1.5
1.5
15
4
11
1
6
3.7
1.4
1.7
15
0.7
0.8
1.8
1.5
16
4
13
6
7
5
9
3
I
c
E
k
TABLE 5.3.
F U E L CONCENTRATE BATCHES ADDED DURING LOW-BOWER EXPERIMENTS
Penguin
Date
11/4
11/5
Concentrate Added
1
U235 Added
Uranium Added
Time
t
Ib
wt %
9
9
Ib
227
0.5004
59.671
135
126
0.2778
C-19
722
1.5917
59.671
43 1
4 03
0.8885
0354
C-14
247
0.5445
59.671
147
137
0.3 020
0439
C-16
285
0.6283
59.671
170
159
0.3505
0529
C-13
131
0.2888
59.671
78.2
73.0
0.1609
0619
c-5
123
0.2712
59.671
73.3
68.5
0.1 51 0
2045
C-15
92
0.2028
59.671
55
51.4
0.1 143
2204
C-18
87
0.1918
59.671
52
48.6
0.1071
2258
C-17
457
1.0075
59.671
273
2345
c-12
84
0.1852
59.671
50
46.7
0.1 030
01 00
c-10
84
0.1852
59.671
50
46.7
0.1 030
No.
9
1710
c-11
2009
I
2 55
I
I
0.5622
L
11/6
11/7
023 1
120*
21,681
47.798
27.557
558 1
5975
t
t
12.304
1
k
*Container No. 120 was not a penguin but a specially prepared batch of concentrate which provided the excess
uranium required for the experiments and to compensate tor burnup.
8
A t 0142 on November 5 the installation of the
Y
second rod was completed and the experiment was
resumed.
Four more injections o f fuel were made
and the rod movement was noted. At 0630, during
the sixth fuel injection, an unexpectedly high
burst of gamma-ray activity (55 mr/hr) was observed on a “Cutie-Pie” monitoring the fuel addit i o n a t the injection station over the pump. Further
fuel addition was stopped until the cause o f the
high gamma-ray reading could be ascertained.
While investigation was under way the level o f the
pump was trimmed to avoid any possible hazard
from the uranium held in the pump. About 71 Ib o f
fuel was removed from the system.
Later i n the
day the cause o f the high gamma-ray a c t i v i t y was
determined t o be a vent l i n e passing close t o the
position o f the “Cutie-Pie.”
Apparently, the
“Cutie-Pie”
had been held close enough t o the
vent l i n e t o read the fission gases (evolved during
operation) as they were discharged through the
vent line.
?
A t 1947 the experiment was resumed and f i v e
more injections were made uneventfully. A photograph o f the control room taken during t h i s time i s
presented in Fig. 5.2; it shows two members o f the
ARE operating group examining the fission chamber
recorders after a fuel injection for rod calibration;
other members of the evening crew are shown i n
their nominal operating stations.
A n inventory o f the uranium added during t h i s
The final large
experiment i s given i n Table 5.4.
amount (47.8 Ib) of concentrate containing 12.30 lb
U 2 3 5 that was added after the experiment is also
shown i n the table, as well as a record of the
samples and withdrawals.
The f i n a l amount of
U235 in the system was 138.55 Ib, w i t h a system
volume o f 5.33 ft3.
The data obtained for calibration of the regulating
rod as a function o f fuel addition are given i n
Table 5.5, which l i s t s the values of M, A M , AM/M,
the recorded movement o f the rod during each fuel
addition, the average position o f the rod for each
fuel addition, and the calculated value o f (Ak/k)/in.
The results of the calibration are shown in Fig.
5.3, where (Ak/k)/in. i s plotted as a function o f
rod position. The movement o f the rod for each
point i s shown as a horizontal l i n e through the
point. The values of (Ak/K)/in. for both rods were
used, the values o f reactivity for rod No. 1 being
corrected by the ratios o f the weights o f the two
rods. From these data it i s evident that there i s no
good reason for presuming that the value o f
(Ak/k)/in.
over the whole length of rod i s not
constant. Therefore, the average value taken over
the f i r s t ten runs gives
(Ak/k)/in.
=
0.033%/in.
45
i
i
i
46
s
t
u)
P
L
.-E
X
e
W
0
t
.-0
L
-
P
u"
-8
o?
m
a
L
l=
.n
0
u)
t
L
fm
e
.-
E
.-e
X
..
W
0
al
e
al
0
m
E
0
0
-o?
E
L
+
0
0
u
-
.-n.
$
2
.-m
U
k
I
T A B L E 5.4.
URANIUM INVENTORY DURING LOW-POWER EXPERIMENTS
~
~
Samoles Removed
F u e l Concentrate Added
Date
Time
Run
No,
Weight
Ob)
Volume
(ft3)
for A n a l y s i s
Fuel
Removed
(Ib)
11/4
1315
1710
2009
11/5 0354
0439
0529
0619
1015
1020
1100
2045
2204
2258
2345
11/6 0100
0530
0535
0231
0423
4
f
i
4
1
1
d
lln
4
J4
i
.
0
1
2
3
4
5
6
7
8
9
10
11
12
0.5004
1.5917
0.5445
0.6283
0.2888
0.2712
0.2028
0.1918
1.0075
0.1852
0.1852
47.798**
U235
Removed
(Ib)
6,269
2.064
0.291
2.686
2.839
0.317
0.335
2.918
0.361
Tota I
Concentrate
Volume
(Ib)
(ft3)
Plus Carrier
0.0017
0.0056
0.0019
0.0012
0.0010
0.0009
53.486*
17.523*
2.5807
Uran ium-235
Fuel Mixture
T~~~~weight
0.0007
0.0004
0.0020
0.0004
0.0004
0.1667
1147.45
1147.95
1149.54
1150.08
1150.71
1151.00
1151.27
1097.78
1080.26
1077.68
1077.88
1078.07
1079.08
1079.27
1079.45
1076.76
1073.92
1121.72
1118.80
5.5454
5.5471
5.5527
5.5546
5.5558
5.5568
5.5577
5.2944
5.2148
5.2023
5.2030
5.2034
5.2054
5.2058
5.2062
5.1932
5.1795
5.3462
5.3323
Weight
Densip
Added
(Ib/ft)
(Ib)
Weight
in System
concentration
Ib,ft3
W t
%
(Ib)
206.9
206.9
0.2778
207.0 0.8885
207.05 0,3020
207.1
0.3505
207.1
0.1609
207.1
0.1510
207.1
207.1
207.1
207.2 0.1143
207.2
0.1071
207.3 0.5622
207.3 0.1030
207.3 0.1030
207.3
207.3
209.8 12.304
209.8
132.763
133.041
133.930
134.232
134.583
134.744
134.895
128.626
126.562
126.271
126.3d5
126.492
127.054
127.157
127.260
126.943
126.608
138.912
138.551
23.94 11.57
23.98 11.59
24.12 11.65
24.17 1 1.67
24.22 11.70
24.25 11.71
24.27 11.72
24.27 11.72
24.27 11.72
24.27
11.72
24.29 11.73
24.31 11.73
24.41 11.77
24.43 11.78
24.44 11.79
24.44 11.79
24.44 11.79
25.98 12.38
25.98 12.38
* T w o large batches o f f l u o r i d e mixture were removed from the system i n order t a trim t h e l i q u i d l e v e l in t h e pump.
**Of t h i s amount, only 15.419 Ib was Na,UF6, the remainder being some of t h e f l u o r i d e mixture t h a t was removed from t h e s y s t e m when the
pump l e v e l was trimmed.
0 05
0 04
4
-._
,
-z
c
-
0 03
?
0
+
a
W
[r
0.02
0.04
0
2
3
4
5
6
7
REGULATING
Fig.
8
IO
9
ROD POSITION
(in I
5.3. Calibration of Regulating Rod from F u e l Addition.
12
14
T A B L E 5.5.
REGULATING ROD REACTIVITY CALIBRATION FROM F U E L ADDITION
(Exp.
L-2)
Movement o f
M,
Run
No.
Weight of
UZ3’
AM,
Weight of
Average
U235
AWM
Ak/k
(%I
Regulating Rod
Regulating Rod
Position
Reactivity,
During F u e l
(Ak/k)/in.
Add i t ion
(%/in.)
i n System
Added
(Ib)
(Ib)
1
133.041
0.2778
0.00209
0.0493
11.80
2.4
2
133.930
0.8885
0.00663
0.156
9.35
7.3
0.0400 *
3
134.232
0.3020
0.00225
0.0531
10.70
1.6
0.0335
4
134.583
0.3505
0.00260
0.0614
8.80
1.9
0.0323
5
134.744
0.1609
o.00119
0.0281
6.80
0.85
0.0330
6
134.895
0.1510
0.001 12
0.0264
6.00
1 .o
0.0264
7
126.385
0.1143
0.000904
0.0213
6.35
0.7
0.0304
a
126.492
0.1070
o.000847
0.0200
6.00
0.6
0.0333
9
127.054
0.5622
0.00442
0.104
5.10
3.6
0.0289
10
127.157
0.1030
0.000810
0.0191
12.25
0.5
0.0383
11
127.260
0.1030
0.000809
0.0191
2.95
0.67**
0.0305**
(in.)
(in.)
.
Average reactivity, weighted according to rod movement
*Corrected by the ratio of the weight of the new rod to the weight of the original rod:
o.0308*
0.0331
36/19.2 = 1.88.
**Average of three readings.
The value from run 1 1 was not included i n t h i s
average because three different values of rod f i n a l
position were recorded i n different places for t h i s
run.
R E G U L A T I N G ROD C A L I B R A T I O N FROM
REACTOR PERIODS
The calibration of the regulating rod by use of
reactor periods (exp. L-5) u t i l i z e d the inhour equat i o n i n a form i n which r e a c t i v i t y is given i n terms
of the reactor period for a given rate of f u e l f l o w .
Figure 5.4 shows a plot of the reactivity ( A k / k ) as
a function o f reactor periods for 0- and 48-gpm
fuel flow, as calculated from the inhour relation.
Details of the calculational method used are presented i n Appendix I.
In the experimental procedure followed, the regul a t i n g rod was suddenly withdrawn a known distance, while the shim rods were i n a set position
and the reactor was operating a t a nominal power
The reactor was then allowed t o
o f about 1 w.
r i s e i n power on a constant period to a peak o f
50 to 100 w, a t which time it was manually
48
scrammed. From the induced period and the known
flow, the excess r e a c t i v i t y introduced during the
run was obtained from the inhour formula.
This
number was then divided by the travel of the regul a t i n g rod t o find the value o f (Ak/k)/in.
for the
A total of eight reactor excursions o f t h i s
run.
nature were made, s i x at 48-gpm flow, and two a t
zero flow.
Some of the reactor excursions during t h i s experiment are shown i n Fig. 5.5, as recorded by the
The straight l i n e s that indicate
l o g N recorder.
the slope of the power increase r e f l e c t periods on
the order of 20 to 25 sec. Figure 5.6 presents some
o f the period traces recorded on the period recorder
during the experiment. The i n i t i a l high peaks for
each rod movement are due t o the transient condition,
h e interpretation 04 which i s given i n
Appendix S. I n plotting the period for a particular
rod motion, the average o f the periods obtained
from the log N and the period recorders was used.
A p l o t of (Ak/k)/in. as a function of rod p o s i t i o n
as obtained from this experiment i s shown i n
Fig. 5.7. The horizontal lines on each s i d e o f the
i
ORNL-LR-DWG
4
6402
0 0035
i
0 0030
0 0025
e
2
00020
i
k
?
t-
0
w
a
00015
LL
I
0 0010
I
0 0005
4
0
f
10
0
20
30
40
50
60
70
80
90
too
REACTOR PERIOD ( sec 1
i
,
Fig.
=
i
5.4.
Regulating Rod R e a c t i v i t y vs Reactor Period for F u e l Flow Rates of
experimental points show the extent of the rod
movement for that point. Again, there i s no conc l u s i v e evidence that the reactivity of the rod per
unit length was not constant.
Therefore, the
average of the 48-gpm flow data weighted over the
rod movement for each run gives
i
I
(&/K)/in.
1
For the two zero-flow points the rod was moved
over the middle 10 in, of i t s 12-in. travel i n two
6-in. overlapping runs covering the upper and lower
halves o f the rod, respectively,
The weighted
average of these points gave a value o f
I
(Ak/k)/in.
?
= 0.029Win.
*
= 0.033%/in.
,
which i s i n excellent agreement with the fuel
addition calibration. This value was settled upon
as the best value of reactivity per inch over the
whole length of the regulating rod. The values for
the reactivity vs rod position are given i n Table
5.6.
0 and 48 gpm.
C A L I B R A T I O N O F SHIM RODS VS
REGULATING ROD
With the reactivity of the regulating rod known,
the shim rods could be calibrated against it (exp.
L-6). This calibration could then be compared
with the previous calibration made during the fuel
addition. It was assumed that the three shim rods
were enough a l i k e that a calibration o f one rod
would be representative, and rod No. 3 was chosen
for the c o l i bration.
The calibration procedure consisted of setting
shim rods Nos. 1 and 2 i n a fixed position, and
then, with the reactor a t a power level of 1 w on
the servo mechanism, rod No. 3 was moved u n t i l a
specified travel of the regulating rod was obtained.
A t the start o f the experiment the rods were adjusted so that shim rod No. 3 was nearly a l l the
way out (position indicator at 35 in.) and the regulating rod nearly a l l inserted (position indicator a t
3 in.). After recording the position of a l l rods,
49
d
i
50
In
m
z wJ
In
;1
wJ
2 2
rlL
N
z
9
O
N
-
M
(
u
0
-
1)1 I
2 I-
N
-
*
i
P
I
h
i
t
a
P
i
&
5-
l.
h
h
b
t
L
J
?
t
t
I
I
!
t
!
i
k
I
f
i
ia
b
0 40
-<
035
z
I
t.
t
2
9
0 3 0
n
0 25
0 04
003
-
c
,
-z
> 002
t
2
Io
W
4
n
0 01
0
2
3
4
5
6
7
8
9
to
11
12
43
14
REGULATING ROD POSITION ( i n )
Fig.
5.7.
Calibration of Regulating Rod from Reactor Period Measurements.
shim rod No, 3 was inserted u n t i l the servo action
had withdrawn the regulating rod a compensating
The action was then
distance of about 10 in.
stopped and the new rod positions were recorded.
With shim rod No. 3 set, shim rods Nos. 1 and 2
were then adjusted to bring the regulating rod back
to i t s original position, and again the rod positions
were recorded. Again rod No. 3 was inserted u n t i l
a 10-in. withdrawal of the regulating rod was
attained.
This process was repeated u n t i l rod
No, 3 had been inserted about 14 in., and each
new position had been recorded. This calibration
agreed well with the calibration against fuel addition. Details of the results are given i n Appendix J.
F U E L SYSTEM C H A R A C T E R I S T I C S
The fuel system characteristics with the fuel
concentrate i n the system were determined for the
52
f i r s t (and only) time the day after the reactor f i r s t
became c r i t i c a l (exp. L-3). The final amounts o f
concentrate (i.e., the excess to provide for burnup
and poisoning during the subsequent power runs)
had not yet been added t o the system.
Consequently, the fuel density at the time the data were
taken was 3.32 g/cm3 as compared w i t h 3.36 g/cm3
for the final fuel composition, and 3.08 g/cm3 for
the carrier; a l l densities were determined a t 1300OF.
The measured and calculated f l o w data as a
function of pump speed are shown i n Fig. 5.8. T h e
flow was calculated for two pump speeds by using
the time between p i p s on the f i s s i o n chamber as
concentrate was f i r s t added t o the system. The
straight l i n e joining the two calculated points i s
considered t o be a good approximation because o f
the regime i n which the pump was operated.
i
T A B L E 5.6.
REGULATING ROD REACTIVITY CALIBRATION FROM REACTOR PE RlOD MEASUREMENTS
(Exp.
No.
Run
Fuel F l o w
(gpm)
L-5)
Average
Average
h/k
Period
from lnhour
(set)
Equation
Movement of
Regulating Rod
Reactivity,
Regulating Rod
(Ak/k)/in.
(in.)
(%/in.)
Position
(in.)
~~
1
48
48
0.00036
7.5
1.1
0.0318
2
48
22.2
0.00060
7.6
2.C8
0.0289
3
0
21
0.00206
9.6
6.05
0.0330
4
48
26
0.00054
3.8
2 .o
0.0270
5
48
22
0.0006 1
11.1
2.2
0.0278
6
0
21.6
0.0020
5.6
6.1
0.0328
7
48
20.1
0.00065
5.2
2.08
0.0313
0.00060
9.3
2.06
0.0291
8
22.2
48
Average for 0-gpm flow
Average for
48-gpm
flow
0.0329
0.029
ORNL-LR-DWG
6407
30
25
(I)
20
/
a
&/<FLOW
5
.-
FROM FLOWMETER
I
+
I
a 400
n
w
a
;15
200
[L
3
/,
u)
W
v)
a
0
10
5.8. F u e l F l o w R a t e a s
a Function of
Pump
5
The flow data as a function o f the system head
loss from the reactor i n l e t to the pump tank are
The f l o w data are plotted digiven i n Fig. 5.9.
r e c t l y vs head (by using the flow rates determined
from pump speed as given by the upper curve i n
Fig. 5.8) and then as corrected for a “live” zero
0
Fig.
Speed.
0
10
20
30
40
50
FLOW RATE ( g p m )
F i g . 5.9. F u e l F l o w R a t e vs the Pressuee Head
from the Pump Suction t o the Reactor Inlet.
53
on the pressure transmitter from which the head
values were obtained.
The latter data fit the
Furthermore, these
theoretical curve very we 11.
system data fit the pump performance data, i f it i s
assumed that there i s a 7-psi drop from the pump
discharge to the reactor i n l e t at design flow (i.e.,
46 gpm); there i s no experimental measurement of
t h i s pressure drop. The pump performance characteristics determined i n a test stand run with fuel at
130OOF are given i n Fig. 5.10.
E F F E C T OF F U E L FLOW
ON REACTIVITY
An experiment was performed i n which the effect
of fuel flow on reactivity was observed (exp.L-7).
For this experiment the reactor was brought to a
nominal power of 1 watt and given over to the
f l u x servo mechanism after f u l l fuel flow of 46 gpm
had been established. After recording the positions
of the regulating and shim rods the flow rate was
reduced step-wise unti I zero-flow was reached.
Each reduction of the flow was accompanied by a
change of position of the regulating rod. Reducing
the flow had the effect of allowing more delayed
neutrons to contribute to the flux level i n the
reactor, and therefore the servo mechanism inserted
the regulating rod to compensate for the excess
reactivity created by the delayed neutrons. Each
rate of fuel flow and the corresponding regulating
rod position were recorded.
The results of the
experiment are shown in Fig. 5.11, where A k / k i s
plotted as a function of flow rate. It i s observed
that 12 in, of rod movement, or 0.4% A k / k , was
needed to compensate for the reactivity introduced
when the flow rate was reduced from f u l l to zero
flow.
This curve i s comparable t o that shown i n
Fig. D.2 of Appendix D. The data for t h i s experiment are tabulated i n Table 5.7.
Although not a part of experiment L-7, additional
insight into fuel a c t i v i t y may be obtained from
examination o f Fig. 5.12, which shows activity as
Before the
recorded by f i s s i o n chamber No. 2.
data shown i n the figure were recorded the reactor
had been operating at a power o f about 1 w with
the fuel pump stopped. As shown, the reactor was
60
60
50
50
40
40
f
k
L.
I
c
t
i
d
.-..
e
..-.
*
r
>
u
30
30
0
I
LI
LL
W
20
20
4
10
IO
0
3
Fig. 5.10. F u e l Pump Performance Curves,
54
ORNL-LR-OWG
ORNL-LR-DWG 6408
04
105
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2
g
0
I
d
3854
42 0
03
90
I
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2 02
+
0
75
; 60
1
_I
3
8
:
LL
4
IL
W
5
45
I
0
REACTIVITY CHANGE DUE TO
STOPPING SODIUM FLOW
TpHdQ,a
v)
01
30
r
W
X
15
0
0
0
Fig.
tivity.
5
10
5.11.
T A B L E 5.7.
15
20
25
30
35
FLOW RATE (qpm)
No.
1
2
3
4
5
6
7
45
50
Effect of F u e l Flow Rate on Reac-
E F F E C T OF F U E L FLOW RATE
ON REACTIVITY
(Exp.
Run
40
L-7)
Fuel
Sodium
Movement of
Chonge in
Flow
Flow
Regulating Rod
Reactivity,
(gpm)
(gpm)
(in.)
45.5
41
35
28.5
23.8
149
149
149
149
149
149
149
1 49
149
149
149
149
149
149
0
149
0
0.7
1.4
1.9
2.4
3 .O
3.45
12.05
3.45
12.05
3.65
3.65
0.9
0.96
1.12
0.96
ia
a
13.5
0
9
10
11
12
13
14
15
16
13.5
0
13.5
13.5
43.5
41
41
41
Ak/k
(n)
0
0.023 1
0.0462
0.0627
0.0792
0.099
0.1 14
0.398
0.114
0.398
0.120
0.120
0.0297
0.031 7
0.0370
0.0317
then scrammed' and the fuel pump started. The
fuel which had been in the reactor before the scram
was more active than the remainder o f the fuel and
'Scram, o s used in this report, refers to on intentional
shut down of the chain reaction b y suddenly dropping
t h e control rods into the core.
Fig. 5.12.
Circulating F u e l Slugs after Starting
Pump Following a Reactor Scram.
it was t h i s excess activity which showed up as the
fuel was circulated after the reactor was scrammed.
LOW-POWER M E A S U R E M E N T O F T H E
TEMPERATURE COEFFICIENT
A measurement of the temperature coefficient of
reactivity (exp. L-8) was made next.
With the
reactor isothermal a t 1312°F and controlled by
the flux servo mechanism, the heat barrier doors
were raised and the helium blower turned on to
200 rpm to cool the fuel i n the heat exchanger.
As the temperature of the reactor decreased, the
servo began t o drive the regulating rod i n t o compensate for the increased reactivity. The recording
o f the regulating rod position vs time i s shown i n
Beginning a t 0218, the rod was inFig. 5.13.
serted by the servo quite rapidly, and by 0221 the
With the
rod was approaching i t s lower limit.
regulating rod on servo, one of the shim rods was
T h i s shim rod
inserted as rapidly as possible.
insertion overcompensated for the increase i n
reactivity due to the temperature drop, and the
55
ORNL-LR-D'fiG
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0235
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EXP L - 8 , NOV 8 , 1 9 5 4
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0217
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REGULATING ROD
0258
LOWER END OF
j
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0214
0213
0212
12
I3
II
10
9
8
7
6
5
4
3
2
4
0
REGULATING ROD INSERTED (in.)
Fig, 5.13.
Coefficient
.
Regulating Rod P o s i t i o n During Low-Power Measurement of the Reactor Temperature
servo quickly withdrew the regulating rod.
Insertion o f the shim rod was stopped when the
regulating rod had been withdrawn to near i t s
upper limit, and the regulating rod was again inserted by the servo to compensate for the reactivity
introduced by the cooling o f the fuel. By 0230, the
outlet temperature i n the heat exchanger was
approaching i t s lower limit, 1150°F, so the helium
blower was stopped and the run was ended.
A trace from the chart for recording the mean fuel
temperature i s shown i n Fig, 5.14, and superimposed on i t i s a plot o f the displacement of the
56
regulating rod as obtained from the previous figure.
A measure o f the temperature coefficient may be
obtained by comparing the slopes o f the two curves.
At 0228 the rod was moving a t exactly 1 in./min,
corresponding t o an increase in reactivity of
3.3 x
(Ak/k)/min.
At the same time, the
mean temperature was dropping a t the rate o f
5.1°F/min.
The resulting reactor temperature
coefficient, as obtained from these data, would
(Ak/k)/OF.
3 y comappear t o be -6.48 x
paring the slopes of the curves obtained earlier
during the run, a considerably higher coefficient
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EXP. L - 8
N O V 8,1954
1
I
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LL
0
v
-
1350
C
t-
a
t7
LT
W
W
z
t
a
I
W
1
L
0
W
4
2
'$
3
:
24
1300
16
a
tl1
I
n
0
0
J
W
K
3
LL
8
1250
0
0215
4
0220
0225
0230
0235
T I M E OF DAY
Fig. 5.14.
Regulating Rod Position Superimposed on F u e l Mean Temperature Chart.
can be obtained that might approximate that for the
fuel.
In any case, the recorded mean temperature
i s believed to be in error (as explained in Appendix K) i n such a manner that the actual rate o f
temperature change i s about 40% higher than that
The measured value of the
given in Fig. 5.14.
temperature coefficient was corrected accordingly
t o give a value o f -4.6 (Ak/k)/O F for the low power.
T h i s value i s comparable t o that subsequently
obtained during high-power operation and, as noted
in the subsequent discussion, the data from experiment H-8 yielded t h e best values for both t h e fuel
and the reactor temperature coefficients.
The apparent time lag between the reactivity and
the minimum mean temperature, as f i r s t noted during
the subcritical measurement of the temperature
coefficient, was s t i l l i n evidence here. Possible
explanations of the time lag, as well as other
detai Is o f the various temperature coefficient
measurements, are given i n Appendix 0.
i
A D J U S T M E N T O F CHAMBER POSITION
4
'
During the c r i t i c a l experiment and the low-power
operation, it was desirable t o have the nuclear
instruments positioned t o read a t as low a power
as possible.
The chambers were therefore as
close t o the reactor as possible, Before increasing
the power, i t was necessary to adjust the positions
o f the nuclear chambers so that they would read a t
This was accomplished by
higher power levels.
adjusting the power level of the reactor t o about
100 w and then moving the compensated log N
chamber away from the reactor so that i t s reading
was changed from 0.032 to 0.001, or i t s upper l i m i t
was extended by a factor of about 30. With the
reactor power at about 1 kw, the micromicroammeter
chamber was moved away from the reactor s o that
i t s reading was changed from 41.4 on the 5 x
amp range t o 54 on the 1 x lo-* amp range, T h i s
extended the upper l i m i t o f the meter by a factor o f
While one chamber was adjusted the
about 40.
reactor power was held constant by the other
chambers so that the correlation between instrument readings was not lost. The nuclear power for
the experiment was determined on t h i s basis.
I n the final analysis, the correlation between
chamber readings and reactor power can be based
on .the 25-hr xenon run during which a good value
for the extracted power was obtained. The micromicroammeter read 52.5 on the 2 x lo-' scale
when the log N read 22.5, and both corresponded
t o an extracted power o f 2.12 Mw.
57
i
6. HIGH-POWER EXPERIMENTS
I
I
The low-power tests were concluded by noon on
November 8. During the last few days of the lowpower tests a l l last minute details in preparing the
system for high-power operation had been under
way. These preparations included lubricating a l l
rotating equipment, r e f i l l i n g o i l reservoirs, changing filters, replacing motor brushes, checking a l l
instrumentation, etc. By 1445 on November 8 the
p i t s had been sealed, i.e.,
covered with three
layers of 2 4 - f t - t h i c k concrete blocks and calked,
and the experiment was ready for the high-power
phase of the operotion, the chronology of which i s
given i n Fig. 6.1.
Prior t o the start of the high-power experiment
the final addition of fuel concentrate, which was
t o provide the reactor with sufficient excess
reactivity t o overcome burnup and fission-product
poisoning, was made. This last addition consisted
of 15.419 Ib of concentrate diluted with 32.379 Ib
of partially enriched fuel (part of that removed from
the system earlier when the pump level was trimmed)
and was injected into the system through the
sample l i n e on the morning of November 7. The
sample l i n e was used for t h i s final enrichment
operation because the transfer I ine through which
the concentrate had hitherto been injected into the
system had again developed a leak. It was necessary t o lower the melting point of the concentrate
(by dilution) because it was not certain that the
sample line could otherwise safely maintain the
high temperature required.
High power, which may be defined for the ARE
as anything over 0.1 Mw, was approached i n a
series of successive steps of increasing power.
The power regime was not attained at once, because
leakage of fission a c t i v i t y from the system into the
p i t s and subsequently from the p i t s into the occupied parts of the building required that provision
be made to maintain the p i t s a t a subatmospheric
pressure.
The experiment could then continue
safely, and a power of greater than 2 Mw was f i r s t
attained early during the evening of November 9.
The remaining time (70 hr) of the ARE operation
was devoted t o several experiments to determine
the high-power behavior of the reactor; these experiments are described i n detail i n the following
sections. Starting i n the evening of November 9,
the fuel temperature coefficient of reactivity (Exp.
H-4) and the over-all reactor temperature coefficient (Exp. H-5 and Exp. H-9) were measured.
58
The early hours of November 10 were used t o take
the reactor t o power from subcritical by making use
of the negative temperature coefficient (Exp. H-6).
Later in the day the sodium temperature coefficient
was measured (Exp. H-7), some other reactor
kinetics were studied (Exp. H-8), and the moderator
temperature coefficient was measured (Exp. H-10).
During t h i s time the operating crews had opportunity
t o familiarize themselves with the handling of the
reactor at power.
At 1835 on November 10, a 25-hr ful I-power xenon
run was started. For t h i s run the reactor was held
in steady-state operation a t 2.12 Mw by the negative
temperature coefficient and a t a mean temperature
Since equilibrium temperature condiof 1311°F.
tions prevailed during t h i s run, a very good measurement of extracted power could be made. The
run ended a t 1935 on November 11. Later that
night, the effect on the power of stopping the
sodium flow was observed (Exp. H-12), and at
2237 a !,O-power run (Exp. H-13) was started to
observe the effect of xenon buildup and continued
for a period o f 10 hr. During t h i s run the reactor
At 0835 on
power was approximately 200 kw.
November 12 no effect of xenon buildup had been
observed, and therefore the experiment was terminated.
The morning of the last day, November 12, was
devoted t o some more reactor kinetics studies and
a measurement of the maximum extracted power
available (Exp. H-14), which turned out t o be about
2.5 Mw. On the afternoon of November 12 the
building was opened t o visitors with proper clearance, and the behavior of the reactor was demonstrated t o them. At 2004, an estimated total operating time of 100 Mw-hr having been accumulated,
the reactor was scrammed and the ARE experiment
was terminated.
b
-
k
L
b
i
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i
APPROACHTOPOWER
t
I
The i n i t i a l phase of high-power operation of the
ARE began a t 1445 on November 8. As shown i n
Fig. 6.1, the power was raised in a series of
At each power level, when
successive steps.
steady-state conditions had been obtained, a
complete set of control room data was taken. The
f i r s t experiment, H-1, was a 20-kw run. During
t h i s run the safety chambers were withdrawn and
At 1619 the
reset t o scram a t about 260 kw.
reactor was shut down because of the presence of
I-
i
.
i
Y
3
1
1
0
0
e
W
r
L
0
0
N
N
1
3
u
I
h-l
I
-
n
X
m
z
c
w
z
0
Z
X
m
5
0
v
r
C
7,
s1
0
Z
m
I
0
0
m
EXP. H-3
ul
\
T
0
0
A
I
.
-,-I-
TIME
J
/
0
Z
rn
0
-
s1
///'I-\\
OF DAY
s
0
0
0-4
90
Zm
I
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8
0
0
80
2:
0
0-
-lX
sz
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0
0
0
0
W
m
XJ
z
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z
0
<
n
VI
W
2
n
ul
W
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0
P
ul
W
A
W
cn
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m
X
73
m
30
m
d
W
0
I
r
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<
0
0
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6
t
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i
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f
airborne radioactivity in the basement. Apparently
the gas f i t t i n g s t o the main fuel pump were leaking
and fission-product gases and vapors were leaking
into the p i t s and from the p i t s into the basement
through defective seals i n some of the piping and
electrical junction panels connecting the basement
with the pits.
In order t o prevent the leakage of activity into
the basement it was decided t o operate the p i t s a t
a slightly subatmospheric pressure. Accordingly,
a 2-in. pipeline was run in a direction south of the
ARE building for a distance of 1000 ft and termi-
I
P
I
d
1
-.
nated in an uninhabited valley. Then, by means o f
portable compressors and a jet, the p i t pressure
was lowered by about 6 in. H,O, with the exhaust
from the compressors being bled into the 2-in.
pipeline. The p i t s were maintained at a negative
pressure for the balance of the experiment.
At 1125 on November 9 the reactor was again
brought t o 20-kw power (Exp. H-2) for the purpose
of testing the new off-gas system t o determine
,
whether or not the negative pressure of 6 in. HO
in the p i t s was enough t o keep the hot gases from
diffusing into the basement. During t h i s run the
p i t activity, as recorded on the p i t monitrons, was
watched, and i t s decay after shutdown of the
reactor a t 1246 was observed. From t h i s run it was
ascertained that the subatmospheric p i t pressure
would prevent p i t a c t i v i t y from leaking into the
bui Iding,
At 1520 on November 9 the approach t o power
was started again (Exp. H-3). At 1523 the reactor
was brought t o a nominal power of 200 kw on a
17-sec period and leveled out. s i x minutes later
the barrier doors were raised and the f i r s t attempt t o extract power was made. A t 1529 the fuel
helium blower was turned on and the blower speed
The nuclear power
was increased t o 300 rpm.
again began t o increase, but the safety chambers
had not been withdrawn far enough and they
scrammed the reactor when the power level had
reached about 260 kw.
After another false start, the safety chambers
were withdrawn again and set t o scram a t about 4
Mw.
The reactor was then brought t o 100 kw,
leveled out, and placed on flux servo a t 1625. At
1626 it was taken o f f servo and the blower speed
was increased, t h i s time t o only 170 rpm. The
reactor power slowly rose and a t 1643 had leveled
off on the reactor temperature coefficient a t about
500 kw. After a set of data had been taken and
the reactor mean temperature elevated, the blower
speed was increased t o 400 rpm and the power level
continued t o rise.
B y 1715 the power level had
leveled out a t near 1 Mw.
The sodium blowers were turned on (at 1740) and
brought t o a speed of about 500 rpm. Half an hour
later the fuel blower speed was increased t o 1590
rpm, and i n 10 min a steady-state power level of
about 2 Mw was reached. At t h i s time the nuclear
and process instruments read as follows:
Log
N power,
'
1.32
1.38
1203
1440
1321
23 7
46
1281
MW
Micromicroammeter power,' M w
Reactor inlet fuel temperature,'
O F
Reactor outlet fuel temperature,'
OF
Reactor mean fuel temperature,2 "F
Fuel temperature difference across reactor,'
O F
Fuel flow, gpm
Reactor sodium mean temperature, OF
Sodium temperature difference across reactor,
O F
Sodium flow, gpm
10
152
From these data the extracted power was calculated
as follows:
Power extracted from sodium, Mw
1.16
0.05
Power extracted from rod cooling, Mw
0.02
Power extracted from fuel, Mw
~
Total extracted power,2 Mw
1.23
However, it was subsequently discovered that the
temperatures upon which this extracted power was
calculated were i n error and, as discussed in
Appendix K, the actual temperature gradient a t t h i s
Therefore, the extime was about 40% higher.
tracted power was correspondingly higher than that
listed, A general discussion of the extracted power
appears in a following section.
The above-described power operation was rnaintained for about 1 hr, during which time no further
trouble was occasioned by the fission gases.
Power was reduced a t 1900 and the reactor scrammed
a t 1915, and thus the i n i t i a l phase of the highpower operation was concluded.
T E M P E R A T U R E C O E F F I C I E N T MEASUREMENTS
Several experiments were performed in an attempt t o measure the effect on reactivity o f temperature changes of the various components of the
'Set on the b a s i s o f the power calibration from fuel
activation.
'The
temperatures and consequently the extracted
power volues are i n error a s noted i n the discussion
following the data.
61
reactor.
In addition t o the over-all reactor coefficient, experiments were run t o obtain data on
the instantaneous fuel temperature coefficient and
the coefficients attributable t o the sodium and the
beryllium oxide moderator.
There were several
anomalies that developed in the course of these
experiments, the two most significant being (1) the
interpretation of the lag i n the response between
the rod movement and the change in reactor temperatures as power was abstracted from the reactor
and (2) the discrepancy between the thermocouple
indications a t the reactor (fuel and sodium i n l e t
and outlet temperatures) and those along the lines
t o and from the reactor.
These anomalies are
discussed in detail in Appendixes 0 and K,
respectively. It was concluded that the time lag
was not real (i,e., the fuel temperature should show
changes as soon as the reactivity does) and that
the l i n e temperature most accurately reflects the
actual stream temperatures.
In the determination of a temperature coefficient
by varying the power extracted from the reactor,
the recorded temperatures are subject t o correction,
as described in Appendix K, However, when the
temperature coefficient was determined, as i n
Exp. H-5, by withdrawing the shim rods so that the
reactor slowly heated i t s e l f while the extracted
power remained constant, no correction needed t o
be made for the temperature.
Furthermore, t h i s
technique minimized the temperature lag anomaly,
and the fuel and reactor temperature coefficients
derived from Exp. H-5 are therefore believed t o be
the best experimental values for these coefficients.
F u e l and Reactor Temperature Coefficients
The fuel and reactor temperature coefficients
were both obtained from the same experiment, the
fuel being the instantaneous coefficient, the
reactor the “equilibrium” coefficient. The i n i t i a l
attempts (Exp. H-4) t o measure these coefficients
were not satisfactory because o f the difficulty i n
interpreting the results, and therefore the measurement was repeated (Exp. H-5) in a somewhat
different manner.
In the f i r s t experiment, H-4, the procedure was
simi lor t o that described previously under ‘‘LowPower Measurement of the Temperature Coefficient,’]
and the data are presented in Figs. 6.2 and 6.3.
The f i r s t of these figures i s the recording of the
position o f the regulating rod as it was inserted by
theservo as the fuel was cooled by the helium flow
i n the heat exchanger. Figure 6.3 i s a recording of
62
the mean fuel temperature w i t h the rod displacement
Deas determined from Fig. 6.2 superimposed.
pending upon the time at which the slopes of the
two curves are compared, the temperature coefficient
appears t o decreasefrom a value of --1.6 x
(Ak/k)/OF to -4.43 x
(Ak/k)/”F
toward the
end o f the run, These values represent the recorded
data and are of course subject t o the correction for
temperature, as discussed i n Appendix K, even as
the data are subject t o various interpretations because of the lag i n the temperature data (see
Appendix 0). Furthermore, the changes were made
quite rapidly, the entire run lasting only 5 min.
In an attempt t o minimize the effect of the temperature anomalies, the temperature coefficients were
then determined from an experiment (Exp. H-5) in
which the extracted power was held constant and
the shim rods were withdrawn, Starting with the
reactor a t a mean temperature o f 126OoF, the withdrawal o f t h e shim rods allowed the reactor to
gradually heat the whole system t o 1315OF. Since
the reactor was a slave t o the heat extraction
system, which was not changed, the extracted
power remained virtually constant during t h i s time.
T h i s process took about 35 min during which the
nuclear power was held constant a t a nominal power
of 200 kw by the flux servo. The regulating rod
gradually withdrew t o compensate for the decrease
i n reactivity as the reactor temperature increased.
The data are tabulated in Table 6.1, and a plot of
rod withdrawal vs mean reactor temperature i s
Inspection o f Table 6.1
presented i n Fig. 6.4.
shows that the rate o f change of t h e i n l e t and
outlet fuel temperatures followed the change in
mean fuel temperature very closely, and thus gave
assurance that the fuel throughout the reactor was
changing temperature a t a uniform rate.
T h e most reliable values for the temperature
coefficients of the ARE were obtained from t h i s
experiment.
For the f i r s t 6 min the plot of rod
withdrawal vs reactor mean temperature shows a
Since 1 in. of rod movement
slope of 0.296 in./OF.
Ak/k, the i n i t i a l
i s equivalent t o 3.3 x
(Ak/k)/OF.
temperature coefficient was -9.8 x
After the f i r s t 6 min and up u n t i l conclusion o f the
run, the curve in Fig. 6.4 demonstrates a constant
value of the temperature coefficient o f -6.0 x
(Ak/k)/OF.
These are the best values for these
coefficients that were obtained, It should be noted
that the value for the over-all temperature coefficient
agrees t o w i t h i n 25% w i t h the value previously
obtained from the low-power experiments. However,
i
i
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ORNL-LR-DWG
6412
2127
-INSERTION
WITHDRAWAL-
-OFF
SERVO CONTROL
2126
2125
E X P . H - 4 , NOV.
>a
n
9,1954
UPPER E N D OF
ROD T R A V E L
2124
ROD WITHDRAWN
BY INSERTING SHIMS
TO OVERRIDE SERVO
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0
W
5
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rl
LOWER E N D
ROD TRAVEL
2123
-
2122
m
BLOWER ON
REGULATING ROD
TRAVEL
2121
2120
12
11
io
9
8
7
6
5
4
3
2
1
R E G U L A T I N G ROD I N S E R T E D (in.)
F i g . 6.2.
ment H-4.
OF
( 3 5 0 rpm)
i
0
Regulating Rod Position During Measurement of Reactor Temperature Coefficient.
Experi-
ORNL-LR-DWG
32
6413
1400
I
LL
28
W
-
[z
c
C 24
t,
3
+
a
1350
20
5
W
16
d
i
z
a
a
E 12
n
n
W
1300
[z
g 8
0
t-
4
W
o
a
[z
0
4
,
t-
_1
i
-
6
a
5
W
2121
2122
2123
2124
2125
2126
2127
1250
2128
TIME OF DAY
Fig.
6.3.
Regulating Rod Position Superimposed on F u e l Mean Temperature Chart.
Experiment
H-4.
63
T A B L E 6.1. REACTOR TEMPERATURES AND ROD POSITIONS DURING
100-kw MEASUREMENT O F TEMPERATURE COEFFICIENTS (EXP. H-5)
Time
2223
24
25
26
27
25
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
54
55
56
57
58
59
2300
01
02
03
04
05
06
07
08
09
10
64
Regulating Rod
Position
(in.)
Rod
Withdrawal
3.72
3.84
4. 15
4.43
4.69
4.96
5.23
5.50
5.77
6.02
6.28
6.49
6.68
6.90
7.10
7.35
7.50
7.69
7.92
8.14
8.32
8.55
8.75
8.94
9.11
9.31
9.50
9.68
9.85
10.05
10.23
10.45
10.60
10.78
10.99
11.17
11.41
11.54
11.80
11.95
12.20
12.38
12.57
12.75
12.90
13.19
13.25
0
0.12
0.43
0.71
0.97
1.24
1.51
1.78
2.05
2.30
2.56
2.77
2.96
3.18
3.38
3.65
3.78
3.97
4.20
4.42
4.60
4.83
5.33
5.22
5.39
5.59
5.78
5.96
6.13
6.33
6.5 1
6.73
6.88
7.06
7.27
7.45
7.69
7.82
8.08
8.23
8.48
8.66
8.85
9.03
9.18
9.47
9.53
(in.)
Reactor Mean
Reactor Inlet
Reactor Outlet
Temperature
Temperature
Temperature
(OF)
(OF)
1262.5
1263.5
1264
1265.5
1266
1267.5
1269
1271
1272
1273
1274.5
1275.5
1277
1278
1279
1280.5
1281.5
1282.5
1283.5
1284.5
1285.5
1286.5
1287.5
1289
1290
1291
1292
1293
1294
1295
1296
1297
1298
1299
1300
1301
1302.3
1303
1304
1305
1306
1307
1308
1309
1310
131 1
1283
1285
1286
1287
1288
1290
1291
1292
1293
1294
1296
1297
1298
1300
1301
1302
1303
1304
1305
1307
1308
1309
1310
131 1
131 2
1313
1314
1315
131 6
1317
1318
1319
1320
1321
1322
1323
1324
1325
1326
1327
1328
1329
1330
1331
1332.5
1333
(O
F)
1263.5
1265
1266
1267
1268
1269
1270.3
1271.5
1273
1274
1275
1276.5
1277.5
1278.5
1279.5
1281
1282
1283
1283.5
1285
1286.3
1287.5
1288.5
1289.5
1291
1292.5
1293.5
1294.3
1295
1296.5
1297.5
1298.8
1299.5
1300.5
1301.3
1302.5
1303.5
1305
1306.3
1307
1308
1308.8
1309.8
1310.5
131 1.5
1312.5
1313.5
-
L
1
r
b
k
b
d
t
P
I:
= 5I
a
1
5
k
ORNL-LR-DWG 6444
1
?
s
1265
1260
1270
1275
1280
4285
1290
1295
(300
1305
1310
1315
1, REACTOR MEAN TEMPERATURE (OF)
;
i'
-
1
Fig.
f
i
6.4.
Regulating Rod Movement a s a Function of Reactor Mean Temperature.
the instantaneous value o f -9.8 x
(Ak/k)/'F
i s much more important from the reactor control
standpoint, and it was t h i s large fuel temperature
coefficient which made the ARE demonstrate the
excellent stability described below under t h e
subtitle "Reactor Kinetics."
The value for the reactor temperature coefficient
was subsequently confirmed in a later experiment
(Exp. H-9) in which the system temperature was
changed by gradually cooling the sodium, which in
turn, cooled the fuel and the reactor.
While the
curve obtained from t h i s experiment was very
similar t o that given in Fig. 6.4, the i n i t i a l slope
could not be that due t o the fuel temperature coefficients, because the fuel i s one of the l a s t constituents i n the reactor to feel the temperature
change. The "equilibrium" slope, however, should
be that determined by the reactor temperature coefficient, and a value o f -6.3 x
(~k/k)/~F
Experiment
H-5.
was obtained. No temperature correction was required because the extracted power in the fuel
system did not change (although that i n the sodium
T h i s vatue agrees very w e l l with
system did).
that obtained i n the preceding experiment.
Sodi um Temperature Coeff icienf
In order t o measure the sodium temperature coefficient an experiment (Exp. H-7)was performed
in which the sodium was cooled and the corresponding changes in sodium temperature and reactivity
were measured. In t h i s experiment it was mandatory that the sodium be cooled rapidly and the
data recorded before the moderator had time t o
cool.
Otherwise, since the sodium bathes the
moderator, the moderator temperature would follow
the sodium temperature and introduce an extraneous
effect due t o i t s (i.e.,
moderator) temperature coefficient. Accordingly, with the reactor a t about
65
33 kw and no power being extracted from either
the fuel or the sodium, the sodium blower speeds
(2 blowers) were quickly raised from 0 t o 2000
rpm.
The reactor period immediately started t o
change and went from i n f i n i t y t o 50 sec in 1.5
min, which, as shown i n Fig. 5.4, corresponds t o
A k / k . The cona reactivity change of 3.5 x
sequent rate of reactivity change was 2.33 x
(Ak/k)/min.
In t h i s experiment the regulating rod,
as well as the shim rods, was held in a fixed
position.
During the 1.5-min interval, the fuel
temperature was changing at a rate of about
-1"F/min and the sodium temperature was changing
at a rate of -2.3'F/min.
(The rate of change of
fuel mean temperature was corrected for the discrepancy i n the recorded mean temperature, as
described i n App. K, but no comparable correction
was necessary i n the rate of change of the sodium
temperature.) B y applying the previously mentioned
(Ak/k)/OF for the instantanevalue o f -9.8 x
ous fuel temperature coefficient t o give a rate o f
reactivity change for the fuel of 0.98 x
(Ak/k)/min
and subtracting from the rate of reactivity change observed upon cooling the sodium,
it i s found that the reactivity change caused by
the decrease in t h e sodium temperature i s
2.33 x
(Ak/k)/min
- 0.98
x
(observed)
(Ak/k)/min
1.35 x
=
(fuel)
(Ak/k)/min
.
Then, by applying the observed rate of change o f
the sodium temperature, the sodium temperature coefficient i s found to be
1.35 x
(Ak/k)/min
-2.3O F/m i n
=
-5.88 x
(Ak/k)OF.
T h i s value i s v a l i d i f it i s assumed that the
transient took place rapidly enough that there was
no appreciable change i n the moderator temperature.
The measurements were, however, subject
t o considerable error because the temperature
changes involved were so small as t o be quite
d i f f i c u l t t o detect and a correction was necessary
in the fuel temperature because of thermal lags.
Moderator Temperature Coefficient
An experiment (H-10) was conducted in order t o
determine the temperature coefficient of the moder-
66
ator. During t h i s experiment the fuel temperature
was held constant and the speed o f the blowers
for cooling the sodium was increased t o change
the temperature of the moderator coolant. Since
t h i s was done very slowly, the moderator did cool
down, in contrast t o the earlier experiment (H-7)
in which the sodium temperature was changed SO
rapidly that the moderator temperature was unable
t o follow the sodium temperature,
The earlier
experiment gave information as t o the temperature
coefficient of the sodium alone, whereas experiment H-7 gave the combined effect of sodium and
moderator changes, The reactivity was indicated
by the position t o which the regulating rod was
adjusted by the flux servo.
T h e changes i n the sodium and the moderator
temperatures slightly increased the heat loss of
the fuel, and the power had t o be increased slightly
In
t o keep the fuel mean temperature constant.
fact, in the middle of the run the reactor mean
temperature was lower than a t the beginning, but
before the final reading w a s taken, the reactor
mean temperature was brought back t o i t s original
value.
Table 6.2 gives the pertinent data. The
sodium i n l e t temperature was taken from a recording o f the temperature of the reflector coolant inlet
4 in. from the bottom of the reactor and the sodium
outlet temperature was taken from a recording of the
temperature of the reflector coolant outlet 3 in. from
the top of the reactor. The time recorded i n column
one i s the time when a reading was taken. The
change of the blower speed preceded t h i s time by
a few minutes t o allow the temperature equilibrium
t o be established. As an indication of the establishment of the equilibrium, use was made of the
leveling off of the trace on the sodium temperature
differential recorder.
A s can be seen from the table, the decrease of
the average sodium temperature from 1273 t o
1246OF corresponds t o a withdrawal of t h e regulating rod from the &in. position t o the 8.9-in. position. A t the lower temperature the reactor was less
reactive and showed that the temperature coefficient
was positive. The magnitude of t h i s temperature
coefficient was small, (0.9 in./27OF) x 3.33 x
(Ak/k)/in.
= +1.1 x
(Ak/k)OF.
T h i s coefficient is, however,the sum o f the sodium temperature coefficient and the moderator temperature
coefficient. Since the sodium temperature coefficient
the moderator temperature COwas -5.9 x
efficient must be +6.9 X lo-' (Ak/k)/OF.
This
value is, of course, subject t o a l l the inherent
t
+
i
=
t
r
i
I
k
I
i
t
loe5
B
t
i
T A B L E 6.2
Time
I
MODERATOR TEMPERATURE COEFFICIENT DATA
I
Fuel
Fuel
Blower
Sodium
Regulating
Mean
Temperature
Temperature
Gradient
Speed
(rpm)
Temperature
Rod
,
(O F)
~~~
(Mw)
(" F)
P F)
No. 1
No. 2
Inlet
Outlet
Average
(in.)
1735
1313
206
960
1050
1265
1282
1273
8.0
1.98
1745
13 14
206
1160
1050
1255
1280
1267
8.0
1.98
1752
13 13
203
1170
1260
1248
1278
1263
7.6
1.98
1802
1308
200
1340
1250
1238
1272
1255
6.6
1.98
.
i
Power
Position
~~
b
Nuclear
i
I
i
i
i
1809
Increased servo demand signal so as t o withdraw rod
1812
1311
204
1480
1240
1230
1268
1249
8.4
2.12
1825
13 13
205
1470
1480
1225
1268
1246
8.9
2.12
errors of the measurement of the sodium temperature
coefficient, as well as the additional errors i n the
temperatures recorded for t h i s particular experiment.
M E A S U R E M E N T O F T H E X E N O N POISONING
7
A t 1825 on November 10 a 25-hr run a t a power
o f 2.12 Mw was started for the purpose o f measuring
the amount o f xenon b u i l t up i n the fuel (Exp. H-11).
A s discussed in Appendix P, the reactor should
have been poisoned by about 2 x
h k / k after
i t i s assumed that no xenan escaped from
the molten fuel. N o t only did the 25-hr run demonstrate that very l i t t l e o f the fission-product gas
25 hr, if
November
1825
remained i n the fuel but, also, that the reactor
possessed phenomenal stability.
Except for a
minute withdrawal of the regulating rod t o compensate for a barely detectable drop i n the mean reactor temperature, a l l readings on both reactor and
process instrumentation held constant within experimental error for the 25 hr. It was assumed that
the rod withdrawal was due t o xenon buildup in the
fuel. However, it could have been due to any of a
number o f minor perturbations, and therefore the
experiment demonstrated an absolute upper l i m i t
on the xenon poisoning.
An abstract o f the log book and data sheets
during the experiment follows:
i
10, 1954
Run started; reactor power, 2.12 Mw; rod position, 9.00 in.;
November
4
reactor mean temperature,
1311OF
11, 1954
0635
Reactor mean temperature had decreased to
1309OF; rod withdrawn t o 9.05
0750
Reactor mean temperature up t o
1020
Reactor mean temperature
1310.5OF; rod withdrawn to 9.25
1120
Reoctor mean temperature
1311OF
1435
Reactor mean temperature down t o
1559
1311OF
1932
1312OF; r o d s t i l l a t 9.30 in.; end of run
in.
131OOF; rod withdrawn to 9.15 in.
131OOF; rod withdrawn
in.
to
9.30 in.
I
!
67
It i s t o be noted that the temperature recorded
was actually one degree higher a t the end o f the
experiment than a t the start; t h i s indicates that
the withdrawal o f the regulating rod by 0.3 ’in. may,
indeed, have been too much. The withdrawal was
unquestionably an upper l i m i t on the compensation
needed for xenon poisoning and corresponded to a
Ak/k of 1 x
This was
or 5% o f the value
t o be expected i f the xenon had not l e f t the fuel
(see Appendix P). Removal o f the xenon probably
occurred by means of the swirling action of the
fuel as it went through the pump. A s mentioned
previously, fission-product gases that probably
came from a leak in the gas f i t t i n g s were detected
i n the p i t s early in the experiment.
At the conclusion o f Exp. H-11 a t 1935, the reactor was operated a t various power levels for 2 hr
i n an experiment (Exp. H-12) for determining the
effect o f the sodium flow rate on the extracted
power. A t 2237 the reactor was set a t 200 k w (onetenth full power) and h e l d there by the flux servo
for 10 hr (Exp. H-13). I f any appreciable amount
o f xenon had been built up i n the fuel from decay
o f iodine formed during Exp. H-11, the poisoning
effect should have been observed by a compensating
rod withdrawal during t h i s 10-hr period. N o appreciable rod withdrawal was observed, and therefore
it was concluded that if there was xenon poisoning
from Exp. H-11, it was negligible.
k0
POWER D E T E R M I N A T I O N F R O M
HEAT EXTRACTION
The most reliable method for determination o f
actual reactor power was that based on the energy
removed from the reactor i n the form of heat. The
inlet, outlet, and mean temperatures, the temperature d i f k r e n c e of the fuel across the reactor, and
the temperature difference o f the reflector coolant
were continually recorded. The rates o f flow of
both the fuel and the sodium were recorded and
could also be determined from the speeds of the
pumps. Since the heat capacities o f both the fuel
and the sodium were known, the power level o f the
reactor could be determined by the sum of the
values obtained from the following relations:
P,
‘N
a =
=
0.11 q , A T ,
0.0343 q N a A T N a
where
P , = power from fuel heat extraction (Mw),
p N a = power from sodium heat extraction (Mw),
68
q = volume flow rate (gpm),
AT = temperature gradient across reactor of
the fuel or the sodium
(OF).
Since both the fuel and the sodium were cooled
by helium passing across a liquid-to-helium heat
exchanger and the helium was cooled by passing
it over a helium-to-water heat exchanger, the reactor power could also be determined from the
water flow and temperature differences across the
he1ium-to-water heat exchangers. Since the fuelto-helium and sodium-to-he1 ium heat exchangers
are close t o their respective helium-to-water heat
exchangers, l i t t l e heat was l o s t by the helium and
therefore practically a l l of the heat removed from
the fuel and sodium was transferred t o the water.
The heat balance thus obtained was known as the
secondary heat balance, while that obtained directly
from the fuel and sodium systems was known as
the primary heat balance.
I n i t i a l comparisons of the primary and secondary
heat balances revealed discrepancies o f the order
of 50%. Subsequent investigation revealed that
the temperature drops o f the primary heat balance
were i n error. The temperatures were obtained from
a few thermocouples located on fuel and sodium
lines within the reactor thermal shield that read
considerably lower than several thermocouples
located external to the thermal shield on the fuel
and sodium lines to and from the reactor. T h i s
anomaly i s discussed further i n Appendix K. A l l
reactor i n l e t and e x i t temperatures were subsequently based on the external thermocouple readings.
A detailed analysis o f the extracted power was
made (App. L) for the 25-hr xenon experiment
primarily because of the certainty that equilibrium
conditions had been established; also, t h i s highpower run at more than 2 Mw contributed almost
two-thirds of the megawatt-hours logged during the
entire reactor operation. For t h i s run the primary
heat balance showed an extracted power o f 2.12
Mw, and the secondary heat balance gave an extracted power of 2.28 Mw. The difference between
the two determinations was 7%. Furthermore, for
t h i s run about 75% of the power was removed by
the fuel, about 24% by the sodium, and about 1%
by the rod cooling system.
Heat balances and, hence, power determinations
were made for a number of other experiments during
the high-power operation; these are tabulated in
Appendix L. One such heat balance that was of
particular interest was obtained during the maxi-
t
*~
L
A
L
t
1
5
-
c
i
t
c
f
t
I
0
L
t.
mum power run (Exp. H-14). For t h i s run the mean
temperature was raised t o 134OOF (to avoid a lowtemperature reverse), and the fuel and sodium
system blower speeds were increased t o their
respective maximums. At equilibrium the power
levels indicated by the primary and secondary heat
balances were 2.45 and 2.53 Mw, respectively.
The l i n e temperatures throughout the fuel and
sodium systems during t h i s experiment are shown
in Figs. 6.5 and 6.6, respectively. The temperatures indicated along the lines are the actual
thermocouple readings of each point at equilibrium.
It may be noted that the outlet fuel l i n e temperaThe temperature
ture averaged about 158OOF.
measured at the reactor was considerably lower,
as shown i n Fig. 6.7, which shows a portion of the
instrument panel during this experiment.
R E A C T O R KINETICS
4
1
1
?
i
3
r
A d i s t i n c t i v e control characteristic of any circulating-fuel reactor is that the reactor power i s
determined solely by that part of the system which
i s external to the reactor, i.e., the heat extraction
equipment.
In the power regime, control rods do
not appreciably influence the steady-state power
production, but the power extracted from the reactor
does influence the reactivity of the reactor and
hence renders the reactor controllable.
In order that such a system be acceptable as a
power reactor, it i s requisite that the reactor have
a negative temperature coefficient of reactivity.
One of the most gratifying results of the ARE
operation was the successful demonstration of i t s
large negative temperature coefficient.
This
temperature coefficient made it possible for the
reactor to maintain a balance between the power
extracted from the circulating fuel and coolant and
the power generated within the reactor. The temperature c y c l i n g of the fuel was the mechanism by
which equilibrium was maintained.
A thorough understanding of control processes
i n a circulating-fuel reactor with a negative temperature coefficient i s necessary for an appreciation
of the k i n e t i c behavior of such a power reactor i n
the power regime.
An important purpose of the
ARE was the observation o f the kinetic behavior
of the reactor under power coupling t o i t s load
when perturbations i n the reactivity were introduced. Transient conditions could be induced both
by control rod motion and by variation of the
external power load.
Information on the kinetic
behavior was obtained from a number o f experiments
that were conducted during the period of power
operation. These are described i n the following
sections. Preceding the discussion of these experiments i s a detailed qualitative description of the
temperature c y c l i n g of the circulating fuel which
i s inherent to a l l kinetic phenomena.
Reactor Control
by Temperature Coefficient
The control of a circulating-fuel reactor w i t h a
negative temperature coefficient can best be understood by following the course of a “slug” of fuel
as it traverses the ARE system. For a description
of the course of a slug, it i s assumed that at time
zero the system i s i n equilibrium, with isothermal
temperatures throughout; i n this condition no power
At time t the fuel system
i s being extracted.
It i s also assumed
helium blower i s turned on.
that at this time the slug of fuel under observation
i s just entering the heat exchanger. In passing
through the heat exchanger the slug i s cooled by
the he1ium blowing through the heat exchanger.
About 20 sec later the cooled slug enters the reactor and i s registered as a decrease on the i n l e t
temperature indicator.
Because it i s cooler than
the fuel it i s displacing, i t s density i s greater and
therefore the number of uranium atoms per unit
volume i s greater. T h i s results i n a greater fission
rate and, hence, a greater reactivity which, i n turn,
increases the power generated.
The temperature
of the slug rises as i t passes through the reactor
because of increased power generation. The r i s e
i n temperature of the slug results i n i t s expansion
and decrease in reactivity. This, i n turn, lowers
the rate of power generation. Eight seconds after
the slug enters the reactor it passes out into the
outlet l i n e and i s registered as an increase i n
temperature on the outlet fuel temperature indicator. I n 47 sec from the start of i t s journey i t i s
back at the heat exchanger to be cooled again.
The masses of fuel behind the i n i t i a l slug follow
the same pattern so that, since the fuel i s a continuous medium, the power generated i n the reactor3 w i l l r i s e u n t i l the increased reactivity due
t o the incoming fuel and decreased reactivity due
A t this
to the outgoing fuel attain a balance.
point equilibrium i s reached between power extracted by the helium and the power generated i n
~~
~
’With a power reactor i t i s appropriate to speak o f two
types o f power: the nuclear power, i.e., the total power
generated within the reactor; and the extracted or useful
power, i.e., the power removed from the fuel and coolant
b y cooling.
i
70
W
x
v)
P
i
t
I
i
I
i
i
(0
e!
f
i:
S
a
M
i
f
Fig.
6.7. Portion of Instrument Panel During Maximum Power Run. Experiment H-14.
the reactor. The reactor power can thus be varied
a t w i l l merely by changing the rate o f cooling, that
is, the demand for power. T h i s type of reactor i s
said t o be a slave t o the demand.
The nuclear power i s proportional t o the neutron
flux for any given c r i t i c a l concentration.
The
neutron flux i s conventionally measured by a
neutron detector, such as a L o g N chamber system,
which consists essentially o f a neutron detector
coupled t o an electronic system which has a
logarithmic-type of output signal. T h i s signal i s
proportional to the neutron flux (and, hence, the
power) level, and i s indicated and recorded on a
logarithmic scale.
The extracted power, on the other hand, i s proportional to the product o f either the fuel flow and
the temperature difference between the i n l e t and
outlet sides of the heat exchanger (or reactor) or
the secondary (or tertiary) coolant flow and i t s
temperature difference.
I n the ARE the fuel was
cooled by helium flowing across it in the fuel-to-
72
helium heat exchangers. The heat picked up by
the helium was then removed by water i n a heliumto-water heat exchanger. It was possible i o measure
the extracted power i n both the fuel loop and the
water loop, as explained i n a preceding section.
When no power i s being extracted from a power
reactor, the nuclear power i s determined by the
control rod and shim rod settings, and the reactor
i s controlled by some type of servo mechanism
which keeps the power a t some desired level.
When the reactor i s taken o f f the servo and allowed
t o respond t o the demand of the cooling produced
i n the heat exchanger, i t w i l l seek a power level
determined by the rate o f heat removal. Once an
equilibrium between nuclear and extracted power
has been established, changing the shim rod or
regulating rod settings merely changes the nuclear
power level but does not alter the amount o f power
being extracted.
T h i s change in nuclear power
results i n a raising or lowering of the reactor mean
temperature, and, once the temperature change has
-i
e
E
4
J
1
1
i
i
sl
i
t
been effected and equilibrium re-establ ished, the
nuclear power and the extracted power w i l l again
4
be equal.
The control o f the ARE by extracted power demand
i s illustrated i n Fig. 6.8, which shows the tracings
made by several o f the control room instrument
recorders during power operation between 1258 and
1329 on November 10, 1954. The reactor behavior
i s demonstrated by reading the tracings right t o
left, proceeding as follows: A t 1258 the reactor
was i n an equilibrium condition a t about a 50-kw
power level. A t 1259 the fuel system helium blower
was turned on and i t s speed was increased t o 500
rpm. The reactor power, as noted on the L o g N
chart, rose on a 10-sec period to slightly over 1
Mw, "overshot" i t s mark, fluctuated somewhat i n
the manner of a damped oscillator, and, finally,
some 4 min later, came to an equilibrium power of
but not immediately, after
around 900 kw.
the blower was turned on, the reactor fuel i n l e t
temperature began to drop and the fuel outlet temperature began t o rise, while the reactor mean
temperature began t o drop slightly. Each of these
temperatures showed the osci I latory phenomenon
noted with the L o g N recorder. The reason for the
drop i n the mean temperature was that both the
decrease and rate of decrease of the i n l e t temperature were greater than the corresponding increase
and rate of increase o f the outlet temperature. The
i n l e t temperature dropped from 1335 t o 12560F6 a t
a rate o f 1.54OF/sec and the outlet temperature
ros0 from 1405 t o 1475OF a t a rate o f 1.36OF/sec;
as a result the mean temperature f e l l from 1370 to
1 3 6 5 O F a t a rate o f about O.l0F/sec. The fact that
the mean temperature dropped (and this was a
characteristic phenomenon noted throughout the
power operation) rather than remaining constant
can be explained, at least in part, by the distortion
o f the flux patterns within the reactor. The highest
fluxes were toward the i n l e t fuel passages, and
the lowest fluxes were toward the outlet fuel
passages (cf., App. 0).
A t 1310 the blower speed was increased from 500
1
9
4 0 n several occasions t h e nuclear power, a s indicated
o n the L o g N recorder, rose temporarily above 2.5 Mw,
b u t the extracted power, l i m i t e d b y the system capacity
for removing heat, never exceeded 2.5 Mw.
5Errors i n marking the charts could conceivably account for a s much a s 1 / 2 rnin o f t h e time differences
noted. T h i s phenomenon o f time l a g i s discussed l a t e r
i n t h i s section.
6 T h e temperature readings given here and elsewhere
i n t h i s section have been corrected b y the method given
i n Appendix K.
t o 1000 rpm, and the power rose from 1 t o 1.8 Mw
on a period of 7.5 sec, with corresponding changes
of the inlet, outlet, and mean temperatures. Although the power increase i n t h i s case was comparable to the f i r s t power increase, the rates of r i s e
of both the power and temperatures were much
less, being on the order o f one-half as much for
the temperatures.
The effect o f withdrawing the regulating rod 3 in.
i s demonstrated with the next rise at 1312 i n
Fig. 6.8. Up to t h i s time the trace o f the regulating rod position was constant, since it was not
on servo. With the withdrawal of the rod the r i s e
of the nuclear power was immediate. After some
delay, a l l of the temperatures showed a rise.
Interestingly enough, the outlet temperature rose
16OF, the mean 12OF, and the i n l e t 9OF. T h i s
pattern, with the outlet temperature showing the
greatest change, was characteristic of the temperature changes incurred by shim rod and regulating
rod movement during power operation. After the
rod movement the temperatures leveled out a t the
higher values, but the nuclear power level slowly
drifted back t o the equilibrium position.
At 1321 the blower was shut off and the charts
show the result of the s h i f t toward isothermal (no
power) conditions.
Two minutes later, when the
nuclear power had decreased to about 600 kw, the
blower was again started and i t s speed was increased to 1500 rpm. The nuclear power level rose
i n i t i a l l y to 2.6 Mw, and i t again went through an
oscillatory c y c l e before settling down to 2.2 Mw.
The corresponding temperature oscillations were
again noted. These oscillations were o f sufficient
strength to determine an oscillation period o f about
2.25 min. An interesting observation a t t h i s point
was that the time lag o f temperature response for a
higher i n i t i a l power (600 k w compared with 50 kw)
was only about one-half the l a g during the f i r s t
r i s e to power a t 1259. T h i s phenomenon had been
noted previously in connection with the various
temperature coefficient experiments (cf., App. 0).
Table 6.3 shows the data obtained from the
nuclear and process instruments during the typical
operation period shown i n Fig. 6.8.
Startup on Demand for Power (Exp.
H-6)
T o illustrate how completely the reactor was a
slave to the load, an experiment was performed i n
which the reactor was brought to subcritical and
then taken to c r i t i c a l on temperature coefficient
(i.e., by the power demand and without use of
73
j
1
I
i
1
i
I
1
e
I
74
?
?
T A B L E 6.3. NUCLEAR AND PROCESS DATA OBTAINED DURING
A TYPICAL OPERATION PERIOD (NOV. 10)
Operation Performed
Type of
T y p e of
Data
F u e l System
F u e l System
Regulating
F u e l System
F u e l System
Mea suremen t
Obtained*
Blower Speed
Blower Speed
Rod
Blower Speed
Blower Speed
lncrea s i ng
Increasing
Withdrawn
Decreasing
Increasing
1300
13 10
13 12
1321
1323
p , , Mw
0.0566
0.99
1.58
1.56
0.57
p,, Mw
1.07
1.70
1.94
0.57
2.17
8P,
1.013
0.71
0.36
0.99
1.60
28
41
9
60
81
0.0363
0.0 173
0.04
0.016
0.198
10
75
33
63
54
1405
1485
1527
1546
1491
Time
L o g N power
St,
Mw
sec
6P/&,
Mw/sec
7; sec
i
Fuel outlet
OF
%,l’
temperature
1
J
To,2* O F
6To, O F
1474
1527
1543
1491
1566
69
42
16
-55
75
St,
45
56
24
75
58
1.54
0.75
0.67
-0.73
1.29
Till, O F
1335
1269
123 1
1239
13 14
Ti,,,OF
1256
1227
1240
1314
1208
8Til
-79
-42
9
75
- 106
58
53
42
75
88
- 1.36
-0.79
0.2 1
1.00
-1.20
sec
6To/6t,
*
1
Fuel inlet
temperature
OF/sec
OF
at, sec
6T/6t,
Fuel mean
temperature
OF/sec
Tm, 18
OF
1370
1377
1379
1388
1402
Tm,2n
O F
1365
1377
1391
1402
1387
-5
0
12
14
-15
33
75
73
ST,,
St,
O F
-0.096
0
0.36
0.19
-0.20
AT^, O F
70
2 16
296
287
177
AT,,
2 18
300
3 03
177
358
&AT), O F
148
84
7
-1 10
181
St, sec
52
54.5
33
75
73
2.87
1.54
0.21
-1.47
2.49
6Tm/6t,
,
Reactor f u e l
AT
52
sec
F/sec
OF
AT)/&, OF/sec
Regulating rod
movement
. . .
dl, in.
7.60
d,,
10.37
in.
75
TABLE 6.3
(continued)
Operation Performed
T y p e of
Data
F u e l System
F u e l System
Regulating
F u e l System
F u e l System
Obtained*
Blower Speed
Blower Speed
Rod
Blower Speed
Blower Speed
Increasing
Increasing
Withdrawn
Decreasing
Increasing
Type of
Measurement
*
Regulating rad
movement
ad,, in.
2.77
81,
9
sec
ad/&,
F u e l system h e l i u m
blower speed
J
0.32
in./sec
Ak/k, %
0.091 4
( A k / k ) / a t , %/sec
0.01 1
s i , rpm
0
500
1000
0
so, rpm
500
1000
0
1500
8s. rpm
500
500
-1000
1500
at, sec
14
14
60
42
&-/at, rpm/sec
36
36
-17
36
* T h e f o l l o w i n g symbols are used:
Quantity
Subscript
P power
t
time
7 period
T temperature
d position
k m u l t i p l i c a t i o n factdr
speed
change i n quantity
b difference between t w o quantities
1 initial condition
2 final condition
o
i
m
outlet condition
inlet condition
mean
8
The a t ' s are the t i m e s required t a go from the i n i t i a l c o n d i t i o n t o the f i n a l condition a t the greatest observed rate of
change. T h e temperatures quoted have a l l been corrected by the method described i n Appendix K.
rods). The experiment consisted o f two runs. In
the f i r s t run the reactor was taken directly from
100 kw to slightly over 2 Mw. I n the second run
the reactor was taken to power from a subcritical
condition.
The progress o f these experiments, as recorded
by the thermocouple recorders on the i n l e t and
outlet sides o f the six individual fuel tubes, i s
shown i n Fig. 6.9. Reading from right to left, the
reactor was a t 200 k w power a t 2300 on November
9 a t the beginning o f the experiment. The fuel
system helium blower speed was increased to 1700
rpm slowly, starting a t 2307. Ten minutes later
the sodium system coolant blowers were turned on.
A s the power rose to about 2.5 Mw, a temperature
difference o f some 32OOF appeared across the reactor tubes.
T h i s i s shown i n Fig. 6.9 by the
parting o f the i n l e t and outlet temperature indications o f the reactor fuel tubes. T h i s temperature
76
difference remained constant u n t i l the blower speed
was reduced o t 2340. A t 2345 the i n l e t and outlet
tube temperatures were nearly the same again at
200 kw power. A sharp drop in the tube ternperatures a t t h i s point corresponded to the insertion
o f the shim rods.
The sharp increase i n the temperature differential
across the fubes a t 2353 was the result of a steep
rise i n power when the blower speed was increased
from 0 to 1700 rpm. A t 0005 on November 10, the
regulating rod was f i r s t entirely inserted (from
7 in. withdrawn) and then fully withdrawn. T h i s
motion was followed by a drop and then a sharp
r i s e i n the fuel tube temperatures.
T h i s action
marked the end o f Run 1.
During Run 2 the reactor was brought subcritical
by turning o f f the fuel system helium blower and
inserting the regulating rod, starting a t 0016. The
sodium system helium blowers had been operating
i
4
1
.1
4
i
1
?
4
d
:
0
0
z
0
0
a
v
w
0
0
r- -
0
0
-
Io
c
cp
0
0
0
0
-
t
0
0
0
0
0
c
0
w-
stroo
OPOO
SEOO
OEOO
SZOO
0200
GI00
0100
so00
OOPZ
OP€Z
SZEZ
OZEZ
S IEZ
j GOEZ
0
0
0
c
w
In
5)
ai
U
In
!2
c
6
>
0
z
C
t
U
.-al
.al
Y
d
L
0
Q
t
s
e
Q
x
I0
n
.
C
t
0
C
s
al
c
t
L
.-E
n
X
w
0
-0
C
al
K
t
--
0
.-bs
al
s
m
77
some of the time during Run 1 t o extract power from
the sodium. A s soon as the reactor became subc r i t i c a l the operation of the sodium system blowers,
which were then removing about 400 kw of power
from the sodium, made the reactor go c r i t i c a l again
and stabalize at the 400 kw level w i t h i n 5 min. It i s
interesting t o note that because the mean sodium
temperature was some 65OF lower than the mean fuel
temperature, the fuel temperatures dropped 65OF
during this time from a mean temperature o f 1395 t o
133OoF, even though the nuclear power was inA t 0026 the fuel system blower was
creasing.
turned on again and the nuclear power temporarily
rose t o 3 M w . ~ Two minutes after the fuel system
blower was turned on, the fuel heat exchanger
outlet temperature went below 1158OF and the
automatic heat-exchanger-temperature interlock reduced the blower speed until the heat exchanger
outlet (reactor inlet) temperature had risen above
1150°F. A t 0030 the reactor power had leveled
out a t 1 Mw. It hod thus been successfully demonstrated that the power demand would bring the
reactor critical; therefore, a t 0033, the fuel and
sodium system blowers were turned o f f and the
experiment brought t o an end.
E f f e c t of O n e D o l l a r of Reactivity (Exp.
H-8)
One o f the objectives of the ARE was the observation o f effects of introducing excess Ak into the
reactor during high-power operation.
The introduction o f the excess Ak could be most easily
accomplished by movement o f the regulating rod,
which was worth 0.4% excess k (one dollar of
reactivity).
One experiment consisted merely of
withdrawing the regulating rod from an entirely
inserted position, and recording or noting the
effects and, then, after equilibrium had been established, inserting the rod to i t s original position.
Since the regulating rod had a travel rate of
0.32 in./sec, reactivity could be introduced a t the
Fig. 6.10 shows the
rate of 0.011% Ak/k-sec.
history o f the experiment.
A t 1110 on November 10, with the reactor a t an
i n i t i a l power o f 2.2 Mw, the regulating rod was
withdrawn i t s f u l l 12 in. of travel. The reactor
went on an observed 42-sec period7 u n t i l the
nuclear power was 3.9 Mw 35-sec later. The re'This period was measured from the L o g N recorder
trace.
The period recorded on the period meter w a s
partially a time rate o f change of period since & / k
was
not constant but w a s increasing during this time.
A
discussion of this i s given i n Appendix S.
78
actor fuel temperatures rose an average o f 45OF i n
Once the rod withdrawal had been completed, the fuel i n l e t and outlet temperatures
leveled o f f a t new values. The extracted power
rose a few per cent when the rod was withdrawn,
mainly because, with the higher fuel mean temperature, the rate o f cooling and, hence, the rate of
heat removal i n the heat exchanger was greater.
Since the extracted power rose only slightly during
t h i s time, the nuclear power slowly drifted downward from i t s peak o f 3.9 Mw to 2.9 Mw, as can be
seen i n Fig. 6.10. Note that the mean temperature
continued i t s upward trend.
Three minutes later
the rod was completely inserted.
The nuciear
7
power decreased from 2.9 to 1.4 Mw on a 47-sec
per:od and then slowly drifted back to 2.2 Mw, the
starting power.
The fuel temperatures then returned to values slightly higher than their original
A characteristic behavior o f the outlet
values.
temperature was again noted during t h i s experiment,
namely, that movement o f a rod affected the outlet
temperature most and the i n l e t temperature the
least. Additional data on t h i s experiment are given
i n Table 6.4.
Again, with this experiment, a time lag was
noted; the i n l e t temperature lagged behind the outl e t temperature by about the fuel transit time o f
47 sec. A reasonable explanation for t h i s lag i s
that the time required for the fuel i n the reactor a t
the start o f the experiment to affect the i n l e t l i n e
thermocouple i s the transit time through the system.
Because o f t h i s difference i n response of the
thermocouples on the i n l e t and outlet fuel lines
when the rod was withdrawn, the temperature
differential across the reactor rose from 341OF to
a peak value of 395OF i n about 1 min and then
dropped t o a more or less constant value o f 355OF,
which was 14OF higher than at the start.
This
means that the extracted power rose slightly, about
8%. Examination o f the tracings o f the fuel and
sodium heat exchanger cooling water recorders for
t h i s time indicated that the fuel power extraction
went up 6% and the sodium power extraction roughly
2%. The rise i n power extraction was due mainly
to the higher mean temperatures and thus the higher
cooling rates attained. The extracted power rose
to about 2.4 Mw at the new equilibrium. The difference (500 kw) between t h i s value and 2.9 Mw,
as shown by the L o g N recorder, went to heating
the reactor. T h i s was evidenced not only by the
continuing increase in reactor mean fuel temperature but also by the r i s i n g travel i n i n l e t and outlet
49 sec.
i
i
t
L
L
t
i
r
c
::
L
t
-
c
k
E
-REGULATING
ROO BEING
ROD
FULLY INSERTED
u.
I
o, 4600
ROD FJL-Y
WITHORAWN
_
I
_
I
II
I
ROO BEchG
ROD FLLLY
W~THORAWN~NSERTED
__
NSERTEO
_.,
1650
ORNL-LR-DWG 6557
ROO ACTION
I
I
I
I
I
I
I
I
REACTOR FUEL OUTLET TEMPERATURE-,
W
d 5 1550
2%
8 8 1500
1
5$
W
4450
$1400
3
0
I
r
J
i; 300
e
i
.k
275
8
a 2
250
K Q
J
w”
225
2
200
(75
I
I
I
I
I
I
I
L
I
I
4
-----+-I
I
I
1
7
I
J
I
I
I
1116
1115
I
/REACTOR
I
T
1
I
c
[ II
I
I I
I
I
I
I
I
PERIOD
I
I
ch
+I
r , \I
1
Y
I
I
I
I
I
I
I
I
I
0.2
1117
I
1114
-TIME
1113
1112
1111
1f10
4409
OF DAY, NOV. 10, 8 5 4
,
Fig. 6.10.
4
i
K i n e t i c Behavior of Reactor Upon Introduction of a
Dollar of Reactivity.
T A B L E 6.4 NUCLEAR AND PROCESS DATA OBTAINED DURING EXPERIMENT H-8 FOR
DETERMINING THE E F F E C T O F ONE DOLLAR OF REACTIVITY (NOV. 10)
ODera t ion Performed
Type of Data
Type o f Measurement
Regulating
Regu l a t i ng
Rod Withdrawn
Rod Inserted
1110
1114
2.34
2.92
3.94
1.41
1.60
1.51
35
35
0.046
0.043
42
47
1573
1620
T0,2* O F
1623
1572
8To, O F
50
-48
6t,
41
41
1.22
-1.17
Tj,,,O F
1217
1258
Ti,2, O F
1257
1221
BTi, O F
40
-3 7
at, sec
57
56
0.70
- 0.66
OF
1395
1439
OF
1440
1396
45
-4 3
49
49
0.92
-0.88
OF
356
362
AT^, O F
366
351
&an,
10
-1 1
Si, sec
49
49
0.20
-0.2 1
dl, in.
2.o
14.0
d2, in.
14.0
2.0
6d.
12.0
-12.0
0btai ned*
Time
Log
power
Pi, Mw
Pqr Mw
8P,
.
Mw
at, sec
6P/6t,
5
F u e l o u t l e t temperature
Mw/sec
sec
To,l'
OF
sec
6To/6t,
F u e l i n l e t temperature
'F/sec
6Ti/6t,
F u e l mean temperature
F/sec
Trn,1'
Trn,2'
ST
,,
St,
OF
sec
6Tm/6t,
Reactor fuel
AT
AT,,
OF/sec
O F
6 6 T )/st,
Regulating rod movement
*See footnote to T a b l e
80
6.3.
in.
F/sec
k
€
1
TABLE 6.4 (continued)
Type of Data
Type o f Measurement
Regulating rod
movement
Obtained*
37
37
0.32
-0.32
hk/k, %
0.4
0.4
(hk/k)/&,W s e c
0.01 1
-0.11
rpm
1750
1750
5 2 , rpm
1750
1750
6s,
rpm
0
0
at,
sec
-
-
0
0
at, sec
6d/6t,
d
Fuel system helium blower speed
Operation Performed
_______~
Regulating
Regulating
Rod Withdrawn
Rod Inserted
SI,
in./sec
6s/6t, rpm/sec
_ _ ~
~~
*See
footnote
to
Table 6.3.
fuel temperatures just before the rod was inserted
a t 11 14.
a
,
d
I
i
4
1
d
Power C y c l i n g of Reactor
The l a s t day o f operation o f the ARE, November
12, was devoted t o maximum power runs and to
demonstrations of the reactor operation to visitors.
The complete history o f the reactor operation during
that time i s shown i n Fig. 6.11. Chart A shows
the trace o f the sodium temperature differential
across the reactor and chart €3 shows the trace of
the outlet temperature o f the water from the fuel
heat exchangers. Clearly seen i n chart A are the
three maximum power runs a t 1O:OO AM, 12:30 and
7:45 P M . Both charts show the cycling o f the
extracted power that the system underwent, but
the cycles are especially well defined i n chart B,
which shows the heat exchanger water temperatures
varying from a low of 65OF to a high o f 140°F,
corresponding t o total extracted powers of from
around 100 k w to over 2 Mw. I n the 12 hr between
8:OO AM and 8:OO PM, 24 complete cycles were
recorded. I n the afternoon hours between 1:00 and
4:OO PM, when the most visitors were present, the
reactor was cycled an average o f 3 times per hour.
Of interest to the visitors were the tracings of the
reactor fuel tube AT recorders. The photograph
reproduced as Fig. 6.12, which was taken during
one of the cycles, shows each of the six fuel tube
AT recorders simultaneously tracing out the same
power cycle pattern. The picture was taken about
12:30 P M during one o f the maximum power runs.
The recorders show AT’S of around 250OF. The
actual fuel tube differences a t t h i s time were more
nearly 355OF.
Additional information on the power cycling may
be obtained from Fig. 6.13 which shows the three
temperature cycles recorded i n the hour between
1 : l O and 2:lO P M on the l a s t day. Shown i n this
figure are traces from the s i x fuel tube AT recorders, a time condensation* of the micromicroammeter recorder (which traced the power), one of
the heat exchanger outlet temperature recorders,
and the individual fuel tube temperature recorder.
Referring to the rnicromicroarnrneter trace and
reading from right t o l e f t starting at 1317, the
events are elaborated for the f i r s t cycle. The other
two cycles are similar.
A t 1317 the fuel system helium blower was turned
off, and the reactor power was allowed to decrease
from a level of 1.8 Mw. Eight minutes later the
power had fallen t o about 130 kw, and the blower
was turned on t o 1700 rpm. The power rose quickly
to 2.3 Mw, at which point a low-heat-exchangertemperature blower reverse occurred; that is,
when the he1ium cooling lowered the heat exchanger
outlet fuel temperature to 115OOF (as noted above),
a relay was automatically actuated that decreased
the helium blower speed until such time as the
* T h i s condensation was necessary because the o r i g i chart ran a t a speed o f 80 in./hr, w h i l e charts u s e d
subsequently ran a t a speed o f o n l y 4 or 5 in./hr.
nal
ai
4
82
x
'6
6
d
c
c
t
k
*
i
P
I
L
L
i
...
i
k
Fig. 6.12.
Reactor F u e l Tube AT'S During High-Power Operation.
outlet temperature rose to above 115OOF. During
t h i s time the power fell to 1.1 Mw. A t 1328 a
t s
permit" signal was given and the power was
again increased. When 2 Mw was reached the reactor was leveled o f f on extracted power and far
10 min the effect o f regulating rod movement was
demonstrated to visitors.
The maximum nuclear power attained during the
first c y c l e was 2.8 Mw4 at 1340. During the third
cycle at 1410 a nuclear power o f 3.4 Mw was
reached.
The s i x fuel tube AT recorders (Fig. 6.12) followed exactly the same pattern as the micromicroIt should be noted, however,
ammeter recorder.
that the actual AT'S were of the order of 100
degrees higher than those shown on the charts.
A t the high power peaks of the third cycle the fuel
outlet l i n e temperatures, as read in the basement,
The i n l e t
were indicated to be about 1625OF.
temperatures ranged from an actual (vs observed)
high of 1 3 4 O O F to a low of 1150OF during the lowheat-exchanger-temperature blower reverse.
The c h m g e i n tube AT'S as a r e s u l t of rod movement was due partly t o the phenomenon of the t i m e
lag between the i n l e t and outlet thermocouples.
Had there been no lag the effect of rad motion on
these couples would have been small.
The fuel heat exchanger outlet temperatures
showed the temperature cycles i n reverse.
The
f i r s t high peak corresponds to the condition at
1300°F w i t h the blower off. A s the blower was
turned on the temperature decreased t o 115OOF,
a t which point, as noted, the blower reverse occurred with the resultant decrease i n blower speed.
The heat exchanger outlet temperature then rose
t o 125OOF before the blower speed was allowed to
increase again.
The data from these and other recorder charts
during t h i s time, together with calculated rates
of change, are given i n Table 6.5.
83
E
a
-
m,
-
0
0
f
0
2
'0
w0
,
0
N
W
,
P
0
8
8
- 0
e
0
0
0
ul
0
cn
;
0
PF)
AJ
(OF)
AT YF)
AT
N
0
0
0
N
0
s
aI
W
T
I
0
m
I
s
I
/
N
8
0
01
N
8
0
TEMPERATURE (OF)
I
6
--
0
a
0
A
5
VI
0
VI
*
0
5
C-
z0
8
W
0
cn
W
0
VI
W
0
d
0
2
%
W
-
:
0
0
0
ul
8
T
I
0
W
I
I
0
e
O
c
n
I
I
0
VI
I
.,
0
0
I
2
0
m
-
(OF1
0
u)
0
0
N
I
I
I
0
0
Z
G
0
0
S
0
0
TEMPERATURE (OF)
i0 ;
0
N
VI
0
0
VI
N
i
I
0
/
>
SET BACK
0
s
s
--
AT
-___
0
VI
t
0
0
<0
0
0
I
I
0
d
~ - 1
0
0
0
5
SPEED INCREASING
TEMP. BLOWER REVERSE
HEAT EXCHANGER
,
BLOWER OFF ((317)
-BLOWER
N-'LOW
I
0
m
MICROMICROAMMETER READING (ARBITRARY SCALE)
-
0
0
W
W
0
0
0
W
0
0
W
0
0
P
zN
-
I
8
W
0
cn
W
i
r!
h
K
K
W
0
iU
4
LL
0
0
f
+
-I
U
U
K
4
I4
n
v)
W
Vr-
n
K
U
0
-I
U
3
Z
!-
U
2
K
W
iU
4
K
4
X
U
2
W
rn
.J
4
I-
P
W
0
f
t
0
N
n ! m
-
9
0
I - d
v)
=r
0
h
0
h
-2
m
Q
I - m
0 : W
rg
ZI hm
0
g N
h
-
' 9 -
h
I
'0
0
z
0
v)
s
d
0.
2
00
2
d
rg
Q
2
7
t
f
i
tem characteristics against the i n i t i a l power, since
t h i s represents the behavior from a given i n i t i a l
condition o f the reactor.
One of the properties investigated i n the above
manner was the reactor period. Some of the observed reactor periods are plotted as a function o f
Since the reactor
the i n i t i a l power in Fig.,6.14.
could be put on any period from i n f i n i t e to a given
minimum, the points in Fig. 6.14 are scattered.
Most of the longer periods represent periods observed during such times as the i n i t i a l rise t o
As
power, which was approached with caution.
the operators became more familiar with the operation a t power, deliberate attempts were made t o
see how fast a period the reactor would attain due
t o blower operation or rod movement. A s a result,
definite lower l i m i t s were established beyond
which reactor periods could not be induced by the
controls available t o the operator. The solid l i n e
in Fig. 6.14 represents the lower l i m i t for periods
during blower operation and the dashed l i n e i s that
corresponding to shim and regulating rod move~nent.~
The smallest period observed during highpower operation was the 10-sec period represented
by the rise i n the L o g N chamber a t 1259, Fig.
6.8. A few smaller periods were observed in the
very low-power regime just above critical.
The
two lines shown i n the figure indicate that the
lower the power the greater the transient that can
Reactor Transients
;
?
*
One of the characteristics o f the reactor that was
most important from the operator’s point of view
was the behavior under transient conditions, because i t i s only through transient effects that the
operator acquires a “feel” for the way the reactor
responds to his control under varying conditions.
The various ways in which transients may be
introduced into the reactor are the following:
1.
2.
3.
4.
i
operation of fuel system helium blower,
operation of sodium system helium blowers,
operation of rod cooling system helium blowers,
changing the fuel flow rate,
5. changing the sodium flow rate,
6. movement of shim rods,
7. movement of regulating rods.
A discussion of the effects on the ARE of some
of these operations i s given i n succeeding paragraphs. Effects 3 and 5 were very small although
observable.
Effect 4 was not observed a t high
power because o f the danger of freezing fuel i n the
pump, although it can be calculated from the inhour
curves (Fig. 5.11) and the known rate of change
of fuel pump speed.
Table 6.6 l i s t s calculated
rates of interjection of A k / k into the reactor by
the various means listed above.
It has already been pointed out that the ARE
reacfor and system were very sluggish in responding
t o demands at high power. Furthermore, it was
observed that the response a t low power (less than
100 kw) was much different from i t s response at
higher powers (at 1 Mw or greater). One way to
examine the behavior of the reactor i s to p l o t sysT A B L E 6.6.
91f the reactor hod been token to power w i t h the shim
rods inserted to greater depths, smaller eriods would
hove been observed for shim rod motion tRan are represented b y the dashed line, since t h e r e a c t i v i t y value
o f the rods would h a v e been greater for those insertions.
CALCULATED RATES O F INCREASE O F
hWk
IN ARE FROM VARIOUS OPERATIONS
i
Operotion
F u e l system helium blower speed increase
Regulating rod movement
Shim rod movement, single rod
Range
0 to 1500 rpm
1000 t o 1500 rpm
0.0035
Whole rod insertion or withdrowal
0.01 1
At
At
Shim rod movement, group operotion
(%Ak/k)/ s e c
4 in. insertion
8 in. insertion
A t 4 in. insertion
At
8 in. insertion
F u e l flow rate
0 to 46 gpm
Sodium flow rote
0 to 150 gpm
0.001 1
0.0039
0.0067
0.01 2
0.019
0.013
1 x
87
ORNL-LR-DWG 6 5 6 0
t
t
"i
P PERIODS RESULTING FROM FUEL SYSTEM
e:
HELIUM BLOWER OPERATION
PERIODS RESULTING FROM ROD MOVEMENT
__
-~
__
_ _
_.
~
FASTEST PERIODS
/'
6 FASTEST PERIODS
i;
FROM SHIM ROD AND
REGULATING ROD MOVEMENT
_ _ _ _.~
-
t
0.05
3{
0 2
05
10
2
5
12
INITIAL EXTRACTED POWER (Mw)
Fig.
6.14.
Induced Reactor Periods as a Function of I n i t i a l Extracted Power.
*
be introduced into the reactor, i.e., the higher the
power the harder it becomes t o introduce a transient
condition.
Extrapolation of the solid l i n e t o a
power o f 1 k w indicates that at t h i s power reactor
periods of 1 sec or less could have been introduced by blower operation.
It should be pointed out that periods could be
observed which were not true periods but were a
combination of the observed period and i t s time
derivative.
Whenever a sudden reactivity change
occurred, such as when a rod was moved slightly,
the period meter would register a transient peak
period that was very small compared with the true
period. The beginning peak i n the 'period curves
of Fig. 5.4 are of t h i s nature. Whenever a continuing change o f A k / k was taking place, the observed period was not a true period but, again, a
combination o f the true period and i t s time derivative. In Fig. 6.10 the period meter did not register
a constant period when the regulating rod was
moved but, rather, a continually changing period.
Also, it i s to be noted that the L o g N power recorder did not register straight slopes during the
power changes but, instead, constantly changing
ones, which again indicated that the period was
changing (cf., App. S).
The time behavior of the fuel temperatures as a
result of blower operation was obtained by plotting
i n l e t and outlet fuel temperature time rates o f
change against the i n i t i a l power. These plots are
shown i n Fig. 6.15. Larger rates o f change o f fuel
temperatures were observed a t lower powers, and
they corresponded to the smal ler periods observed.
T h e greatest rates of change of the fuel temperatures were observed a t an i n i t i a l power of 280 kw,
a t which time the i n l e t temperature changed a t a
rate o f -2.75'F/sec
and the outlet temperature
The only reason higher
a t a rate of +2.6'F/sec.
rates of change were not observed at powers
lower than t h i s was that in the low-power regime
the kinetic behavior was not well known and therefore a conservative approach t o transient conditions was used.
The rates of decrease of temperatures when the
blower was turned o f f and the reactor was allowed
to come to a lower power are also plotted i n Fig.
6.15. The trend of rates o f change of the temperatures seems to be reversed.
b
p
i
k
k
c
14
E
L
L
-
?
i
i
I;
c
I
1
i'
ORNL-LR-DWG
4
65<
3
,..
1
, .
1
2
0
a
Y)
,
0
2
e
W
a
z
1
0
I
W
Lz
2
I
2
0
W
E
Iw
-1
0
A
c
W
LL
-2
f
J
i
-3
FUEL OUTLET TEMPERATURE RATE, HELIUM BLOWER SPEED DECREASED
A FUEL INLET TEMPERATURE RATE, HELIUM BLOWER SPEED DECREASED
-4
0
0 5
( 0
15
2 0
2 5
INITIAL EXTRACTED POWER (Mw)
Fig.
d
d
!
1
rc
6.15.
Time Behavior of Reactor F u e l Temperatures as a Function of I n i t i a l Extracted Power.
Some very general observations on the transient
behavior o f the ARE reactor are summarized below.
Reactor Periods. I n general, the lower the i n i t i a l
power the faster was the transient introduced by
any type of operation t h a t resulted i n upsetting the
equi I ibrium between nuclear and extracted power.
The smaller periods were associated with lower
i n i t i a l powers.
The characterO s c i l l a t i o n of Reactor Power.
i s t i c behavior of the nuclear power during an estabIished transient was to "over-shoot" and then
oscillate around a mean value before settling down
t o that power. The period o f the oscillation appeared to be of the order of 2 min and lasted for
about 2 cycles.
Movement of Rods. The reactor was always more
prompt t o respond to rod motion than it was t o
blower operation. Also the reactor periods observed
at a given power were smaller due to rod movement
than those due t o blower operation.
Nuclear Power.
Whenever blower speed was
changed the nuclear power rose (with oscillations
as noted above) t o a higher power level than the
corresponding extracted power demand.
When a
rod movement occurred, the nuclear power changed
considerably, which resulted in higher or lower
reactor over-all temperatures, and then leveled out
to a new balance with the extracted power a t about
the original power level.
Extracted power was essenExtracted Power.
t i a l l y a function only of the demand and changed
due t o a change in operation of one of the various
heat exchanger blowers. During rod movement a t
a given demand power, the extracted power remained essentially constant. The unbalance between extracted and nuclear power resulted in a
change o f the mean reactor temperature.
Reactor Outlet Fuel Temperature.
The outlet
fuel temperature always rose and f e l l in the same
direction as the nuclear power level. For blower
operation the rise i n the outlet temperature was
always less than the f a l l o f the i n l e t temperature.
The opposite was true for regulating rod movement.
Reactor I n l e t Fuel Temperature. The i n l e t fuel
4
I
4
89
P
temperature rose and f e l l i n the opposite direction
to the change i n nuclear power during blower operation and i n the same direction as the nuclear
power during rod movements. The i n l e t temperature change was greater than the outlet temperature
change for blower operation.
The opposite was
true for shim rod operation.
Reactor Mean Temperature.
The reactor mean
temperature followed the trend of the reactor i n l e t
temperatures i n a l l cases, except that the changes
and rates of change were much smaller.
T i m e Lags. The system was very sluggish because of the long tansit time (47 sec) of the fuel.
I n addition to the sluggishness of the system, there
appeared to be time lags between the responses of
various temperature indicating instruments for the
same action. These lags were of the order o f 2
min for low-power operation (less than 100 kw) and
of the order of 1 min for full-power operation i n the
megawatt range.
The topic of time lags i s discussed i n greater detail i n a following section.
Also the mean reactor fuel temperature i s given by
+
= - (1T o
,
Ti)
2
Tm
and
.
-
Ti
-
2Tm ,
A T = To
From t h i s it i s seen that
AT = 2To
t
and
P = 2kq(To
-
Tm)
.
Now consider a change AP in the power:
AP
= 2kq(ATo
-
,
ATrn)
but the temperature coefficient r e a c t i v i t y
a
=-
a
is
Ak/k
i
‘Tm
11
so that
h
t
Calculated Power Change Resulting from a
Since
Regulating Rod Movement
Ak/k
A formula was developed which would estimate
a power change i n the reactor from a given change
i n the reactor outlet temperature and a regulating
rod movement.
T h i s was possible primarily because of the relationships e x i s t i n g between PI
AT, .and Ak/k. The development of the relationship i s described below.
= [(Ak/k)/in.],,
,
x Ad,
where
Ad,
[(Ak/k)/in.I,,
= regulating rod movement,
=
c
0.033% in. for the regulating
k
rod,
by putting i n the experimental values,”
that
i t i s seen
L
r
AP
=
=
2(0.11)(46)
(10.12 ATo
L
1
ATo
1
0.72
+ 47.75
Ad) x
When equilibrium e x i s t s between extracted power
and nuclear power, the power level i s given by
P = k q AT = k q (To
-
Ti)
,
where
k
I
= a constant (specific heat o f the fuel),
q = fuel flow (gpm),
AT = the difference between i n l e t and outlet
fuel temperatures,
T
= outlet fuel temperature,
Ti = i n l e t fuel temperature.
90
9.8
10-~
Mw
.
The change i n power, as calculated from t h i s
formula, checked closely the observed power change
for several different cases which were examined.
The calculated data are given i n Table 6.7.
b-
I
f
?
L
Reactor Temperature Differential as a Function
o f Helium Blower Speed
A relationship between the fuel system helium
blower speed and the resulting fuel AT was obtained from the data taken during ARE operation.
”The
factor 0.72 is u s e d to correct t h e temperatures
i n the manner described in Appendix K.
L
t
z
k
L
I
T A B L E 6.7.
SOME CALCULATED AND OBSERVED POWER CHANGES FROM REGULATING ROD MOTION
I
Observed Outlet
Date
AT^
Regulating Rod
C a l c u l a t e d Power
Observed Power
Movement
Change
Change
(OF)
(in.)
(Mw)
(Mw)
1110*
55
12
1.59
1.60
1 1 14*
53
12
1.55
1.51
1321
13
0.37
0.36
Time
Temperature Change
~
Nov.
10
*Data recorded during Exp.
2.8
H-8.
1
Some of the experimental fuel AT’S are plotted in
Fig. 6.16 as a function of their corresponding
blower speeds. These points could be represented
by a parabola of the form
AT^
=
K~
ORNL-LR-DWG
6562
t
,
i
where
K
!
= a constant of proportionality,
s = the blower speed.
i
The curve that f i t the experimental points best was
found t o be
AT
=
( 0 . 7 5 ~ ) ” ~,
where AT i s expressed i n hundreds of degrees F
and s i s given i n hundreds of rpm.
The extracted power can now be expressed i n
terms o f blower speed:
0
400
800
I200
1600
2000
FUEL SYSTEM HELIUM BLOWER SPEED ( r p m )
/-
Reactor
F i g . 6.16.
Helium Blower Speed.
P = k q AT = k q ( 0 . 7 5 ~ ) ” ~
46
i1
( 0 . 7 5 ~ p ’ ~ i io2
=
[OJI
=
0.506 ( 0 . 7 5 ~ ) ”Mw
~
a Function of F u e l
10-~
.
lo2
The factor
i s used because AT i s expressed
i n the empirical formula i n hundreds of degrees.
The
factor expresses power i n megawatts.
The results obtained with the formula checked
with the observed values of power in numerous
cases. As an example, during experiment H-3, for
a blower speed of 1590 rpm, an extracted power
of 1.75 Mw was observed. The formula gives
P = 0.506r0.75 (15.9)1’12 = 1.75 Mw
AT a s
.
The Phenomenon o f the Time L a g
One of the most noteworthy phenomena enc0u.ntered during the experiment was the time lag that
seemed to be an i n t r i n s i c part of the system behavior. lt was remarkably demonstrated in Fig. 6.8
when the blower was turned on. It was 0.75 min
late: that the i n l e t temperature records showed a
response, about 1.5 min before the mean temperature hegan dropping, and almost 2 min before the
outlet temperature showed a corresponding rise.
The reasons for the lags are obscure, but i n an
attempt to analyze the phenomenon several things
must be considered: the extensiveness of the system, the reactor geometry, i.e., location of i n l e t
and outlet fuel passages w i t h respect to reactor
core, heat transfer properties and possible phenemona w i t h i n reactor, location, design, and attachment of the thermocoupl es within the reactor, and
other unknown factors. There i s a good p o s s i b i l i t y
that the time lag phenomenon was connected w i t h
the temperature discrepancies noted in Appendix K.
Errors i n marking the charts could possibly
account for as much as 0.5 min, but certainly this
cannot wholly account for the lag which was noted
91
L
I
i
consistently throughout the progress of the experiment.
There i s no doubt the system was v&
sluggish and slow t o respond. T h i s sluggishness
was undoubtedly due, i n large part, t o the length
of the fuel system. The transit time i n the system
was about 47 sec a t full pump speed, as observed
during the c r i t i c a l experiments.
There does not
seem t o be any convenient mechapism for explaining longer time lags. A partial explanation
i s advanced i n Appendix 0, where geometrical
considerations show that the instantaneous temperature change of the fuel was greater near the center
o f the reactor than a t the points o f the thermocouple locations.
Thus the nuclear power was
observed t o change before the i n l e t and outlet
thermocouples indicated changes.
An attempt was also made t o analyze the heat
transfer conditions within the reactor. Since the
fuel and sodium flows were i n opposite directions
w i h i n the reactor, the fuel i n l e t l i n e thermocouples
were located near sodium outlet lines and the fuel
outlet l i n e thermocouples were located near sodium
i n l e t lines. The temperature differences existing
between the sodium (and moderator) and the fuel
near the i n l e t and outlet thermocouples were o f the
order o f 100°F or more and could possibly have
influenced the readings o f the thermocouples. T h i s
phenomena was investigated,"
and no delay approaching 2 min could be ascribed to the heat
transfer characteristics o f the system. The greatest
time lag which could be found in the heat transfer
mechanism was o f the order o f a few seconds. The
time lags o f the reactor AT and mean temperature
thermocouples were also investigated because
these thermocouples had been insulated from the
metal o f the system. The time lags found were o f
the order o f 15 sec or less, and thus they were
about the same as those for uninsulated thermocouples.
An experimental check showed that
helium from the rod-cooling system had no appreciable effect on the readings o f fuel temperatures
when the blower speed was changed. Therefore,
the phenomena o f lag and low temperature readings
could not be attributed t o t h i s cause.
It may well be that what may be described as
e1
thermal inertia" played a part in the time lag.
Since the reactor as a whole did not respond to a
temperature change very fast, even though the
fuel did, the thermocouple readings could have
been affected to the extent that they received heat
"H.
neeri ng.
92
F.
Poppendiek,
Reactor
Experimental
Engi-
not only from the fuel tubes but from other parts
o f the reactor. Better design and location o f the
thermocouples might have prevented such time lags.
Reactivity Effects of Transients i n the
Sodium System
The introduction of transient conditions i n t o the
sodium system was expected t o have only small
effects on reactor behavior. T h i s proved t o be the
case i n the few instances for which data are available.
The two operations of the sodium system
which were observed t o have effects on reactivity
were changing the rate o f sodium flow and changing
the rate o f sodium cooling.
The f i r s t effect was observed as a part o f one o f
the low-power experiments (Exp. L-7), i n which,
with the reactor on servo a t 1-w power, the fuel
flow rate was varied and the change in the regulating rod position was noted.
The concluding
portion of the experiment consisted in stopping the
sodium coolant flow and observing the change i n
regulating rod position with the fuel flow a t i t s
normal value o f 46 gpm. During t h i s time the regulating rod changed from 7.85 to 7.69 in., a movement o f 0.16 in., which corresponded t o a A k / k o f
T h i s experiment was not repeated
only 5.3 x
a t high power, and therefore no data are available
on reactor power and temperature changes as a
function of sodium flow.
The second effect, the effect on the reactor system o f changing the sodium cooling rate, was observed i n Exp. H-12 during high-power operation.
I n t h i s experiment the sodium system helium
blowers were turned o f f and the following power
and temperature changes were observed. A t 2110
on November 11 the reactor was a t 2.1-Mw extracted
power; of this amount, approximately 1.6 Mw was
being extracted from the fuel and 0.5 Mw from the
sodium. A t 2113 the sodium system helium blowers
were turned off, and the power dropped, i n the next
90 sec, to 1.96 Mw a t a rate o f 1.5 kw/sec, and,
1
in the next 9 4 min, it dropped a t a much slower
rate o f 280 w/sec and leveled o f f a t 1.80 Mw. The
total decrease i n power was 300 kw. It i s t o be
noted that since the extracted power from the
sodium at the beginning of the experiment was
500 kw, about 200 k w of heat was s t i l l being l o s t
by radiation and other means through the barrier
door openings and other parts o f the circuit, even
with the helium circulation stopped.
When the
helium blower was turned on again a t 2124 the
reactor responded with an i n i t i a l rise o f 120 k w t o
i
c
.
*
6
c
k
E
I
I
k
1
fE
1
d
d
.
;
'
4
2
decrease noted. There was no drop i n temperature
which matched the very slow rate of decrease i n
power after the i n i t i a l decrease. Upon starting the
sodium blowers again a t 2124 the mean temperature
rose 13OF i n 5 min to a new value o f 1323OF, 10°F
higher than at the start o f the experiment. The
increase over the i n i t i a l temperature was partly
attributable to the thermal inertia of the reactor.
The reactor did not lose heat rapidly, but, when the
blowers were again turned on, the nuclear power
production increased by 300 kw and raised the
temperature level to above that a t the i n i t i a l condition.
It may be concluded from these results that
neither stopping the sodium flow or the sodium
cooling had much effect on the reactor behavior.
I t i s interesting t o observe that the mean reactor
fuel temperature went down with the decrease i n
nuclear power, which resulted from stopping the
It might be suspected that the
sodium blowers.
reactor mean temperature would have risen because
of lack o f cooling i n the moderator.
JI
4
1
4
1.92 Mw a t the rate of 1.33 kw/sec and then a very
slow rise over a period o f nearly 18 min to 2.06 Mw
a t a rate only 0.1 as great.
Any over-a1 I reactor temperature effect from
sodium transients would have been reflected i n a
change i n the mean fuel temperature, and therefore
the mean fuel temperature was observed.
When
the sodium blower was turned o f f a t 2113 the reactor mean fuel temperature dropped from 1313 t o
1310°F in 3 min, corresponding to the i n i t i a l power
.
J
t
Reactivity Following a Scram
1
1
1
1
,
-
The behavior o f a reactor after a scram i s determined primarily by delayed neutrons and photoneutrons,12 which i n the ARE were produced i n
the beryllium oxide moderator. These processes
keep the flux o f the reactor from decaying immediately.
It was observed that after every scram the reactor decay took place i n a series o f descending
oscillations, the time between oscillations being
equal to the fuel transit time o f 47 sec. Figure
6.17 shows a trace of one o f the fission chambers
after the scram a t the conclusion of Run 7 o f Exp.
L-5, one o f the low-power runs. The scram took
place a t the extreme right-hand edge of the trace.
Six prominent pips and many lesser ones can be
seen as the decay progressed.
12S. Glasstone and M. C. Edlund, EIements of Nuclear
Reactor Theory, Van Nastrand, p 88-89 (1952).
T h i s behavior can be explained on the basis that
the slug o f fuel passing through the reactor a t the
instant before the scram carried the l a s t group o f
delayed neutron emitters produced a t power. T h i s
phenomenon was particularly pronounced at the
time the photograph of Fig. 6.17 was taken because
it was a t that time that an experiment13 was under
way i n which the reactor was allowed to r i s e from
1-w power t o a predetermined power level of from
50 t o 100 w on a constant period before it was
scrammed. A t the instant of scram the power of.the
reactor was 100 w, whereas one transit time previously (47 sec) the power was only about 14 w. A s
a result, immediately after the scram the delayed
neutrons were particularly intense i n the slug o f
fuel which was i n the reactor a t the time o f the
scram. The transit time o f 47 sec i s comparable
to the 56-sec group o f delayed n e ~ t r 0 n s . l ~A s
this slug traversed the system and went back
through the reactor, the delayed neutron emitters
gave r i s e to neutron multiplication within the reactor and, a t the same time, the associated gamma
rays f e l l on the beryllium o f the moderator and
gave r i s e t o more neutrons by the reaction
Be9
+
y
+ Be8 +
n
.
The total effect of both types o f reactions was
to give a strong multiplication every 47 sec. During
the f i r s t pip shown in Fig. 6.17, the flux rose by
a factor of 2; with succeeding pips, the multiplication became less. The average mean decay time
of the f i r s t four pips was 72 sec.
For the two
longest lived groups of delayed neutrons the mean
lives are 32 and 80 sec.14 Therefore, the attenuation of the pips closely followed the theoretical
A remarkable feature o f
delayed neutron decay.
Fig. 6.17 i s that the pips could be distinguished
for about 12 min. Since after the f i r s t 3 or 4 min
the delayed neutron emitters were gone, the remaining effect was due solely to photoneutrons.
The scram behavior a t high power was very
similar, as observed with the use of the safety
chambers.
The fission chambers could not be
used a t high power, and therefore the f i r s t few
seconds after a scram could not be observed with
them.
l3In experiment L-5 the r e g u l a t i n g rod was c a l i b r a t e d
b y the periods induced b y rod m o t i o n (cf., chap. 4).
14Glasstone and Edlund, op. cit., p
65.
93
ORNL-LR-DWG
3855
P
t
CR
M
NO.!;
RUN 7, EXP. L5
t
*
t
t
t
L
f
L
HER TO GIVE A GRAPHIC REPRESENTATION OF THE DECREASE IN COUNTING RATE IN
THE CIRCULATING SLUGS FOLLOWING A SCRAM. THERE
IS, HOWEVER, SOME ERROR IN BOTH ABSCISSA AND
ORDINATE ACROSS THE JOIN.
I
k
h
W
x
2.
e
e
.-
L
W
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I
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t
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i
h
E,
h
k.
&.
1
I
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I
I
I
I
I
I
I
1
I
L
l
I
-
I
l
.
l
I
I
I
TIME
Fig.
94
6.17.
Circulating Slugs After a Scram.
I
I
I
t
1
I
I
I
c
!
F I N A L O P E R A T I O N AND SHUTDOWN
i
1
J
The l a s t scheduled experiment conducted on the
reactor was the measurement o f the xenon buildup
The 10-hr
following the 25-hr run a t 2.12 Mw.
period o f operation at one-tenth f u l l power was
concluded at 0835 on November 12. During the
following 11%-hr period from 0835 t o 2004, when
the reactor was shut down the final time, the operation of the reactor was demonstrated for A i r Force
and ANP personnel who were gathered for the
The
quarterly ORNL-ANP Information Meeting.
demonstrations included repeated cycling o f the
load by turning the blowers on and off, and group
movement o f the three shim rods t o change the
reactor mean temperature.
The information obtained during t h i s time on’the dynamic behavior of
the reactor i s described above under the subtitle
It
Reactor Kinetics.”
A l s o during the final 11 P2-hr period of operation,
two complete surveys o f system temperatures were
taken with the reactor a t maximum power. It was
during one o f these runs that the equilibrium fuel
outlet temperature of 158OOF and the total reactor
power o f 2.45 Mw was attained (cf., App. L).
Since a l l operational objectives of the experiment
had been attained and it was estimated on the
basis of time and the assumed reactor power that
the desired integrated power o f 100 Mwhr would be
attained by 2000 on Friday, November 12, i t was
decided t o terminate the experiment at that time.
Colonel Clyde D. Gasser, Chief, Nuclear Powered
Aircraft Branch o f WADC, who was then v i s i t i n g
the Laboratory, was invited t o officiate at the
termination o f the experiment. At 8:04 P M on
November 12, with Colonel Gasser at the controls,
the reactor was scrammed for the l a s t time and the
operation of the Aircraft Reactor Experiment was
brought t o a close. A photograph taken in the control room a t that time i s shown as Fig. 6.18.
At the time the experiment was terminated i t was
believed that the total integrated power was more
than the 100 Mwhr which had been prescribed as
a nominal experimental obiective, but subsequent
graphical integration of the L o g N charts, Appendix R, revealed that the actual total integrated
power was about 96 Mwhr. At the time the reactor
was scrammed the sodium system had been i n
operation (circulating sodium) for 635.2 hr and the
fluoride system for 462.2 hr. Of t h i s total fluoride
circulating time 220.7 hr were obtained with the
reactor c r i t i c a l and 73.8 hr after the reactor was
f i r s t brought t o power.
Although the nuclear operation was concluded
on Friday evening, the fuel and sodium were permitted t o circulate until the following morning, a t
which time they were dumped into their respective
dump tanks.
The final phase of the experiment,
including the dumping operation, subsequent analys i s and recovery of the fuel, and examination of
the systems and components for corrosion, wear,
radiation damage, etc., are t o be discussed i n a
subsequent report. Since much o f the data cannot
conveniently be obtained u n t i l the radioactivity
has decayed, the l a s t report w i l l not be forthcoming
immedi ately.
I
i
95
96
J
t-
I
!
i
7, RECOMMENDATIONS
N o comprehensive report i s complete without
'
1
C
conclusions, discussion, and recommendations. In
t h i s report the conclusions are presented in the
summary a t the beginning, and the discussion i s
incorporated in the text and, especially, i n the
Appendixes. Furthermore, certain alterations and
modifications that would have been desirable in
the conduct o f the experiment are implicit in the
discussions throughout the report. However, the
changes, i f effected for any similar future experiments, could lead t o substantial improvements
in design, instrumentation, operation, ( - i d interpretation.
Accordingly, a Ii st o f recommendations
based on the ARE experience i s presented; no
significance i s inferred by the order o f listing.
1. The operating crew should maintain the same
electricians,
shift schedule as the craft labor
assigned
instrument mechanics, pipe fitters, etc.
t o the operating crew, and a l l members of one crew
In
should have their off-day a t the same time.
particular, there should be four operating crews
that maintain the same rotating schedule as the
rest o f the plant.
2. The cause o f the obvious discrepancy between
the temperqtures read by the I ine thermocouples
close t o the reactor and those further removed
from the reactor should be ascertained. No physical phenomena, with the possible exception o f
radiation, have been proposed, t o date, which could
account for the observed difference.
-
-
3. In order t o measure the extracted reactor
power from the secondary or tertiary heat transfer
mediums, these systems should be adequately
instrumented for flow rates and temperatures.
Such instrumentation might also make possible
a thermodynamic analysis of the performance of
the various heat exchangers.
d
4
a,
1
4. Thermocouples on the outer surface o f fuel
and sodium tubes in the heat exchangers should be
installed so that they read wall temperatures rather
than an intermediate gas temperature.
5. Of the two thermocouples which were required
t o measure the reactor mean temperature and the
two required t o measure the temperature gradient,
one was electrically insulated from the pipe wall.
T h e insulation effected a time lag between the
wall and thermocouple temperatures o f about
15 sec. Thermocouple installations for obtaining
time-dependent data should be made so that the
wall temperature can be read without a time lag.
6. T o heat small (up t o
in. OD) lines and
associated valves, tanks, etc. to uniform high
temperatures (-1400" F) requires a surprising
degree o f precision in the installation o f both
heaters and insulation. Furthermore, where calrod
heaters are employed, they should be installed on
opposite sides of a line, and the heaters and l i n e
should be jointly wrapped with heat shielding
before insulation i s applied.
7. T o control accurately the temperature of any
system,
thermocouples should be installed a t
any discontinuity of either the heaters (i.e.,
between adjacent heaters) or the system (i.e., a t
the junction o f two lines, etc.), and, w i t h the
exception of obviously identical installations,
each heater should have i t s own control.
8. Where gas lines are subject t o plugging due
t o the condensation of vapor from the liquid i n the
system to which the lines are connected, it i s
necessary either t o heat the lines t o above the
freezing point o f the vapor or t o employ a vapor
trap.
While a vapor trap proved satisfactory in
preventing the fuel off-gas l i n e from plugging, it
was not possible t o heat the sodium off-gas l i n e
sufficiently (because o f temperature limitation o f
the valves) t o prevent the gradual formation of a
restriction.
Gas valves that can be operated at
higher temperatures and are compatible with
sodium are needed.
9. The double-walled piping added a degree o f
complexity t o the system that was far out o f proportion t o the benefits derived from the helium
annulus. Parts o f the annulus were a t subatmospheric pressure, and no leak tests were made on
the fuel system once it was f i l l e d with the fluoride
mixture.
The helium flow was not needed for
distributing heat in the sodium system, and it was
Conseof uncertain value in the fuel system.
quently, future systems should not include such
an annulus.
10. The use o f any type o f connection other
than an inert-arc-welded joint in any but the most
temporary fuel or sodium l i n e should be avoided.
11. A l l joints, connections, and f i t t i n g s in the
off-gas system should be welded, and, in general,
they should be assembled with the same meticulous
care that characterized the fabrication o f the fuel
and sodium systems in order to minimize the poss i b i l i t y o f the unintentional release o f f i s s i o n
gases.
i
i
i
L
I
,
[
i
97
I
!
z
c
4
h.
I
12. A s a secondary defense in the event of the
release of activity in the reactor c e l l (i.e., pit),
the c e l l should be leaktight. The leak-tightness of
the numerous bulkheads out of such a c e l l should
be carefully checked.
13. The air intake t o the control room was located on the roof o f the building, and thus adverse
meteorological conditions readily introduced offgas activity into the control room. With an airtight
control room equipped with i t s own air supply (or
a remotely located filtered intake), the control room
operations could continue without concern for the
inhalation of gaseous activity.
14. The off-gas monitrons should be shielded
from direct radiation. Furthermore, although these
monitrons were not needed during the experiment,
they are known t o build up background activity
which would eventually mask that o f the gas they
are to measure. The development o f monitrons in
which activity w i l l not accumulate i s recommended.
15. The radiation level and the airborne activity
throughout the building were measured by monitrons
and constant air monitors, respectively. T h e data
from both instruments should be continuously
recorded. Furthermore, the output o f the a i r monitors should be modified by the addition of a differentiating circuit so that the recorded data would
give better measures of the airborne activity.
16. The helium ducts leaked t o the extent that
i t was possible t o attain only a fraction o f the
desired helium concentration therein. They were
o f the conventional bolted-flange design and should
have been modified t o permit seal welding o f a l l
joints; they should be subjected t o stringent leak
tests.
17. The helium consumption for the l a s t three
and one-half months of the ARE experiment was
million standard cubic feet (over 3000 standard
cylinders).
The average consumption rate during
the l a s t two weeks o f the experiment was about
8.5 cfm, and peak consumption rates o f 25 cfm were
recorded.
T h i s extraordinarily high helium consumption could be reduced by the use o f dry air
or nitrogen for much of the pneumatic instrumentation in which helium was used.
18. Various valves in both the gas and liquid
systems leaked across the valve seats. T h e valve
development program should be emphasized unti I
valves are obtained that can be depended upon as
reliable components of high-temperature (* 1200O F)
sodium, fluoride, or gas systems.
Furthermore,
either l i m i t switches that would operate a t higher
5
98
temperatures t o indicate valve open or closed
should be developed, or the existing l i m i t switches
should be located i n cooler regions.
19. The use o f frangible disks t o isolate the
standby fuel pump did not enhance the feasibility
o f continuing the experiment i n the event o f the
failure o f the main fuel pump during high-power
operation. The frangible disks are objected t o i n
that they require a nonreversible operation.
In
general, leak-tight valves should be developed
that may be used rather than the frangible disks.
20, Although the ARE, as designed, was t o
incorporate numerous “freeze sections,” only two
were included in the system as finally constructed,
and the operability of these two was so questionable thcrt their use was not contemplated during
the course of the experiment. Such sections should
either be eliminated from consideration in future
systems, or a reliable freeze section should be
developed.
21. Much useful information would be obtained
i f provisions could be made for sampling each
liquid system throughout the operation. T h i s was
possible on the ARE only prior t o the high-power
experiments.
22. The variable inductance-type flow and level
indicators should be converted t o the null-balance
type o f instruments in order t o eliminate the
temperature dependence of the pickup c o i l signal.
Furthermore, these c o i l s should be located in a
region a t much less than 1OOOOF in order t o increase c o i l life.
23. Although numerous spark plug probes were
satisfactorily employed t o measure levels in the
various tanks, the intermittent shorts that were
experienced with several sodium probes might
have been avoided i f clearances between the
probe wire and the stand pipe in which the probes
were located had been greater.
24. The use of mercury alarm switches on
vibratory equipment where there i s l i t t l e leeway
between the operating and alarm condition wi I I
give frequent false alarms and should therefore be
avoided.
25. Weight instruments are o f questionable
accuracy in a system in which the tank being
weighed i s connected through numerous pipes to
a fixed system, especially when these pipes are
covered with heaters and insulation and are subject
to thermal expansion.
26. The flame photometer i s an extremely sensit i v e instrument for the detection o f sodium and
L
13
b
t
v
t
f
E
I
k
i
NaK.
However, in employing t h i s instrument t o
detect the presence o f sodium (or NaK) i n gas, the
sampling l i n e should be heated t o about 300OF.
27. The magnet faces i n the shim rods should
be designed so that d i r t particles cannot become
trapped thereon and thus require higher holding
currents.
28. The two control points, i.e., the upstairs
control room and the basement heater and instrument
panels, should have been more convenient t o one
another i n order t o effect the greatest efficiency
of operation.
29. The communications system i n the building
was inadequate.
Except for the auxiliary FM
system, which was frequently inoperative, there
was only one phone in the control room, which
was on the main station of the PA system. However, the PA system was not capable of audibly
supporting two conversations a t the same time.
Accordingly, the capacity of the PA system should
be increased so that as many as 4 or 5 conversations may be simultaneously effected.
Also,
more outlets should be provided both in the control
room and a t other work areas i n the building.
30. A megawatt-hour meter should be installed
(possibly on the log N or the AT recorder) i n order
t o provide a continuous measure o f the integrated
power as the experiment progresses.
31. In order t o analyze various related time-
4
dependent data (as required, for example, i n the
determination o f the various temperature coefficients), it i s necessary that the recorder charts
be marked a t the start o f the experiment. While
the charts can be marked by hand, t h i s was a
source o f error which became very important in the
analysis of data from fast transients. Accordingly,
a l l such charts should be periodically and simultaneously marked by an automatic stamper. Furthermore, each chart should be stamped with a dist i n c t i v e mark that would positively identify it.
32. T o analyze the kinetic behavior of a reactor
system i t i s necessary to have a recorder chart of
the time behavior of a l l equipment which can
introduce transients in the reactor, as w e l l as
charts of the process data and conventional nuclear
data.
Therefore the helium blower speeds and
shim rod positions should have been recorded
conti nualI y.
33. The recorded data and the data sheets
comprise a f a i r l y comprehensive picture o f the
experiment.
However, even i f a l l t h e data are
recorded they are of value only t o the extent that
meaningful interpretations can be made.
While
automatically marking the data charts w i l l be
helpful, it w i l l s t i l l be necessary to rely on log
book information which should be recorded in
great detail, possibly t o the extent o f making t h i s
the only responsibility of a s h i f t "historian."
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t
b
= I
L
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Appendix B
SUMMARY OF DESIGN AND OPERATIONAL DATA
i
f
This appendix tabulates both the design and operational data pertinent t o the aircraft reactor experiment. Most of the design data wereextracted from two design memoranda,’a2 although some values had t o
be revised because of subsequent modifications i n the design. In addition t o the design data, the experimental values of the various system parameters are included, All values obtained experimentally are
shown in ita1 ics and, for comparative purposes, are tabulated together with the corresponding design
values. There are, of course, numerous design numbers for which it was not possible t o obtain experimental numbers. The flow diagram o f the experiment is presented in Fig. 6.1. The values of temperature,
pressure, and flow given on t h i s drawing are design values - not experimental values.
!
j
t
’W. B. Cottrell, A R E Design Data,
ORNL CF-53-12-9 (Dec. 1, 1953).
2W. B. Cottrell, A R E Design Data Supplement, ORNL CF-54-3-65 (March 2, 1954).
4
i
DESCRIPTION
d
L
1. The Reactor Experiment
4
4
d
Type of reactor
Neutron energy
Power (maximum)
Purpose
Design lifetime
Fue I
Moderator
Reflector
Primary coolant
Ref lector coo Iant
Structural material
Test stand
Shield
Heat flow
Circulating fuel, solid moderator
Thermal and epithermal
2.5 Mw
Experimental
1000 hr
NaF-ZrF4-UF4 (53.09-40.73-6.18 mole %)
Be0
Be0
The circulating fuel
Sodium
Inconel
Concrete p i t s i n Building 7503
7P2 ft of concrete
Fuel t o helium t o water
t
4
a
2.
e
P h y s i c a l Dimensions of Reactor (in.)
Hot
(1300O F)
t.
35.60
35.80
32.94
33.30
35.60
35.80
32.94
33.30
47.50
48.03
4.00
4.25
32.94
33.30
4.93
4.98
32.94
33.30
48.00
48.55
2.00
2.02
44.50
45.00
4.00
4.04
66 parallel Inconel tubes containing
Core height
Core diameter
Side ref Iector height
Side reflector inside diameter
Side reflector outside diameter
Top reflector thickness
Top reflector diameter
Bottom reflector thickness
Bottom reflector diameter
Pressure shell inside diameter
Pressure shell wall thickness
Pressure shell inside height
Pressure she1I head thickness
Fuel elements
3.
Cold
(70' F)
the circulating fuel. The tubes
were connected i n six parallel circuits each having 11 tubes i n
series. Each tube was 1.235 in.
O.D., with a 60-mil wall.
Volumes of Reactor Constituents (Cold)
a.
B
Inside Pressure Shell
Volume (ft3)
Be0
Core
Side reflector
Annulus outside reflector
Top reflector
Bottom reflector
Space above reflector and annulus
Space below reflector and annulus
14.55
17.68
0
0
0
0
0
32.23
Total
Total
Fuel
No
Inconel
Rodsa
1.33
0.21
0.24
0
0
-
0.80
0.91
0.58
1.62
1.67
2.14
2.17
-
0.37b
0.10"
0.19
0.08b
0.44b
0.04"
0.52c
0.50
0.26d
0
0.06
0.07
0.03d
0.Md
17.55
18.95
0.77
1.97
2.42
2.21
2.73
1.78
9.89
1.74
0.96
46.60
0
0
f
t
li
&
~
nTotal volume i n s i d e inner rod sleeve.
P
blncludes the Inconel-clad s t a i n l e s s steel r o d sleeves.
i
clncludes a l l three sleeves around f i s s i o n chambers
dlncludes volume of i n s u l a t i n g material around inner f i s s i o n chamber sleeve.
I,
b. Operating Fuel System
bi
b
Volume (ft3)
Cold
Core
External system (to minimum pump level)
Pump (available above minimum level)
Total
108
a
-
1.33
3.48
1.65
-
6.46
Hot
-
1 37
3.60
1.70
-
6.67
= 6
a
-
E
b
r
3
I
~
TO DRAIN
4
508
m
i
t
D
k
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I
I
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L
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i
44
5
f
P
c.
Miscellaneous Fuel System
Volume
d
(ft3)
14.5
Maximum capacity f h d flush tanks
Bottom (waste) in f i l l and flush tanks
Dump l i n e t o f i l l and flush tank No. 2
Dump l i n e t o f i l l and flush tank No. 1
By-pass leg
Heat exchanger leg from pump t o f i l l l i n e t o Tank No. 2
Reactor leg from fill line t o pump (minimum operating level)
Pump capacity, minimum t o maximum level
Carrier available
Concentrate ava iIab I e
s
0.3
0.8
0.75
0.4
rrsasa
1.6
3.0
1.7
15.5
1.3
i
d. Sodium System
Volume
(ft3)
Inside pressure shell
Outside pressure shell
.-
R,
10
10
-
20
Total
MAT E RIALS
1
1
1
-
f
1. Amounts of C r i t i c a l Materials
LL
t
i
s
B e 0 blocks (assuming p = 2.75 g/cm3)
B e 0 slabs (assuming p = 2.75 g/cm3)
Amount of uranium requested
Uranium enrichment
Uranium in core
Uranium inventory in experiment
5490 Ib
48 Ib
253 Ib of U235
93.4% $35
30 t o 40 Ib
126 t o 177 Ib
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2
Composition
of Reactor Constituents
Fluoride Fuel Mixture
a.
Fue I a
Fuel Carrier
Fuel Concentrate
NaF
mole
wt %
53.09
&34
50
20.1
66.7
21.3
40.73
62.12
50
0
79.9
0
6.18
17.54
0
0
33.3
78.7
25
35
10"
47d
Zr F4
mole %
wt %
"F4
mole %
wt %
Impurities (ppm)
<5 b
Ni
Fe
Cr
5b
449
-
84d
2Od
aFuel composition for high-power operation.
Final ana I ys i s before hi gh-power operat ion.
CAverage of a l l 14 batches of carrier.
dEstimate based on a few prel iminary analyses.
b.
Inconela
Constituent
Amount
(wt %)
Con s t ituent
Amount
(wt %)
Ni
Cr
Fe
Mn
78.5
14.0
6.5
0.25
0.25
0.2
Cob
AIb
Tib
Tab
0.2
0.2
0.2
0.5
Si
cu
Wb
0.5
Znb
0.2
Constituent
Amount
(wt %)
Zrb
C
Mo
Ag, B, Ba
Be, Ca, Cd
V, Sn, Mg
0.1
0.08
Trace
Trace
Trace
Trace
aB. B. Betty and W. A. Mudge, Mechanical Engineering, February 1945.
bAccuracy, *loo%. Y-12 Isotope Analysis Methods Laboratory (spectrographic analysis). Y- 12Area Report Y-F2&14.
c.
Impurities
Si
AI
Pb
Ni
Mn
co
Amount
(PPd
1050
213
45
<20
<5
<1
Beryllium Oxide*
Impurities
Ca
Fe
Ztl
Cr
B
Amount
(PPd
780
114
<30
10
2.8
Impurities
Amount
(PPd
Na
330
Ms
K
Li
< 25
Ag
*W. K. Ergen, Activation of Impurities in BeO, Y-12 Area Report Y-F20-14 (May 1, 1951).
112
50
<5
<1
d. H e l i u m
<10 pprn
Oxygen contamination
e. Sodium
~ 0 . 0 2 5wt %
Oxygen contamination
3.
Physical Properties of Reactor Materials
Melting
Point
("C)
Thermal
Conductivity
(Btu/hr.ft."F)
Fuel Carrier:a
NaF-ZrF, (NaZrF,)
50-50 mole %
510
2.5 kSb
Fuel Concentrate:'
NaF-UF, (Na,UF,)
66.7-33.3 m o l e %
635
Fuel:a
NaF-ZrF,-UF,
53.09-40.73-6.18
530
mole %
Heat
Capacity
(Bt u/l b e " F)
Viscosity
(CP)
Density
(s/cm3)
0.30b
p = 3.79 - 0.00093T
600 < T < 800°C
0.5 (estimated) 10.25 at 700OC
7.0 at 800°C
5.1 at 900°C
0.21b
p = 5.598
1.3 (estimated) 8.5 at 600°C
5.7 at 700" C
4.2 at 800°C
0.24b
p = 3.98
8.0 at 600°C
5.3 st 700°C
3.7 at 800°C
- 0.001 19T
600 < T < 800°C
- 0.00093T
600 < T < 800OC
BeOC
2570
16.7 at 1500°F
19.1 at 1300°F
22.5 at 1100" F
0.46 at 1 100°F
0.48 at 1300°F
0.50 a t 1500°F
from 2.27d to 2.83
at 2 0 T
Inconele
1395
8.7 from 20 to
0.101
77 < T < 212°F
8.51 at 20°C
100°C
10.8 at 400°C
13.1 at 800°C
0.38 at 250°C
0.30
0.27 at 400°C
0.18 at 700°C
<-272.2 0.100 at 200°F 0.0267 at 200T 1.248
0.119 at 400°F 0.0323 at 400°C 0 < T < 300°C
0.136 at 600°F 0.0382 at 600°C
98
Sodiumf
He1 i u m g
43.8 at 3OOOC
38.6 at 500T
0.85 at 400T
0.82 at 500OC
0.78 at 700T
1.79 x lo', at 0°C
1.30 x lo-, at 100°C
1.03 x lo-, at 200°C
0.18
0.13
0.38
0.48
lnsu lation
Superex (Si0.J
.
Sponge felt (MgSiO,)
4.
aData from ANP physical Properties Group.
bPreliminary values for the liquid 600 t o 800°C.
=Data from The Properties o/ Beryllium Oxide, BMIT-18 (Dec. 15, 1949).
dPorosity of BeQfrom ANP Ceramics Group, 23% a t p = 2.27, 0% a t p =2.83.
eData from Metals Handbook, 1948 Ed., The American Society for Metals.
{Data from Liquid Metals Handbook, NAVEXOS P-733 (June 1952).
gB. 0. Newman, Physical Properties of Heat Transfer
Fluids, GI-401 (Nov.
10, 1947).
113
REACTOR PHYSICS
1.
Neutron energy
Thermal fissions, %
Neutron flux, n/cm2.sec
Gamma flux, y/cm2-sec
2.
f
$
General
t
Thermal and epithermal
-
60
10’4
-101’
+
h
Power
Maximum power, Mw
Design power, Mw
Power density (core av a t max power) w/cm3
Maximum specific power (av kw/kg. o f U235a t
Power ratio (max/av)
Axiolly
Radially in core
Radially in fuel tube
Maximum power density in fuel, w/cm3
Maximum power density in moderator, w/cm3
3. Neutron Flux
in
1.5
Mw)
1.5
1.o
(2.5)
3
94
(51
(153)
1.5: 1
1.2: 1
1.7: 1
110
(180)
1.3
(2.2)
a
Core (n/cm2*sec)
Thermal (max)
Thermal (av)
Fast (max)
F a s t (av)
Intermediate
4.
5.
Leakage F l u x (per fission)
Reflector
Fast
Intermediate
Therma I
0.001 1
0.066
0.191
Ends
Fast
lntermediate
Thermal
0.013
0.102
0.0002
Fuel
u235
Enrichment, 7%
Critical mass, Ib of U235 in clean core
U235 i n core (i.e., c r i t i c a l mass plus excess reactivity)
U235 in system, Ib
consumption a t maximum power, g/day
u235
6.
93.4
(32.75)
Ib
30 t o 40
(36)
126 t o 177 (138)
1.5
(2.5)
Neutron F l u x in Reflector
Maximum flux a t fission chamber holes, nv/w
Counting rate o f fission chambers, counts/sec.w
1.5 x lo6
2 x 105
i
d
7.
Reactivity Coefficients
Effect
L
T her ma I base
-0.011 from 68 t o 1283OF
-0.009 from 1283 t o 1672OF
hk/k
1
Uranium mass (at k
i
Va I ue
Symbol
=
1)
bk/k
0.25
(0.236)
1
1
.
d
*
Sodium density
-0.05 from 90
to
Moderator density
0.5 from 95
100% p
Inconel
Fuel
to
t
100% p
-0.17 from 100 t o 140% p
density
(Ap/p)
(Ak/k)/O F
temperature
(-9.8 x 10-5)
- 5 ~ 1 0 - ~(-6.1 x
Reactor temperature
(Ak/k)/O F
Sodium temperature
(Ak/k)/O F
(- j.88 x
Moderator temperature
(Ak/k)/' F
(1.1
10-5)
SHl EL D I N G
t
i
T
1. Material
Thickness (in.)
(from source outward)
2.
i
Barytes concrete block in
the reactor p i t only
12
Poured Portland concrete
18
2.3
Barytes
60
3.3
concrete
block
Composition
1
d
Cement
Concrete
Barytes
CaCO, +- SiO, + AI,O,
Cement and aggregate of gravel
Cement and aggregate of BaSO,
3. Relaxation Lengths (cm)
ii
i
]-MeV gamma rays
2.5-Mev gamma rays
7.0-Mev gamma rays
Fast neutrons (1 t o 5 MeV)
Barytes Concrete
Portland Cement
4.14
6.72
9.46
8.2
4.62
6.86
14.0
11.0
115
4.
i
Source Intensities
No. per F i s s i o n
Core gammas
Prompt
Fiss ion-product decay
2.0
2.0
0.94
Capture gammas
5.
Energy (Mev)
2.5
2.5
7.0
Gamma f l u x outside of reactor thermal insulation per watt
2.0-Mev gammas
7.0-Mev gammas
3 x io4 y/cm2.sec
0.8 x lo4 y/cm2.sec
Neutron flux outside of reactor thermal insulation per watt
Thermal t o 250 kev
Fast t o 250 kev
3 x 105 n/cm2-sec
1.5 x lo4 n/cm2.sec
A c t i v i t y in External F u e l Circuit per Watt
Time Out of
Reactor
(set)
0.52-Mev Gamma
Activity
(photons/cm3 Ssec)
I-Mev Delayed-Neutron
Activity
(n/cm3.sec)
0
5
10
20
30
40
8.1 x lo6
6.6 x lo6
5.9 x lo6
5.1 x lo6
4.7 x 106
4.5 x 106
8.5 io3
2.3 103
1.4 103
0.74 103
0.50 103
0.35 io3
E:
c
i
i
=i
REACTOR CONTROL
1.
Control Elements
Source
Regulating rod
Shim rods
Temperature coefficien (Ak/k)/'
2.
1
1
3
--5 x
F
I
6
@
I
low5
(-6.
e
10-5)
Source
Core axis
Po-Be
Locat ion
TYPe
Strength, curies
Neutron intensity, n/sec
15
3.5 x 10'
(7)
(1.6 x 107)
a,
3. Regulating Rod
Location
Diameter, in.
Travel, in.
Cooling
Total Ak/k, %
Maximum (Ak/k)/sec
Speed, in./sec
116
Core axis
2
12
C
b
He1ium
(slowspeed), %
0.40
(0.40)
0.010
(0.011)
0.3 or 3 (0.32) (fast speed not used)
a
:
I
Backlash, in.
0.007
Diehl lA, 115 v, 60 cycle, 3400 rpm, reversible
Manual
Temperature error-signal (not used)
Flux signal,
k 2 % N F ; 5 1.3OF mean T
Servo motor
B
Actuation
Regulation
There were seven regulating rods made up which gave a calculated A k / k from 0.13 t o 1.35%.
used in the experiment was the one which most nearly gave a measured Ak/k of 0.4%.
Rod number
Calculated Ak/k
Measured Ak/k
Weight per inch of rod, Ib
i
1
0.13
2
0.18
3
0.27
0.042
0.061
0.091
4
0.39
(- 0.25)
0.132
5
0.56
(0.40)
0.21
t
The rod
6
0.90
7
1.35
0.32
0.50
t
4. Shim Rods (3)
One a t each 120 deg
2
36
He1ium
Location
Diameter, in.
Travel, in.
Cooling
Material
Magnet release t ime, sec
Maximum withdrawal speed, in./sec
Total Ak/K per rod, %
Maximum (% Ak/k)/sec per rod
Motor
J
i
.
_
on
7.5-im-radius c i r c l e
i
B4C
< 10-2
0.036
5.0
0.005,
(0.046)
(5.8)
(0.0039 at 4-in. i n s e r t i o n )
av over rod
Janette 1.2 amp, 115 v,
60 cycle, 1725 rpm, reversible
5. Nuclear Instrumentation
a.
Fission
Chambers (2)
Function
Sen s i t i v i t y
Count ing-rate signal
0.14 (counts/sec) per (n/cm2.sec)
Range
lo”,
4
i.e.,
lo4 in instrument, lo7 in position and shielding
b. Parallel-Circular-Plate Ionization Chambers (3)
(2 chambers)
(2 chambers)
Sensitivity
Range
Function
c.
Function (1 chamber)
(1 chamber)
Sensitivity
Range
Safety level (scram signal)
Regulating rod temperature servo
5O pa at 10’0 n/cm2-sec
lo3, i.e., from 5 x
to 1.5 N ,
Compensated Ionization Chambers (2)
Micromicro ammeter
Regulating rod flux servo
50 pa at 10” n/cm2-sec
* IO6, i.e., from 10-6 to 3 N,
I
117
6. Scrams and Annunciators
a.
Cause
Neutron level
Period
Period
Reactor exit fuel temperature
Heat exchanger e x i t fuel temperature
Fuel flow
Power
Scram switch
'Shim
I
Automatic Rod Insertion and Annunciation
Nuclear
Scram
Set Point
1.2 and 1.5 N,
1 sec
5 sec
> 155OOF
b
Process
Scram
Reverse'
Annunciator
No.
30
30
32
25
25
25
30
30
X
X
X
<llOO°F
<10 gpm
Off
Scram position
- .
c
I
rods automatically driven in.
bNeutron l e v e l set so that the f a s t scram annunciator, No. 30, would annunciate a t 1.2 N F (normal flux) although
the safety rods would not be dropped until the neutron level reached 1.5 N F .
There was not another neutron l e v e l
annunciator a t 1.5 NF'
b. Nuclear Annunciators
Safety circuit
Count rate meter
Servo
Rod-cool ing helium
Neutron level
118
f
-
Set Point
Annunciator No.
Electronic trouble
Off-scale
Off-scal e
Off
28
26
31
29
27
1.2 N,
P
L
k
,
c.
Fuel System Annunciators
Set Point
4
1
4
Pump power (main pump)
Low heat exchanger fuel flow
High heat exchanger water temperature
LOW
heat exchanger fuel temperature
High fuel level (main pump)
L o w fuel level (main pump)
High pump pressure (main pump)
High fuel reactor pressure
High fuel level (standby pump)
Low fuel level (standby pump)
High pump pressure (standby pump)
Pump power (standby pump)
High temperature differential across reactor tubes
High fuel temperature
Low heat exchanger fuel temperature
High heat exchanger tube temperature
Low heat exchanger tube temperature‘
Pump lubricant (main pump)
Pump coolant (main pump)
Pump !ubricant (standby pump)
Pump coolant (standby pump)
>50 amp
<20 gpm
> 16OOF
<115OoF
Maximum pumping level
Minimum pumping level
> 5 psi
>50 p s i (>41 psi)
> 5 psi
>50 amps
> 235O F (> 400” F )
> 15OOOF
<llOO°F
> 150OOF
<115OoF
< 4 gpm (<2 g p m )
<2 gpm (<z gpm)
(4 gpm (<2 gpm)
<2 gpm (<2 gpm)
Annunciator
NO.
33
34
35
36
37
38
39
41
45
46
47
48
49
51
52
53
54
65
66
69
70
aThe helium blower was interlocked so that when the fuel temperature decreased below t h i s temperature the blower
speed was reduced t o zero until the fuel temperature exceeded 1150OF.
d. Sodium System Annunciators
Set Point
Annunciator
No.
4
1
1
d
1’
d
f
f
Pump power (main pump)
Low sodium flow (back loop)
Low heat exchanger sodium temperature
High sodium level (main pump)
Low sodium level (main pump)
High pump pressure (main pump)
Pump power (standby pump)
L o w sodium flow (front loop)
LOW
heat exchanger sodium temperature
High sodium level (standby pump)
L o w sodium level (standby pump)
High pump pressure (standby pump)
High sodium temperature differential across reactor
Low sodium reactor pressure
Pump lubricant (main pump)
Pump coolant (main pump)
Pump lubricant (standby pump)
Pump coolant (standby pump)
>50 amp
<lo0 gpm
> 160OF
<llOO°F
9
10
11
>65 psi
>50 amp
(100 gpm
> 160°F
< 1 10QoF
>65 psi
>60°F
(60 psig
< 4 gPm
<2 gpm
< 4 gPm
( 2 gPm
(>48 p s i g )
( ( 2 gpm)
(<2 gPm)
(<2 g b m )
(<2 RPm)
12
13
14
15
17
18
19
20
21
22
23
24
42
67
68
71
72
119
e.
Miscellaneous Annunciators
Annunciator
Set Point
Sodium system he1ium blower
Pit oxygen
Pit humidity
Helium supply low
Space cooler water temperature
Low water flow (any of 5 systems)
Rod cooling water temperature
L o w reservoir level
Stack closed
Pit a c t i v i t y
Monitrons
Vent gas monitor
Vent header vacuum
L o w nitrogen supply
L o w air supply
DC-to-AC motor-generator set
AC-to-DC motor-generator set
No.
Sodium system blower on before fuel system blower
<90% He (Disconnected)
> 10% relative humidity
(500 psi
> 160OF
> 10% below design
>160OF
<16 p s i
<5 mph wind velocity or high ion chamber reading
0.8 pc/cm3
> 12 mr/hr (>7.5 m / h r )
0.8 pc/cm3
>29 in. Hg vacuum
<300 psi
<40 p s i
Off
1
2
3
4
5
6
7
8
55
56
57
58
59
60
61
63
64
Off
SYSTEM OP E R A T I N G CON D IT10 NS3
1. Reactor
Fuel inlet temperature, O F
Fuel outlet temperature, O F
Mean fuel temperatwe, O F
Sodium inlet temperature, O F
Sodi um out let temperature, O F
Fuel flow through reactor (total), gpm
Fuel flow rate i n fuel tubes (11.3 gpm), fps
Sodium flow through reactor, gpm
Heat removed from fuel, kw
Heat removed from sodium, kw
Fuel dwell time i n reactor, sec
Fuel cycle time
For 50% of fuel, sec
For rest of fuel, sec
Maximum fuel tube temperature, O F
Maximum moderator temperature, O F
Sodium circulating time, hr
Fuel circulating time,* hr
1315
1480
1400
1105
1235
68
4
224
1270
650
8.3
(1209)
( I 522)
(1365)
(1226)
(1335)
(4 6)
(3)
(150)
(1520)
33.6
46.4
1493
1530
1000
1000
(47)
(47)
(577)
(635)
(462)
*After 462 hr of circulating time, the l a s t 221 hr were attained after the reactor first became critical, and
these after the reactor first operated above 1 Mw.
74 o f
-
~
e
3The design values were taken from drawing A-3-0, "Primary H e a t Disposal System," F l o w Sheet, i n which the
physical properties of the fuel were assumed to be p = 3.27 g/cm3, ,u = 7 to 13 cp, and cp =0.23 Btu/lb.'F.
The
operating values were taken from the 2 5 h r Xenon run (E-.
H-8).
2. Fuel System
a.
Reactor out let
Heat exchanger i n l e t
Heat exchanger outlet
Pump inlet
Pump outlet
Reactor inlet
'There
Fuel Loop
Temperature
Pressure
Flow
(" F)
(Psi$
(gpm)
46
34
12
2 (0.3)
54
50 (39)
40
20 '
20
40 ( 4 6 )
40
40
1450
1450
1150
1150
1150
1150
(1522)
(1522)
(1209)
(1209)
(1209)
(1209)
were two fuel-to-helium heat exchangers i n parallel.
b. Helium Loop
(" F)
Fuel-to-he1ium heat exchanger No. 1 inlet
Fuel-to-he1ium heat exchanger No. 1 outlet
Helium-to-water heat exchanger No. 1 inlet
Helium-to-water heat exchanger No. 1 outlet
Fuel-to-helium heat exchanger No. 2 inlet
Fuel-to-he1ium heat exchanger No. 2 outlet
He1ium-to-water heat exchanger No. 2 inlet
He1ium-to-water heat exchanger No. 2 outlet
Blower outlet
'The
H,O)
(in.
(cfm)
7,300
12,300
12,300
7,300
7,300
12,300
12,300
7,300
7,300
2.2
1.5
1.5
1.2
1.1
0.4
0.4
0.1
2.3
180
620
620
180
180
620
620
180
180
helium passed through two temperature c y c l e s i n each loop.
c.
He1ium-to-water heat exchanger inlet
He1ium-to-water heat exchanger outlet
'The
Flow
Pressure
Temperaturea
Water Loop
Temperature'
Pressure
Flowb
(" F)
(psig)
(gpm)
70 (61)
135 (124)
10
8
65
(103)
65
water entered e a c h heat exchanger a t ambient temperature and was then dumped.
bThere were two helium-to-water h e a t exchangers i n parallel; total flow,
130 gpm.
3. Sodium System
a.
Reactor outlet
Sodium-to-he1 ium heat exchanger inlet
Sodium-to-he1ium heat exchanger outlet
Pump inlet
Pump outlet
Reactor inlet
Sodium Loop
Temperature
Pressure
Flow
(OF)
( P S 4
(gpm)
58
52
51
48 (36)
65
60 (49)
224 (152)
112
112
224
224
224
1235
1235
1105
1105
1105
1105
(1335)
(1335)
(1226)
(1226)
(1226)
(1226)
121
b. Helium Loop
(in.
(OF)
170
1020
1020
170
170
Sodium-to-ha ium heat exchanger inlet
Sodium-to-he1ium heat exchanger outlet
He1ium-to-water heat exchanger inlet
Blower outlet
c.
H,O)
(cfd
4
2
2000
4700
4700
2000
2
0
4
2000
Water Loop
Temperature
Helium-to-water heat exchanger inlet
4. Rod Cooling
Flow
Pressure
Temperature
ure
Pr
Flow
(" F)
(PSid
(gpm)
70 (61)
100 (114)
10
8
77 (38.3)
77
System
a.
Helium Loop
Temperature
Pressure
(in. H,O)
(" F)
240
240
110
110
110
Rod assembly outlet
Helium-to-water heat exchanger inlet
Blower
Rod assembly inlet
Flow
(cfm)
1270
423
333
0
13
l0OOb
1000
UThere were three heat exchangers in parallel.
bCapacity of each of the two parallel blowers; however, the second blower was i n standby condition.
b.
Water Loop
Temperat urea
Pressure
Flowb
(" F)
(psig)
(gpm)
70 (61)
100 (63)
10
8
18 (17.6)
18
Water-to-helium heat exchanger inlet
Water-to-he1 ium heat exchanger outlet
?he
water entered edch h e a t exchanger a t ambient temperature and was then dumped.
b T o t a l flow for three p a r a l l e l heat exchangers.
5.
Water System
Equipment
Spate coolers
Reflector coolant system
Rod cooling system
Fuel coolant system
Pump cooling systems
Total
122
F l o w per Unit
No. of
Flow
(gpm)
Units
(gpm)
7
77
8
2
3
2
4
6
65
3
56
154
18
130
12
370
06)
(77.6)
(17.6)
(206)
(23)
(370.2)
MISCELLANEOUS
1. Welding Specifications'
Process
Base metal
Position
F i l l e r metal
Preparation of base metal
Cleaning fluid
Clearance i n butt joints
Clearance in lap joints
Welding current
Electrode
Sh ielding gas b Ianket
Gas blanket beneath weld
Welding passes
Welders qualifications
I
B
I
Inert-gas, shielded-arc, d-c weld
lnconel
Horizontal rolled, fixed vertical, or horizontal
'/16
to ?32 in. lnconel rod (Inco No. 62 satisfactory)
Machined, cleaned, and unstressed
Trichloroethylene
in. with 100 deg bevel
"/32 to
Flush at weld
d-c, electrode negative, 38 t o 80 amp
!,6 to ?32 in. tungsten, 90 deg point
Argon, 99.8% pure; 35 cfh
Helium, 99.5% pure; 4 t o 12 cfh
1 to5
Welded or passed QB No. l b within 45 days
i
'For details, see Procedure Specifications, PS-1, O R N L Metallurgy Division, September 1952.
boperation Qualification T e s t Specifications, QTS- 1, ORNL Metallurgy Division, September 1952.
2.
Stress Analysis
a.
4
Moments and Stress i n the Pressure Shell Ends
Maximum radial stress (per psi pressure)
Minimum radial stress (per p s i pressure)
Maximum radial moment (per psi pressure)
Minimum radial moment (per p s i pressure)
Maximum tangential stress (per psi pressure)
Minimum tangential stress (per psi pressure)
Maximum tangential moment (per psi pressure)
Minimum tangential moment (per psi pressure)
85 psi
-80 p s i
80 in.-lb
-55 in.-lb
70 psi
-18 psi
180 in.-lb
-10 in.-lb
i
14 psi
-23 psi
13 psi
g
!
b. Stress i n the Pressure Shell Vessel
Maximum circumferential stress (per psi pressure)
Minimum circumferential stress (per psi pressure)
Maximum longitudinal stress (per p s i pressure)
Minimum longitudinal stress (per p s i pressure)
-75
psi
I
I
123
J
c. Summary of Pipe Stresses'
Line
No.
Pipe Size
(in.)
Maximum Cold
Prestress
(Ib)
1
Maximum Hot
Prestress
Temperatureb
Pressure
L
Ob)
(" F)
(PSid
a
Fuel Piping
111
112
113
114
115
116
117"
118
119
120
2
1 '/2
1%
1
1
1%
2
2
1%
2
1
3,850
11,230
24,300
15,400
38,000
28,200
2,170
6,450
13,900
8,850
19,600
14,500
1500
1500
1400
1500
1375
1325
28,000
22,000
15,350
12,800
11,3OO
7,900
1500
1325
1325
39
I34
'34
34
11
60
11
11
60
60
A
-
t.
6
I
i
b
t
&
k
Sodium Piping
303
304
305
306
307
308
3 09
310
313
24
1 and 2
1 and 2
1
2, 3
2
1 2, 3
2
2\
2
\
\
\,
\,
5,260
4,2 80
4,300
13,250
4,170
13,250
2,2 10
4,430
28,600
1050
1050
1050
1050
1050
1050
1050
1050
1325
3,690
3,000
3,020
9,310
2,935
9,3 10
1,550
3,110
16,400
58
58
,58
61
F
P
1
'All piping AS1 Schedule 40.
b N o t a l l temperatures and pressures given were reactor design point values.
= T h i s line consistedprimarily of pipe connections and was not stress analyzed.
3. Fluoride Pretreatment
Treat ment
Purpose
Hydrof I uor inat ion
Hydrogenat ion
Hydrofluor inat ion
Hydrogenation
Remove water
Reduce oxides and sulfates
Fluorinate oxides and sulfates
Remove HF, NiF, and FeF,
Time
(hr)
2
2
4
24 t o 30
Ternperdtur e
(W,
RT t o 700
700 t o 800
800
i
c
t
124
t
Appendix
C
CONTROL SYSTEMDESCRIPTION AND OPERATION’
F. P. Green, Instrumentation and Controls Division
d
1
1
4
i -
CONTROL SYSTEM DESIGN
The control system designed for the ARE had to
be able to maintain the reactor at any given power
and, when necessary, had to also be able t o overcome quickly the excess reactivity provided for
handling fuel depletion, fission-product poisons,
and power level increases.
The motor speeds
and gear ratios were set so that safety-rod withdrawal could not add A k / k a t a rate greater than
0.015%/sec.
The rods were designed to contain
15%A k / k , and the control system was designed to
l i m i t the fast rate o f regulating rod withdrawal to
be usad for high-power operation t o an equivalent
A k / k of O.l%/sec, with a maximum A k / k of 0.40%,
which i s one dollar, for steady-state fuel circulation. For the c r i t i c a l experiment and low-power
operation, the regulating rod withdrawal rate was
limited to an equivalent h k / k of O.Ol%/sec.
A fundamental distinction was made between
insertion and withdrawal of the shim rods. E i t h e r
the operator or an automatic signal could insert
any number of rods u t any time, whereas withdrawal
was always subject to both operator and interlock
permission. Withdrawal was limited, moreover, to
what was needed or safe under given circumstances.
Quantitative, rather than qualitative, indication
of control parameters was used where possible.
Where feasible, only one procedure of manual control was mechanically permitted to minimize timeconsuming arbitrary decisions on the operator’s
part.
The control philosophy o f paramount importance
i n the general operating dynamics was the use of
only absorber rods to control the nuclear process
and the use of only helium coolant to control the
power generation.
Most of the control and safety features o f the
ARE were similar t o those o f other reactors. Considerable care was exercised i n the design of
instrument components, control rods, and control
circuits to make them as nearly fail-safe as was
An instrumentation block diagram o f
practical.
the reactor control system i s shown i n Fig. C.1,
’ T h i s appendix was originally issued as O R N L CF-535-238, A R E Control S y s t e m D e s i g n Criteria, F. P. Green
(May 18, 1953), and w a s revised for this report on
March 7, 1955.
i n which the reactor i s shown to be subject t o
effects o f the control rods and to certain auxiliary
f a c i l i t i e s . The reactor, in turn, affected certain
instruments which produced information that was
transmitted to the operator and t o the control system.
The operator and the control system also
received information from indicators o f rod position
and motion. B y means o f the operator’s actions
and instrument signals, the control system transmitted appropriate signals to the motors and
clutches. The actuators located over the reactor
p i t i n the actuator housing, i n turn, affected the
reactor by corresponding rod motions which closed
the control loop. I n addition t o the above i n d i cations that were a t the operator’s disposal on
the console, there were many annunciators physically located along the top of the vertical board
i n the control room. The group o f eight annunciators labeled “Nuclear Instruments” were provided
t o indicate that conditions were improper for
raising the operating power level or that improper
operating procedures had caused safe l i m i t s to be
exceeded.
INSTRUMENTS DESCRIPTION
The reactor instruments discussed i n the following have been arbitrarily limited t o those directly concerned with the measurement of neutron
level and with f l u i d temperatures and flows associated with the reactor. The parallel-circular-plate
and compensated ion chambers, the fission chambers, and BF, counter were of ORNL design,
and, except for the latter two, were identical t o
those used in the MTR and LITR.2 Since the
fission chambers were located i n a high-temperature region within the reactor reflector, a special
chamber had t o be developed that could be helium
cooled. The compensated ion chambers and the
log N and period meters had a useful range o f
approximately lo6, and thus they indicated fluxes
of 10” n/cm2.sec (maximum) and lo4 n/cm2.sec
(minimum). The BF, counter had a maximum counting rate o f about lo5 n/sec and the fission chambers and count rate meters were capable o f covering
2S. H. Hanauer, E. R. Mann, and J. J. Stone, An O f f O n Servo for the A R E , O R N L CF-52-11-228 (Nov. 25,
1952).
125
a range o f 10” i f they were withdrawn from the
reflector as the flux increased. Since the count
rate meter has a long integrating time and the
resultant time delays would make it an unsatisfactory instrument for automatic control, it was
used as an adjunct t o manual control.
The instantaneous positions o f the four control
rods were reported to the operator by selsyns. I n
addition, the positions of the rods with respect t o
certain fixed mechanical limits were detected by
means of lever-operated microswitches. The I imitswitch signals t i e d i n with the control system and
with signal lights on the control console. The
upper l i m i t switches on the three shim rods operated when the rods were withdrawn t o about 36 in.
from the f u l l y inserted position.
Their exact
location was adjusted to provide optimum sensit i v i t y o f shim rod action. The lower l i m i t switches
served t o cut the rod drive circuits when the
magnet heads reached the magnet keepers.
The seat switches operated indicating I ights on
the console, which served only to indicate the
proximity o f the pneumatic shock absorber piston
t o the lower spring shock absorber.
The lights
told the operator the immediate effectiveness o f a
# I
scrammed” or dropped rod and, hence, that the
rod had not iammed on i t s f a l l into the reactor
core.
The regulating rod was equipped with two travel
l i m i t switches whose actions tied i n intimately
with the manual and automatic regimes of the
servo-control system, which i s described i n a
following section.
Thermocouples were located a t many points on
the reactor and the heat exchangers t o monitor the
fuel temperatures, since temperature extremes or
low flow rates could have resulted i n serious
damage to the reactor.
The temperatures were
recorded, and electrical contacts i n the recorders
interlocked with the scram circuit t o provide shutdown i f the reactor i n l e t temperature dropped below
11W0F or the outlet temperature exceeded 155OOF.
The operator was also warned by an annunciator
alarm of low helium pressure in the rod-cooling
system because loss o f cooling would have increased the danger o f a rod jamming due t o mechanical deformation.
CONTROLS DESCRIPTION
Motion o f the shim rods or the regulating rod was
obtained from two energy sources, gravitational
and electromechanical.
Rod motion by safety
126
action was obtained by gravitational force upon
release of the shim rods from their holders as a
result o f de-energizing the electromagnetic clutches.
The three shim rods, each .with i t s magnet, were
subiect to a number o f different possible sources
of safety signal. The necessary interconnecting
circuitry was centered about an “auction”
or
1 6 sigma”
bus. Instrument signals were fed to this
bus by sigma amplifiers. The sigma bus may be
said t o go along with the “highest bidder” among
the grid potentials of the sigma amplifiers. T h i s
t
I
F
F
important auction effect allowed a l l rods t o be
If a l l three
dropped by any one safety signal.
amplifiers and the three magnets had been identical i n adjustment and operation, a l l three rods
would have dropped simultaneously as the sigma
bus potential rose or f e l l past some c r i t i c a l value.
Usually, only one of the rods dropped initially,
and the others were released by interlocked relay
action.
Two kinds of scrams were possible:
“fast”
scrams caused by amplifier action and ‘ ‘ S I O W ~ ~
scrams caused by interruption of magnet power as
the result of the action o f relays. I n a fast scram
the shim rods were dropped when the reactor level
reached or exceeded a specified flux level, as
determined by the safety chambers. The signal
from the safety chambers was amplified by the
sigma amp1 i f i e r s and subsequently caused the
magnet current t o become low enough t o drop the
rods. The slow scram was the actual interruption
o f power to the magnet amplifiers by relay action.
T h i s relay action could be caused by temperatures
i n excess of safe limits or stoppage o f fuel flow.
The shim rod drives were 3-wire, 115-v, instantaneously reversible, capacitor-run motors rated
at 1800 rpm. The shim-rod head, or actuator assembly, was a worm gear driven through a gear
The
reducer at slightly less than 234 in./min.
sequence o f shim rod scram, rod pick-up, and withdrawal required a minimum o f 25 min.
Both individual and group activation o f shim rods
were selected i n preference to group control alone
t o permit increased (vernier) f l e x i b i l i t y i n approaching criticality. Group shim rod action would
have made d i f f i c u l t the control of flux shading
throughout the reactor, particularly on rod insertion,
since, upon cutting the motor power, the friction
characteristics o f the three drives would have
caused the rods t o coast t o different levels.
The regulating rod drive consisted o f two Diehl
Manufacturing Company, 200-w, instantly reversible,
k
I
c
t
t
L;
6
h
i
t
&
!+
e
b
L
=
f
t
R
c
L
-
m
m
I g
-
q
U
i
4
E
two-phase servomotors that operated from reversing
contactors which received their actuation from a
mixer-pilot relay located i n the servo amplifier.
Only one o f the motors operated a t a time, depending on whether the temperature or flux servo
was coupled into the control loop. For the flux
servo operation, an additional speed reducer was
used so that the speed o f the rod was one tenth
that avoi lable for temperature servo operation.
I n i t i a t i o n of the flux servo “auto” regime was
contingent on reactor power great enough to give
a micromicroammeter reading of more than 20
and a servo-amplifier error signal small enough so
that no control rod correction was called for. With
the servo on auto the regulating rod upper or lower
l i m i t switches actuated an annunciator. Although
not used i n the experiment, i n i t i a t i o n of the temperature servo auto regime was contingent on
reactor power greater than a minimum controllable
power (>2% N F ) and a servo-amplifier error signal
small enough that the servo would not c a l l for rod
motion.
The rod-cooling system consisted of a closed
loop for circulating helium through the annuli
around the shim rods, the regulating rod, and the
f i s s i o n chambers and then through three parallel
he1ium-to-water heat exchangers. The he1ium was
circulated by two 15-hp a-c motor-driven positivedisplacement blowers. The motors were controlled
from the reactor operating console.
The fuel and reflector coolant temperatures were
also controlled from the console.
The helium
coolant removed heat from fuel and sodium system
heat exchangers and passed it to an open-cycle
he1ium-to-water heat exchanger, from which the
water was dumped. The helium i n the fuel system
was circulated by an electric-motor-driven fan that
was controlled from the console. The helium i n
the sodium system was circulated by two hydraulicmotor-driven blowers for which only the on and o f f
controls were located on the console.
In order that the reactor operation would be as
free as possible from the effects of transient
disturbances or outages an the purchased-power
lines, a separate electrical power supply was
available for the control system.
The supply
consisted of a 250-v, 225-ampI storage-battery bank
of 2-hr capacity. The bank was charged during
normal power source operation by a 125-kw motorgenerator set.
T h i s emergency d-c supply fed
emergency l i g h t s and a 25-kw motor-generator set
t o supply instrumentation power (120/208 v, single
phase alternating current) during purchased-power
outages.
An auto-transfer switch provided an
automatic switchover transient o f several cycles
upon loss of the purchased-power source.
The
auto-transfer switch automatically returned the
system to normal power 5 min after resumption o f
purchased-power supply.
The system could be
returned to normal power manually at any time after
resumption of the purchased-power supply.
c
C O N S O L E A N D C O N T R O L BOARD D E S C R I P T I O N
The control console was made up o f two large
panels, one to the right and one t o the l e f t o f the
operator’s chair, and eight subpanels directly i n
front of the operator, Fig. C.2. Over the top of the
console, the operator also had a full view of the
vertical board of instruments which indicated and
recorded nuclear parameters and pertinent process
temperatures. The rotary General E l e c t r i c Company
type SB-1 switches used were of the center-idle,
two- and three-position variety. They were wired
so that clockwise rotation produced an “increase”
i n the controlled parameter; such operation has
been described as potentially dangerous. However,
the scram switch, an exception, produced a scram
i n either direction o f rotation.
The top row o f the left-hand panel o f the console
contained the source-drive control (Fig. C.2). I n
the second row the f i r s t two switches were the
controls for the two-speed rod-cool ing he1ium
blowers, and the third switch was the on-off control for the electric motors which were the prime
movers for the hydraul ic-motor-driven blowers. In
the bottom row were the group control for the three
shim rods and the annunciator acknowledge and
reset pushbuttons.
The center eight subpanels were identical i n s i z e
and shape and contained the s i x selsyns which
indicated the rod and fission chamber positions
and the two servo control assemblies. The temperature c o l i brated potentiometer located near the
panel center was used to set the fuel mean temperature for automatic, servo-controlled operation.
The subpanel on the extreme l e f t contained the
flux vernier (not shown) used to set f l u x demand
into the f l u x servo system for controlled operation.
The d i a l was calibrated t o correspond t o a setting
o f from 20 to 100 on the micromicroammeter Brown
recorder. Associated with a l l controls were the
necessary limits-of-travel indicating lights, along
with the on-off and speed-range indicators.
The right-hand panel contained the switches most
129
i
t
closely associated with the fuel system and the
energy removal controls.
I n the upper row were
the switch for the fuel-loop helium-blower prime
mover, which was a 50-hp electric motor, the scram
switch, and the by-pass switch (not shown on
Fig. C.2). The by-pass switch was provided to
permit the fuel-loop prime mover to be run, even
though the reactor power was low, to remove the
afterheat that was expected to be present after a
shutdown. The bottom row on the right-hand panel
contained the fuel-system-he1ium blower speed
cantrol, the selector switch for barrier-door control,
and the barrier-door drive control.
The four panels i n the center of the vertical
board contained the 12 recorders which pertained
to the reactor. They were, l e f t to right, top row:
reactor i n l e t temperature, reactor AT, reactor outl e t temperature, reactor mean temperature; center
row: count rate No. 1, p i l e period, safety level
No. 1, and micromicroammeter. I n the bottom row
there were: count rate No. 2, log N, safety level
No. 2, and control rod position.
The range of flux through which the reactor
passed from the insertion of the source to f u l l that
power operation was so great, more than
TABLE C.1.
several classes of instruments were used.
The normal flux (NF) was arbitrarily designated as 1, and then the first, and lowest, range
encountered was the source range, from somewhat
less than
to lo-’’; the second was the
counter range, from 10”’
to
the third was
to 3.3 times
and
the period range, from
to 1.
the last was the power range, from 3.3 x
A t greater than
i n the period range, the servo
system could be put into auto operation i f desired.
CONTROL OPERATIONS
I n the operation of the ARE, operator i n i t i a t i v e
was overridden by two categories of rod-inserting
action:
scram, the dropping of safety rods; and
reverse, the simultaneous continuous insertion of
a l l three shim rods, which was operative only
during startup. The fast scram was i n a class by
itself, being the ultimate safety protection of the
reactor, and could never be vetoed by the operator.
The various occasions for automatic rod insertion
and annunciation for nuclear trouble are l i s t e d i n
Table C. 1.
There were five interlocks between nuclear reactor control, fuel and moderator coolant pumping,
CAUSES O F AUTOMATIC ROD INSERTION AND ANNUNCIATION
F a s t Scram
NF)
Neutron l e v e l
(1.5
Neutron l e v e l
(1.2 NF)
1-sec period
Slow Scram
Reverse
X
X
X
X
X
5-sec period
X
F u e l temperature
(> 155OOF)
Annunciation
X
X
X
Loss o f fuel flow (power >10 kw)
X
X
Loss o f control bus voltage
X
X
Loss o f purchased power
X
X
Manual scram
X
Reactor power
(AT
F u e l temperature
> 40OoF)
((1 100°F)
X
F u e l helium blowers on without reflector helium blowers on
X
Servo o f f range
X
Rod coolant helium o f f
X
Count r a t e meter o f f scale
X
Safety circuit trouble
X
130
-
1
,
t
n
I
i
-
and helium-cooling operation. (1) When operating
a t power (above 10 kw), loss o f fuel flow would
have initiated a slow scram. (2) Loss o f fuel flow
would have, likewise, removed a permissive on
operation o f the fuel-loop helium-blower prime
mover. (3) Excessive deviation o f fuel temperature
from set point would have produced a scram for
either too high or too low a temperature. L o w fuel
temperature would also cause the helium p i l o t
motor t o run the magnetic clutch control for the
helium blowers t o the zero position and thereby
range, prime movers operating, reactor i n power
range, and power greater than 10 kw. The operator
was notified by the "permit"
l i g h t when these
conditions were satisfied. A p i l o t decrease was
not interruptible by the operator u n t i l the cause o f
the automatic action was alleviated.
The elementary control diagram i s shown i n Fig.
C.3. The various relay and l i m i t switches referred
t o i n Fig. C.3 are described i n Tables C.2 and
C.3, respectively. These tables and Fig. C.3 show
the procedural features of the control system as it
applied t o operator initiative.
T h i s system was
divided into channels, each consisting o f one or
more relays actuated by the operator subsequent
t o properly setting up interlock permissives. The
interlocks often depended upon the aspects o f
relays in other channels. Primarily, the channels
corresponded t o a mechanical unit involving a rod
control, a process loop control, or an instrument
system control. The channels were:
stop the blowers and thus stop fuel cooling. (4)
Moderator coolant flow was interlocked with the
moderator coolant helium blowers t o prevent helium
circulation when the sodium flow was low or
stopped. A slug of cold sodium would have entered
the reactor when circulation was restarted i f t h i s
precaution had not been taken. (5)Permissives on
increasing he1ium flow were fuel temperature i n
D
1. T w o f i s s i o n chamber d r i v e s
2.
u.
Manually actuated.
b.
T r a v e l l i m i t switches, p a n e l - l i g h t indicated.
Servo control system
(2.
2% N F , for temperature servo,
20 on the micramicroammeter recorder
Auto i n i t i a t e d manually by permission of power greater than
and power great enough t o g i v e a reading o f more than
for any scale range, for f l u x servo, and a dbmand error c l o s e t o zero (neither l i g h t burning)
for either serva system.
b. Auto regime sealed-in l i g h t i n d i c a t e d i f i n i t i a t e d w i t h proper c o n d i t i o n s and power remained
greater than that s p e c i f i e d in
C.
2u
above.
Manual regime regained b y manually dropping out seal or b y manipulation of control rod
switch.
d.
Reverse interrupted rod withdrawal b y any means and caused rod insertion.
e . Manual control rod overrode automatic regime and dropped out seal.
1.
P e r m i t s an manual rod control required t h a t either the reactor to be a t a power l e v e l greater
than
J
1
N or one count r a t e meter be "on scale.''
F
3. Slow scram
a. A series-parallel r e l a y system was used
t o i n i t i a t e a scram. T h i s system was used because
a large number of series c o n t a c t s would have been i n d u c i v e t o f a l s e drop-outs and hence
f a l s e scrams.
The three most urgent c r i t e r i a c a l l i n g for scram had to be f a i l - s a f e and were
therefore placed i n the series-connected section.
b. R-16 was i n a c t i v e i n the operating regime; R-17 and R-18 were n o r m a l l y actuated and were
responsive t o manual scram, scram reset, fuel temperature extremes, and
4.
Reverse, R-23 and R-24, p l u s an indicator l i g h t
a Operated all three rods i f the by-pass s w i t c h was i n "normal"
R-16.
and
(1) the group insert s w i t c h was closed, or
(2) a slow scram occurred (equivalent to a scram follower), or
(3) the servo system was i n the manual regime; the reactor power wos l e s s than 10 kw; and
a period o f l e s s than
On i n i t i a l startup,
i
5
seconds occurred.
No. 4 3 ) i n i t i a t e d the reverse at any power l e v e l (due
t o jumper-shorted
133
1
contact) u n t i l a n e g a t i v e temperature c o e f f i c i e n t o f r e a c t i v i t y was proved.
b.
5.
T h e by-pass s w i t c h prevented roverse (but n o t slaw scram) when i n "by-pass"
position.
Shim rad withdrawal
a
P e r m i s s i v e s were:
(1) no reverse i n progress,
(2)
reading a n - s c a l e on a t l e a s t one count r a t e meter ar neutron l e v e l greoter than
-5
10
NF*
(3) no manual i n s e r t i o n i n progress,
(4) rod heads n o t r e s t i n g on upper travel limits.
6.
f
Shim r o d insert
if
a
Permitted
b.
Actuated i n d i v i d u a l l y and manually.
heads n o t on lower travel limits.
c. Actuated together b y any o c t i o n operating reverse.
7.
Fuel-system helium-blower prime movers
a. T h e prime movers could be started by 5-13 i f
(1)
(2)
c l u t c h r e l a y s were actuated (rods "hung"),
or
if
by-pass s w i t c h
S-5 was on by-pass,
helium blbwer speed l o w e r l i m i t s were closed, and
(3) fuel f l o w was n o t below 20 gpm.
b.
T h e prime movers were a u t o m a t i c a l l y turned o f f
dropped b e l o w
20
if
a shim r a d was dropped or fuel f l o w
0
gpm.
c. R e l a y R-38 started the prime movers by a c t i v a t i n g the starter r e l a y RF-1 l o c a t e d i n t h e
basement.
8.
H e l i u m p i l o t c o n t r o l s (power-loading system)
Permit i n d i c a t o r s on power increase were lit i f
(1) the reactor l e v e l was greater than 10 k w or t h e by-pass s w i t c h was set on by-pass,
(2) the prime movers were running, and
(3) the fuel temperature was above llOO°F.
b. Power increase from the c o o l a n t system could be demanded i f
(1) t h e permit l i g h t s were o n and
(2) p i l o t switch, S-14, was thrown t o "increase.'*
c. I f the power decreased and the c o n t r o l s were n o t already a n shut-off l i m i t s , t h e h e l i u m
a
c i r c u l a t i o n would be shut o f f .
(1) monually by u s i n g s w i t c h 5-14, or
(2)
a u t o m a t i c a l l y by whichever system suffered from l o w f u e l temperature, from prime mover
cut-aut,
or from l o w power
(<lo
kw),
since by-pass s w i t c h c o n t a c t
during normal c r i t i c a l and power operation.
S5-3 was c l o s e d
L o w fuel temperature i n t e r l o c k s automati-
c a l l y decreased the corresponding h e l i u m p i l o t c o n t r o l s
if
the fuel dropped b e l o w t h e
-t
c r i t i c a l temperature at e i t h e r o f the t w o heat exchangers.
9. T h e by-pass s w i t c h f u n c t i o n was somewhat obscure s i n c e it occurred i n the prime movers, p i l o t
controls, and reverse c i r c u i t s . Its i n c l u s i o n r e s u l t e d from t h e p o s s i b i l i t y o f
requiring
additional
fuel c o o l i n g f o l l o w i n g a shutdown and after an e x t e n s i v e running period at power level.
large gamma heat source, p r i m a r i l y i n the moderator (because
of
A
i t s v e r y l a r g e heat capacity),
c a l l e d the afterheat, remained in t h e reactor immediately f o l l o w i n g power operation.
B y u s i n g both hands, t h e operator could throw and h o l d
S5
on by-pass w h i l e h e r e s t a r t e d t h e
prime movers and subsequently increased t h e helium p i l o t c o n t r o l s t o such a p o i n t t h a t the
temperature l c v e l e d off.
T h e by-pass c i r c u i t s permitted t e s t i n g o f c o o l i n g system c o n t r o l c i r c u i t s and p r e s e t t i n g o f
c o o l i n g system c o n t r o l l i m i t s w h i l e the reactor was shut down and a l s o permitted the v i t a l
s u b c r i t i c a l test far fuel temperature c o e f f i c i e n t o f reactivity.
3Wm.
134
B.
C o t t r e l l and
J. H.
Buck,
3
ARE Hazards Summary Report, ORNL-1407
(Nav.
20, 1952).
i.
T h e use of the manual reverse cut-out,
S5-6, operated by throwing the by-pass switch, was
divorced from the above Considerations and devoted only to t e s t operation of the shim rod
motors at shutdown, a t low power, or with the magnet amplifiers turned off.
10. Source drive
a. Manual I y actuated.
b. T r a v e l l i m i t switches, panel-light indicated.
REACTOR OPERATION
I
3
I
J,
J
#
P
.
The preliminary operating plan for the ARE was
described i n 0RNL-1844.4 I n the check operations,
considerable time was spent in loading the inert
fuel carrier and i n the subsequent shakedown run.
Critical loading, subcritical measurement o f fuel
temperature coefficient, regulating rod calibration,
zero-power operation a t 1200 t o 13OO0F, and power
operation are described in Appendix D.
In the starting o f a freshly loaded reactor the
operator i s most concerned about a reliable neutron
signal. Attention must therefore be given t o assuring that the fission chambers are in their most
sensitive positions, and that they are giving a
readable signal on the scalers or count rate meters.
It i s imperative that the source be inserted for the
earliest indication of subcritical multiplication t o
be seen during the f i r s t startup. Use of the source
i n subsequent experiments becomes less important
as the history o f the reactor operation builds up.
For a normal, intentional shutdown the operator
had the c h o i t e o f a variety o f procedures.
For
every scram, however, the interlocks of the system
were such that the helium flow i n the fuel-to-helium
heat exchangers was cut o f f by opening the electrical circuits of the motors. T h i s action was necessary t o prevent the fuel from freezing i n the heat
exchangers.
The various slow scrams included
the following: (1) maximum reactor outlet temperature, (2) minimum reactor inlet temperature, (3) a
1-sec period, (4) N , greater than 3.0 Mw, (5) fuel
flow below 20 gpm. A 5-sec period inserted the
shim rods, which decreased A k / k a t a rate o f
O.O15%/s/ec,
and operated only during l o w l e v e l
startup.
In addition t o the process signals which initiated
slow scrams, there were several other potential
process malfunctions for which it might have been
desirable t o scram the reactor.
However, the
action t o be taken i n these situations was l e f t up
t o the decision o f the operator. There were two
reasons for not putting these process malfunctions
4Design and Installation of the Aircraft Reactor E x periment, ORNL-1844 (to be published).
on automatic scram. First, sensory signals usually
lag the event t o such an extent that a fast scram
cannot provide better protection than a delayed
scram. Second, reaction t o the signals requires
limited judgment.
An annunciator alarm signal
was provided to draw the operator’s attention t o
the possible difficulty for each case that would
have required limited judgment. The cases were
the following: (1) pronounced changes i n differential
temperatures between reactor outlet tubes, (2)
lowering level i n any surge tank, (3) r i s i n g level
i n any surge tank, (4) alarm from a p i t radiation
monitor or monitron, (5) high oxygen concentration
or humidity in the pits.
The operation o f placing the “position”-type
regulating rod servo system on automatic control
for
i n any part of the power range above 2% N,,
temperature servo, and a micromicroammeter recorder reading o f greater than 20 on any scale, for
flux servo, was arranged t o require that the reactor
t o be on a stable, i n f i n i t e period. I n the normal
course o f events, the rod should have been manually
set at mid-range. Six d-c voltage signals were fed
into the servo amplifier, three from the servo ion
chamber, i n l e t temperature recorder, and outlet
temperature recorder, one from the v o k e supply
adjusted by the “temperature demand” potentiometer, and two from the flux demand and micromicroammeter recorders. The operator therefore had a
unique temperature setting for each level of power
operation that produced “zero error” i n the amplifier output, and hence no “error”-light
indications
and no demand for rod movement. It was a t t h i s
setting that the servo “auto”
regime could be
initiated. A small change in reactor mean temperature or flux level could be effected i n the auto
regime by resetting the temperature demand or flux
demand potentiometers, but the power level could
not be changed.
NO clear l i n e could be drawn between normal
operation and operation under d i f f i c u l t y . A few o f
the situations in which the operator would have
expected to feel more than average need for attention t o instrument signals and control actions were
loss or interruption o f power, automatic shutdown
135
1
I
4
!
!
N
o n
c +
v)
n
m
u
2r
'
-
m
p
A
a"
/,
N
a"
-
-
w "
A
0
-a "
A
v
29(9)
"P
*a
A
*a v
8
x "
A
A
p
f
PV
P
L"
b-?
I
r
v)
;
r
I,
,,
'
N
N
m
a v
A
a v
-d
-
75
347
346
-345
-
,v
, y
a
CONTROL
TRANSFER
D
cn
P
a
,*
N
01
P
CONTROL
TRANSFER
-
CONTROL
TRANSFER
-a
a
-a
73
m
I
w
w
3
va
Ov)
N
N
0
N
N
m
n
1
i;
440V
CONTROL
TRANSFER
5;;
y -
a-C
;
3r;]741
j
440V
440v
440"
200
344
315
343
4
?
a
,I
,c
F
0
e
E
,r
P
n
314
342
a
P
187
186
I85
184
183
I82
li
L
L
t
I
k
c
t1
I
k
L
k
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I
t
i
1
t
iI
1
E
i
t
TABLE C.2.
LIST O F RELAY SWITCHES
~~
Contact
~
~
~~~
~~
Set Point
Instrument
.
J
1
i
J
J
RS-9
Count rate recorder No.
1
Closed above
RS-10
1
Opened o f f s c a l e
RS-1 1
2
Closed above
RS-12
2
Opened o f f s c a l e
RS-13
Front sodium flow
Closed above
60 gpm
RS-14
Bock sodium flow
Closed above
60 gpm
RS-15
Log N
Closed when N
RS-16
F u e l heat exchanger No.
2 outlet
Closed above
1150°F
RS-17
2 outlet
Closed above
1100°F
RS-18
Fuel flow
Closed above
20 gpm
RS-19
Not used
RS-20
Lag N
RS-2 1
Log N
RS-22
Log
RS-23a
Safety l e v e l No.
1
Opened above
120
RS-23b
Safety l e v e l No.
2
Opened above
120
RS-24
Fuel flow
Opened above
20 gpm
RS-25
N o t used
RS-26
Reactor o u t l e t temperature (mixed manifold)
Opened above
155OoF
RS-27
N o t used
RS-28
Not used
RS-29
P i l e period
Closed a t l e s s than
RS-30
M i cromi croammeter
Closed above
RS-31
Count rate recorder No. 1
Closed above 1 count/sec
RS-32
Count rate recorder
RS-33
RS-34
RS-35
P i l e period
RS-3 6
R5-40
AT across reactor
AT across reactor
AT across reactor
AT O C ~ O S Sreactor
AT across reactor
RS-4 1
AT
RS-42
N o t used
RS-43
N o t used
RS-37
RS-38
RS-39
> 2% N F
> 10’
Closed when N >
Closed when N > 6.67
No. 2
-
1 count/sec
Closed when N
N
__
1 count/sec
NF
NF
x
NF
5 sec
20 ,UP
Closed above
1 count/sec
1 outlet
Closed above
115OoF
1 outlet
Closed above
llOO°F
Opened a t l e s s than
tube No.
1
Closed above
45OoF
tube No.
2
Closed above
45OoF
tube No.
3
Closed above
45OoF
tube No.
4
Closed above
45OoF
tube No.
5
Closed above
45OoF
across reactor tube No.
6
Closed above
45OoF
5 sec
,
1
139
F
TABLE C.2 (continued)
Contact
Set P o i n t
Instrument
RS-44
Not used
RS-45
AC-to-DC M-G
RS-46
Not used
R S-47
N o t used
RS-48
Not used
RS-49
Not used
R S-50
F u e l flow
Closed above 20 gpm
RS-5 1
Fuel f l o w
Closed above 20 gpm
RS-52
Reactor outlet temperature (mixed manifold)
Opened above 150OoF
RS-53
Wind v e l o c i t y recorder
Wind above 5 m i l e s h r or vent gas monitors
RS-54
Fuel heat exchanger No. 1 outlet
Closed above 115OoF
RS-55
Closed above llOO°F
R 5-56
Fuel heat exchanger No. 2 o u t l e t
Closed obove 1 15OoF
RS-57
Closed above 1 100°F
R 5-58
Main fuel
Closed above 4
RS-59
F u e l heat exchanger No. 2
R S-60
Secondary fuel pump low-water-flow
R 5-6 1
Back sodium heat
RS-62
Front sodium flow
Closed above 100 gpm
RS-63
Back sodium flow
Closed above 100 gpm
R 5-64
H e l ium
RS-65
Closed when set was charging batteries
set
pump
low-water-flow
alarm
alarm
exchanger outlet
t
F
gpm
i
Closed above 4 gpm
Closed obove llOO°F
i
Closed
at l e s s
P i t humidity indicator
Opened
at
> l o % relative humidity
RS-66
Helium supply heoder pressure
Opened
at
<500 p s i g
RS.67
Water reservoir level
Opened
at
< 1 6 p s i g (<18,000
RS-58
M a i n sodium pump tank l e v e l
Opened
a t l o w level
RS-69
Main sodium pump tank l e v e l
Opened
at
RS-70
Standby sodium pump tank level
Opened
at low level
RS-71
Standby sodium pump tank l e v e l
Opened
at
RS-72
M a i n fuel pump tank level
Opened
at low
RS-73
M o i n fuel pump tank l e v e l
Opened
at
RS-74
Standby fuel pump tank l e v e l
Opened
a t l o w level
RS.75
Standby fuel pump tank level
Opened
at
RS-76
Water controller
for front sodium heat exchanger
Opened above 16OoF
R S-77
Water controller
for bock sodium heat exchanger
Opened obove 16OoF
RS-78
Wate: controller
for fuel heat exchanger
Opened above 160°F
R S-79
Standby sodium pump
concen trati on
recorder
motor ammeter
c
t
i
Closed above 1 100°F
outlet
tk
than 90% He
t
gal)
b
high l e v e l
h i g h level
level
high l e v e l
high level
Opened above 50 amp
k
f
i
T A B L E C.2
Contract
i
j
.
(continued)
Instrument
Set P o i n t
RS-80
Water c o n t r o l l e r for space coolers
Opened above
160' F
RS-8 1
Water c o n t r o l l e r for rod-cool i n g heat exchanger
Opened above
160'F
RS-82
l n d i v i d u o l heot exchanger tube temperature
Opened above
150OoF
RS-83
I n d i v i d u a l heot exchanger tube temperature
Closed above
1 150'F
RS-84
M a i n sodium pump motor ommeter
Opened above
50
RS-85
AT
Opened above
400'F
RS-86
Main fuel pump motor ammeter
Opened above
50
RS-87
AT
Opened above
4OO0F
RS-88
Standby fuel pump motor ammeter
Opened above
50 amp
RS-89
AT
Opened above
400'F
RS-90
Standby sodium pump low-woter-flow olorm
Closed above
4
RS-9 1
AT
RS-92
1
amp
i
4
!
1
,
1 .
I
i
c
i
ocross reactor tube No.
2
3
amp
gpm
Opened above
4OO0F
M a i n sodium pump low-water-flow alarm
C l o s e d above
4 gpm
RS-93
AT
Opened above
400'F
R S-9 4
M a i n fuel pump l o w - o i l - f l o w alarm
C l o s e d above
2 gpm
RS-95
AT
Opened above
400'F
R S-9 6
Standby fuel pump l o w - o i l - f l o w alarm
Closed above
2 gpm
RS-97
Standby sodium pump l o w - o i l - f l o w alarm
Closed above
2 gpm
RS-98
M o i n sodium pump l o w - o i l - f l o w alarm
C l o s e d above
2 gpm
RS-99
Sodium o u t l e t pressure a t reactor
Closed above
RS-100
F u e l i n l e t pressure at reactor
Opened above
across reactor tube
4
No. 5
across reactor tube No. 6
48 p s i g
48 p s i g
RS-10 1
M a i n sodium pump tank pressure
Opened above 47 p s i g
2
RS-102
Standby sodium pump tank pressure
Opened above
1
RS-103
Standby fuel pump tank pressure
Opened above 5 p s i g
RS.104
M o i n fuel pump tank pressure
Opened above
5 psig
RS-105
Rod c o o l i n g helium pressure
C l o s e d above
13 p s i g
4
RS-106
F u e l loop water flow
C l o s e d above
110 gpm
1
RS-107
Water f l o w to fuel and sodium pumps
Closed above 10 gpm
RS-108
Rod c o a l i n g water flow
C l o s e d above 15 gpm
RS-109
F r o n t sodium l o o p water f l o w
C l o s e d above
RS-110
Back sodium loop water f l o w
C l o s e d above 65 gpm
RS-1 11
Space cooler water flow
C l o s e d above
45 gpm
RS-112
V e n t header vacuum
Opened b e l o w
29 in. Hg
RS-113
Reserve nitrogen header s u p p l y pressure
Opened a t
<300 p s i g
RS-114
Rod c o a l i n g helium
Opened a t
(40
RS-115
DC-to-AC
1
M-G
set
47 p s i g
65 gpm
psig
C l o s e d when s e t was running
141
4
T A B L E C.2 (continued)
Contact
Set P o i n t
Instrument
RS-116
AC-to-DC M-G set
Closed when set was running
RS-117
Heat exchanger p i t radiation monitor
Opened
at
high level
RS-118
Heat exchanger p i t radiation monitor
Opened
at
high level
RS-119
Vent gas monitor
Opened
at
high l e v e l
RS-120
Vent gas monitor
Opened
at
high l e v e l
RS-121
Sodium
Opened
at
<6OoF
Ar o t
reactor
T A B L E C.3.
LS-1
Fission chamber No. 1 lower l i m i t
LS-2
Fission chamber No. 1 upper l i m i t
F i s s i o n chamber
:”
No. 2 lower l i m i t
_
~
~
~
Closed
on l i m i t
Closed on l i m i t
Closed
on l i m i t
on l i m i t
L s-4
F i s s i o n chamber No. 2 upper l i m i t
Closed
L s-5
Regulating rod lower l i m i t
Closed on l i m i t
L 5-6
Regulating rod upper l i m i t
Closed
LS-7
Shim rod No. 1 upper l i m i t
Closed on l i m i t
L s-8
Shim rod No. 2 upper
limit
Closed
on l i m i t
L s-9
Shim rod No. 3 upper l i m i t
Closed
on l i m i t
LS-lo
Shim rod No. 1 lower l i m i t
C l o s e d on l i m i t
LS-11
Shim rod No. 2 lower l i m i t
Closed
LS-12
Shim rod No. 3
Closed on l i m i t
LS-13
Shim rod No. 1 clutch
Closed when rod
was
attached
L S-14
Shim rod No. 2 clutch
Closed when rod
was
attached
LS-15
Shim rod No. 3 c l u t c h
Closed when rod
was
attached
LS-16
Shim rod No. 1
seat
Closed
LS-17
Shim rod No. 2
seat
Closed on seat
LS-18
Shim rod No. 3
seat
Closed
action, and
shutdown.
buildup o f
xenon
lower l i m i t
poison after a
A s far as loss of power i s concerned, the operator
would have been l e f t figuratively, and nearly
literally, i n the dark were the instrument bus and
control bus t o go dead simultaneously.
In this
very unlikely occurrence, a l l the instrument lights
and a l l relay-indicating lights would have been
out, leaving the operator without visual knowledge
of whether the scram (automatic on loss o f power)
142
: i
Set P o i n t
_
L5.3
- i
t
L I S T OF L I M I T SWITCHES
Device
Conto c t
b
on l i m i t
k
on l i m i t
on seat
on s e a t
was effective. If he had reasonable f a i t h i n the
law of gravity, he would probably have remained a t
h i s post u n t i l power was restored. There would
have been no normal scram indication i f the relay
cabinet main fuse were to have opened. However,
since the amp1i f i e r power cabinet would have
operated from emergency power, there was a check
on release of the shim rods i n the fact that the
magnet currents would have been interrupted.
Whenever the reactor was subiect t o automatic
e
shutdown action, the
vided between the
chances of returning
ation.
The circuits
operator’s concern was dicausative effects and the
the reactor to normal operprevented the operator from
, interfering with shutdown actions as long as the
causative conditions remained. Hence the operator’s reaction t o anything less drastic than a
scram was t o clear the responsible condition.
When the reactor was scrammed, the chances o f
getting back i n t o normal operation were o f course
dependent upon the time history o f previous operation as well as upon the need for remedying the
trouble. However, any length o f complete shutdown
could have been tolerated after prolonged operation
a t normal flux, since xenon poison buildup was
low.
4.
4
rl
J
J
143
Appendix D
NUCLEAR OPERATING PROCEDURES~
The importance and critical nature G / the program, a s well a s the short time scheduled for tbe
operation of the reactor and system, necessitated a tight experimental program and p r e c i s e
operating procedures. The nuclear operating pmcedures were, of course, prescribed in advance
of the experiment It i s of considerable interest to note that the actual experimental program
followed the anticipated pmgram very c l o s e l y thmughouf The only significant deviations were
the inclusion of some additional experiments a s time permitted. The following appendix i s a
verbatim copy o f the pmcedures by which the reactor w a s operated Some minor discrepancies
-
.
1300° F
1. The fuel loading system w i l l be operated with
A D D I T I O N O F FUEL
b
-
-
0.0013 T
(“c)
.
At a temperature o f 13OO0F, the density i s
4.59 g/cm3, or 287 lb/ft3. I n each pound of the
fuel concentrate there i s 0.556 Ib o f U235.
One quart of the concentrate weighing 9.59 Ib
and containing 5.33 Ib o f U235w i l l be forced i n t o
the transfer tank and weighed. With the safety
carrier.
The fuel storage tank w i l l be installed
and carrier forced into the transfer tank and then
into the reactor system. The strain gage weighing
devices w i l l be checked.
2. T h e helium blowers w i l l be operated and the
resultant drop i n mean temperature observed. The
operating crew w i l l practice in handling the helium
flow so as to drop the mean temperature a t a rate
of 10°F/min and a t a rate of 25’F/min.
Care w i l l
be exercised not t o drop the temperature o f the
fuel carrier t o below 1150°F as it leaves the heat
exchangers. A s t h e latter temperature approaches
that value, the helium flow w i l l be reduced and
the system brought back t o i t s mean operating
temperature o f 1300°F.
3. With the helium blowers o f f and starting with
fuel carrier a t rated flow o f 34 gpm and mean
temperature o f 13OO0F, the f!ow rate w i l l be
decreased i n steps t o about one-third rated value.
A l l temperatures w i l l be observed t o note any
spurious changes caused by the decreased flow.
rods completely withdrawn and the neutron source
and fission chambers inserted, the fuel concentrate w i l l be forced into the fuel circuit. At t h i s
time the carrier flow w i l l be 34 gpm and the
Concentrate and carrier temperatures w i l l be
1300OF. The scram level on the safety chambers
w i l l have been set a t about 10 k w and the period
scram a t 1 sec. A BF, counter w i l l have been
installed temporarily in place o f the neutron ion
chamber for the temperature servo control, and the
regulating rod w i l l be on slow-speed drive.
The system volume up to the minimum operating
level i n the fuel pump i s calculated t o be 4.64 ft3.
Assuming a c r i t i c a l mass o f 30 Ib in the 1.3 ft3 o f
reactor core, there w i l l be approximately 124 Ib o f
U235 in the fuel circuit when the reactor f i r s t goes
c r i t i c a l with the shim rods completely withdrawn.
Accordingly, it i s anticipated that approximately
0.78 ft3 or 23.4 quarts of fuel concentrate must
be added for i n i t i a l c r i t i c a l i t y . Twelve quarts w i l l
be added in succession with the shim rods completely withdrawn and the fission chambers f u l l y
’This
appendix was originally issued a s ORNL
CF-54-7-144, A R E Operating Procedures, Part 11,
Nuclear Operation, J. L. Meem (July 27, 1954).
’As previously noted, the original enrichment system
was not employed, although the principles and techniques of enrichment outlined here were followed.
144
i
t
CONCENTRATE^
The fuel concentrate w i l l be added in batches
t o an “eversafe”
container which w i l l contain
115 k g of UZ3’. The density of the fuel concentrate may be represented by the equation:
p (g/cm 3 ) = 5.51
L
i
in operating conditions (Le., 34-gpm fuel flow vs 46-gpm actual flow) and procedures (i.e., use
of the original enrichment system, which w a s replaced) will be noted
It i s anticipated that the reactor and i t s associated circuits w i l l be operated for about 250 hr
with fused salts in the system prior t o the introduction of fuel. A l l process instrumentation and
components (non-nuclear) w i l l be checked out
during t h i s period. T h e mechanical operation of
the nuclear equipment w i l l be checked out a t
temperature.
The safety rods w i l l be raised and
dropped approximately 50 times during t h i s period.
Several tests preliminary to the fuel loading
w i l l have been run. T h e rated flow o f the fuel
carrier w i l l be 34 gpm a t a mean temperature o f
E
c
L
t
t
L
i
b
t
i+
D
b
p.
c
inserted. After each quart o f concentrate i s added,
counts w i l l be taken on both fission chambers and
T h e reciprocal counting rate
the BF, counter.
w i l l be plotted vs. the fuel concentration to observe the approach t o criticality.
After 50% of the fuel concentrate (12 quarts) has
been added to the system, a sample of the mixed
fuel and carrier w i l l be withdrawn for chemical
analysis.
i
a
SUBCRITICAL EXPERIMENTS
.
I
During the addition of the f i r s t 12 quarts o f fuel
concentrate, the safety rods have been completely
withdrawn from the reactor. From here on, the shim
rods w i l l be inserted approximately 25% while a
quart of fuel concentrate i s being added. The shim
rods w i l l then be withdrawn and a count taken on
the fission chambers as before.
The shim rods
f
w i l l be inserted and withdrawn in t h i s fashion for
each succeeding fuel addition. The reciprocal of
the counting rate vs. fuel concentration w i l l s t i l l
be plotted t o indicate the increase i n c r i t i c a l i t y .
When approximately 90% o f the c r i t i c a l mass has
been added, the fuel addition w i l l be stopped and
For
several subcritical experiments performed.
these experiments it i s desired that the k o f the
reactor be about 0.97 t o 0.98. In Fig. D.l i s shown
the relationship between Ak/k and A M / M as a
function o f the c r i t i c a l mass M. Using t h i s curve,
an estimate of the point a t which fuel addition
must be stopped can be made.
For the f i r s t experiment, the fuel temperature
w i l l be decreased at a rate o f 10°F/min by starting
the helium flow.
Assuming a temperature coef(Ak/k)/'F, the resultant Ak/k
ficient of -5 x
should be about 0.25% after 5 min. If AM/M i s
DWG.22502
j"
f
4
1
.
\
ad
1
'
\
\
d
1
!
-2
20
22
24
26
28
30
32
CRITICAL MASS (Ib)
34
36
38
40
Fig. D.1. Reactivity-Mass Ratio as a Function of C r i t i c a l Mass.
145
approximately 4 A k / k (Fig. D.l), the count rate
should increase by an amount corresponding t o the
addition o f 1.0% more fuel. If t k e change in count
rate i s too small to be definite, the experiment
w i l l be repeated a t a rate of decrease o f 25'F/min
u n t i l a definite increase in count rate i s observed,
or until the temperature has been decreased 1OOOF.
if, after t h i s decrease i n temperature, no increase
in count rate i s observed, i t w i l l be assumed that
the temperature coefficient i s negligibly small and
the experiment w i l l proceed. As has always been
the philosophy on the ARE, i f the temperature
coefficient i s observed t o be positive, the experiment w i l l be concluded and the fuel c i r c u i t drained.
For the second experiment, the reactor temperature w i l l be returned t o i t s original value of
1300°F, and with the reactor containing 90% o f
the c r i t i c a l mass, the fuel flow rate w i l l be gradua l l y decreased. I f the available delayed neutron
fraction i s 0.47% Ak at f u l l flow and 0.75% Ak
when the flow i s stopped (Fig. D.2), a Ak o f
0.28% should appear. The count rate should in-
DWG. 2235t
0.008
0.007
0.006
0.005
>-
k
L
'0
0.004
a
W
cc
0.003
0.002
0.001
0
0
10
Fig.
146
D.2.
20
30
40
50
FUEL FLOW RATE ( g p m )
60
Reactivity from Delayed Neutrons as a Function of F u e l Flow.
70
80
-
crease by an amount equivalent t o the AM corresponding t o additional Ak i n the delayed neutrons.
A rough experimental check on the calculated
value of the delayed neutron fraction w i l l thus
be available.
During the period of subcritical operation, the
count rate w i l l be carefully observed for sudden
changes that could be caused by fuel segregation.
No such difficulty i s anticipated, but i f such an
effect i s observed, the reactor w i l l be shut down
u n t i l the reason for such behavior i s ascertained.
INITIAL CRITICALITY
4
k
d
*
Upon completion of the subcritical experiments,
the fuel flow and reactor temperature w i l l be
returned t o the i n i t i a l conditions of 34 gpm3 and
1300°F.
Counts w i l l be taken on the fission
chambers with the shim rods 25% inserted and
completely withdrawn.
The shim rods w i l l then
be 50% inserted and a quart of fuel concentrate
added.
The rods w i l l be withdrawn t o 25% and
a count taken and then completely withdrawn and
a count taken. From here on, two curves o f reciprocal counting rate vs. fuel concentration w i l I
be plotted, one at 25%rod insertion and the original
curve with the rods completely withdrawn. If the
reactor contained 90% of the c r i t i c a l mass during
the subcritical experiments, about two 1-quart
additions of fuel concentrate w i l l bring the reactor
c r i t i c a l with the rods completely withdrawn. The
l a s t fuel additions w i l l be made very slowly. When
c r i t i c a l i t y has been definitely reached, the reactor
w i l l be shut down and a fuel sample taken for
chemical analysis.
After the sample has been taken, the reactor w i l l
again be brought barely c r i t i c a l (a few hundredths
o f a watt) and held at constant power by watching
t h e count rate meters. The gamma-ray dosage w i l l
be measured with a “Cutie-pie” at a l l pertinent
points throughout t h e p i t s and recorded for future
reference on radiation damage and shielding.
ROD C A L I B R A T I O N vs F U E L ADDITION
The reactor w i l l be brought c r i t i c a l a t a very
low power and the shim rods adjusted so that the
regulating rod i s 5 in. above center.
The regulating rod i s estimated t o be worth about 0.04%
Ak/in.
One quart of fuel i s approximately 4%
AM or 1% Ak (Fig. D.l). Therefore, ?$5 of a quart
should be worth about
3Ac+ual, 46 g p m .
1 in. of
regulating rod.
Figures D.3 and D.4 show a calibration of a
regulating rod taken on a mockup of the ARE at
the Critical Experiment Facility, While it would
be fortuitous i f the regulating rod i n the actual
A R E gave the same calibration, it i s expected
that the general shape of the curves w i l l be the
same,
With the shim rods i n a fixed position, the
regulating rod w i l l be f u l l y inserted and approxiof a quart of fuel concentrate w i l l be
mately
added slowly t o the system.
The reactor w i l l
be brought c r i t i c a l on the regulating rod and i t s
new position noted.
T h i s procedure w i l l be
repeated u n t i l the regulating rod has been c a l i brated from 5 in, above t o 5 in. below i t s midposition.
The above experiment w i l l have been run w i t h
the shim rods almost completely withdrawn. The
quart
shims w i l l now be inserted about 25% and
o f fuel added (approximately 0.5% A k ) * The shim
rods w i l l be withdrawn u n t i l the reactor goes
critical.
T h i s procedure w i l l be continued t o
g i v e a rough calibration o f shim rod position v s
fuel addition over one-fourth of the rod.
When
the reactor i s c r i t i c a l with the shims approximately
15% inserted, the shims w i l l be adjusted so that
the regulating rod i s 5 in. above center, and a
second calibration of the regulating rod vs fuel
addition w i l l be run as above.
Calibration o f the shim rods v s fuel addition
w i l l then be continued u n t i l the reactor i s c r i t i c a l
w i t h the rods about 25% inserted, At t h i s time,
a third and final calibration of the regulating
rod vs fuel addition w i l l be run. The reactor w i l l
then be shut down and a fuel sample taken.
t5
’/2
LOW -POWER E X P E R I M E N TS
The nominal power of the reactor i s obtained
as follows: It i s estimated that in the reflector
a t the mid-plane of the reactor the fission producing flux i s 2 x lo6 nv/w, and the average f l u x
over the length of the fission chamber when f u l l y
nv/w.
Preliminary tests
inserted i s 1.5 x
on the fission chambers show a counting efficiency
The reo f approximately 0.14 counts/sec.nv,
lationship between counting rate and power i s
therefore approximately 2 x lo5 counts/sec = 1 w.
T h e nominal power o f the reactor w i l l be based
on t h i s relationship during the preceding experiments.
Until t h i s time the reactor has been operated
on the fission chambers alone a t a nominal power
lo6
147
DWG. 21525A
130
120
110
100
90
-
80
Y)
c
w
c
0 70
W
3
A
2
60
OI
0
50
40
30
20
(0
0
0
2
4
6
8
10
12
14
16
18
20
22
24
26
28
30
32
34
ROD POSITION ( i n . )
Fig.
D.3. Regulating Rod Calibration (Rod 6).
of about 0.01 w (about 2000 counts/sec).
The
reactor power w i l l now be increased to about
1 to 10 w, at which time the neutron ionization
chambers should begin to g i v e readings on the
log N recorder, the micromicroammeter, and the
period recorder. The reactor w i l l be leveled out
and the flux servo turned on. One o f the fission
chambers w i l l be completely withdrawn, which
should reduce i t s counting rate by several orders
A careful comparison o f the new
of magnitude.
count rate vs the o l d count rate w i l l be made.
The reactor w i l l be held at constant power for
exactly 1 hr and then scrammed. A fuel sample
w i l l be drawn off and sent t o the B u l k Shielding
F a c i l i t y for determining i t s activity. T h i s a c t i v i t y
w i l l be compared w i t h that of a previous sample
(Fig. D.3, which was exposed i n a known neutron
148
f l u x i n the B u l k Shielding Reactor.
From t h e
comparison, the f i s s i o n rate or absolute power
of the ARE w i l l be determined for t h i s run (cf.,
A l l neutron level instruments w i l l be
app. H).
calibrated accordingly.
Rod Calibration vs Period. A family of curves
o f reactivity v s reactor period (inhour curves)
w i t h the rate o f flow of the fuel as a parameter
i s shown i n Fig. D.6.
With the fuel f l o w rate a t 34 gpm, the servo w i l l
b e turned o f f and t h e reactor placed on i n f i n i t e
period manually w i t h the regulating rod a t midposition.
From the rod calibration v s fuel p l o t
and the theoretical inhour curve, t h e regulating
r o d withdrawal for a 30-sec period w i l l be e s t i The r o d
mated (present estimate i s about 1 in.).
w i l l be withdrawn accordingly and the neutron
DWG. 24526A
f
?
i
,
e
I
4
: .
1
1
I1
0
-1
-7
0
2
4
6
10
8
12
14
16
48
ROD POSITION
Fig.
4
J
1
D.4.
Regulating
l e v e l allowed t o r i s e about two orders of magnitude, whereupon the regulating rod w i l l be f u l l y
inserted and the induced gamma rays allowed t o
The reactor w i l l then
decay for about 20 min.
be brought back t o i t s original power, as determined by the fission chambers. If the position
o f the regulating rod does not return t o i t s original
value because of photoneutrons, the reactor power
w i l l again be reduced by insertion of the regulating
rod u n t i l the photoneutron effect becomes negligible. The above procedures w i l l be repeated for
20-sec and 10-sec periods.
The 10-sec period
should correspond roughly t o a 2-in. withdrawal
o f the regulating rod. Runs w i l l then be made at
correspondingly longer periods u n t i l the period
for a $-in. withdrawal of the rod i s obtained
At t h i s time a repeat
(approximately 100 sec).
run on the 30-sec period w i l l be made t o ascertain
whether photoneutron buildup i s causing appreciable error i n the measurements.
The shims w i l l then be adjusted so that the
20
22
24
26
28
30
32
34
i
(in.)
Rod Sensitivity (Rod
B).
regulating rod i s 1 in. below center a t i n f i n i t e
period.
The rod w i l l be withdrawn 1 in. and the
period recorded. T h i s procedure w i l l be repeated
a t successive starting positions of the regulating
rod 1 in. apart from 5 in. below center t o 5 in.
A check on the i n i t i a l 30-sec run w i t h
above.
the rod withdrawn from the mid-position w i l l be
made periodically t o ensure that photoneutron
buildup i s not interfering w i t h the measurements.
From the standpoints of safety and experimental
convenience, the regulating rod should be worth
between 0.3% and 0.5% Ak for i t s f u l l 12-in.
If these experiments show that the value
travel.
o f the rod i s not i n t h i s range, a new rod w i l l be
installed at t h i s time and the period calibration
If convenient, a fuel sample w i l l be
repeated.
taken a t t h i s time.
Measurement of the Delayed Neutron Fraction.
The delayed neutron fraction for a stationary fuel
reactor i s about 0.73%. For the ARE at a design
f l o w of 34 gpm, the fraction i s calculated t o be
149
I
i
w
J
5
10 6
L
IO
TIME AFTER SHUTDOWN (hours1
Fig.
0.4795,
D.5, Decay Curves for Fluoride Activated
and for smaller flow rates, the corresponding reactivity may be found from Fig. D.2,
Starting with the reactor stabilized at 34 gpm
and the f l u x servo on, the flow rate w i l l gradually
b e reduced. The calibrated regulating rod should
be inserted so as t o indicate the same reactivity
as shown in Fig. D.2. T h i s w i l l give experimental
verification of the previously calculated inhour
curves, Fig. D.6.
Preliminary Measurements of Temperature Coefficient. With the reactor held at 10 w by the f l u x
servo, the helium flow w i l l be turned on so as t o
drop the fuel temperature at a rate of 10°F/min.
As the mean reactor temperature drops, the servo
w i l l insert the regulating rod so as t o maintain
constant flux, and from the rod calibration, the
Before the
corresponding Ak can be obtained.
r o d has reached the l i m i t of i t s travel, the helium
150
i n the
Bulk
Shielding Reactor,
and the fuel allowed t o return
flow w i l l be shut
t o i t s original temperature.
From the recordings
o f rod position and mean temperature, a plot o f
Ak v s mean temperature can be obtained. T h e
i n i t i a l slope o f t h i s curve w i l l correspond t o the
fuel temperature coefficient. Because of the weak
signal received by the f l u x servo, t h i s measurement w i l l be only approximate and i s t o be
repeated a t higher power.
Depending upon how the reactor responds, the
procedure can be repeated by dropping the fuel
iemperature at rates of up t o 25'F/min.
Care w i l l
b e taken not t o drop the fuel temperature so low
as t o set o f f the low-temperature scram.
L
t
e
E
%
t
P
APPROACH TOPOWER
A t the conclusion of the above experiments and
before the fuel storage tank i s removed from the
0.0035
0.0030
0.0025
1
t
i
>
0.0020
t
?
L
V
W
r
0.0015
0.0040
0.0005
3
0
0
IO
Fig.
4
1
4
20
30
40
50
60
70
REACTOR PERIOD ( s e c )
80
90
100
110
1 20
D.6. Reactivity as a Function of Reactor Period for Several Fuel F l o w Rates.
pits, sufficient fuel concentrate w i l l be added t o
t h e system so that 4% excess reactivity i s available.
T h i s excess reactivity w i l l be absorbed
w i t h the shim rods. The shim rods w i l l have a
rough calibration by t h i s time, and it i s expected
that they w i l l be worth about 12% i n excess
reactivity.
Therefore, they w i l l need t o be inserted about one-third of their length.
After
addition of the final amount of fuel concentrate
t h e liquid level i n the pump w i l l be checked t o
ensure that a t least 0.3 ft3 of volume remains for
expansion of the fuel.
T h i s expansion volume
w i l l allow the fuel t o expand isothermally from
1300 t o 16OO0F, which i s well above the hightemperature scram level. The final sample of fuel
w i l l be taken for chemical analysis.
T h e reactor w i l l now be shut down so that the
concrete block shields can be put over the p i t s
and the p i t s flooded with helium. The fuel storage
tank w i l l f i r s t be removed and a f i n a l check on
a l l process equipment made.
The BF, counter
w i l l be removed and the temperature servo chamber
installed. The regulating rod w i l l be put on fast
drive. A check l i s t i s being prepared of a l l items
t o be reviewed before the p i t s are sealed.
If the temperature coefficient i s of sufficient
magnitude t o control the reactor, the flux servo
w i l l be l e f t in. However, if the magnitude of the
temperature coefficient i s marginal, the f l u x servo
w i l l be removed a t t h i s time and the temperature
servo connected.
After the p i t s have been sealed, the reactor w i l l
be brought t o a power of 1 kw and allowed t o
stabilize.
U p u n t i l t h i s time a l l neutron ion
chambers have been f u l l y inserted.
The ion
chambers w i l l be withdrawn slowly, one by one,
w h i l e the power level i s being constantly monitored with the fission chambers. The ion chambers
w i l l be set for a maximum reading of around 5 Mw.
151
T h e safety chambers w i l l be set t o scram a t the
same level.
T h e reactor i s now ready for f u l l power operation. The rod w i l l be withdrawn and the reactor
put on about a 50-sec period. Somewhere i n the
region from 10 t o 100 kw a noticeable increase
in reactor period and an increase in reactor temperatures should be observed. The helium flow w i l l
be started slowly and gradually increased urltil
a nominal power of 100 t o 200 kw i s reached
(AT from 25 t o 5OOF).
At t h i s time, the reactor w i l l be allowed t o
hr and a l l readings recorded.
stabilize for about
If the temperature coefficient i s of insufficient
magnitude to stabilize the reactor, as determined
by previous measurements, the control o f the
reactor w i l l be turned over t o the temperature
servo.
The power as determined from the heat extraction
by the helium w i l l now be measured.
The heat
capacity o f the fuel i s 0.23 Btu/lb.OF and the
density of the fuel i s represented by
'/2
p (g/cm3)
=
4.04 - 0.0011 T ("C)
.
A t a mean temperature of 13OO0F, the average
The power can be exdensity i s 3.27 g/cm3.
pressed as
p (kw) = 0.11 x AT (OF) x flow (gpm)
.
Therefore a t 34 gpm the power extracted by the
fuel i s 3.74 k w P F .
Having obtained a heat balance, the extraction
o f heat from the fuel in the heat exchanger w i l l
b e increased u n t i l a power of about 500 kw
(AT = 134OF) is reached. Again the reactor w i l l
be allowed t o stabilize and the extracted power
measured. A t h i r d and f i n a l heat balance w i l l be
made at a power of about 1 Mw (AT = 268°F).
T h i s i s close t o the maximum power obtainable
without changing the i n i t i a l conditions of 1300°F
mean reactor temperature and 34-gpm fuel flow.
During the preceding discussion, no mention has
been made of heat extraction other than in the
fuel circuit. An appreciable quantity of heat w i l l
be removed by the sodium in the reflector coolant
circuit.
Before the reactor goes t o high power,
t h e sodium inlet and outlet temperatures w i l l be
near the isothermal temperature of 1300°F. Since
there i s some gamma heating i n the reflector
region, it w i l l be necessary t o lower the i n l e t
temperature of the sodium by extracting heat w i t h
the reflector coolant heat exchangers. The sodium
f l o w rate w i l l be held at 224 gpm and the sodium
152
mean temperature a t 1300°F.
Since the heat
capacity of the sodium i s 0.30 Btu/lbSoF and i t s
specific gravity i s 0.78, the power extracted by
the sodium w i l l be 7.7 k w P F . No mare than 10%
of the power generated i s expected t o go i n t o the
sod iurn.
b
I
b
E X P E R I M E N T S A T POWER
Measurement of the Temperature Caeff icients.
After the reactor power has been cal ibrdted, the
power w i l l be reduced t o about 500 kw and allowed
t o stabilize.
The flux servo w i l l be turned on,
and by means of the helium demand the mean
temperature w i l l be dropped a t an i n i t i a l rate of
(If the temperature servo i s connected,
10°F/min.
it w i l l have t o be replaced for t h i s experiment.)
A s discussed previously the i n i t i a l slope of the
p l o t of Ak v s mean temperature wi I I represent the
f u e l temperature coefficient.
Contrary t o the
procedure a t low power, however, the helium flow
w i l l not be decreased after the temperature has
dropped. The reactor w i l l be allowed t o stabilize,
and, after about 30 min, the regulating rod w i l l
have leveled out a t a new position, and the reactor
w i l l have assumed a new mean temperature. These
readings w i l l represent the over-all reactor temperature coefficient (fuel plus moderator). If, during
t h e above experiment, the servo inserts the rod
t o i t s limit, it w i l l be l e f t a t that position since
t h e temperature should stabilize the reactor.
If the fuel temperature coefficient i s quite small,
t h e experiment can be repeated with an i n i t i a l
rate of temperature decrease of 25"F/min.
Care
w i l l be taken that the reactor does not go on too
fast a period and that the low-temperature scram
l i m i t i s not exceeded.
Maximum Power Extraction. Except for specific a l l y stated instances, the reactor has been
operated continuously up t o t h i s time at a mean
temperature of 1300°F and a fuel flow rate o f
34 gpm, Under these conditions, the maximum
obtainable power i s 1.1 Mw. At t h i s power, the
reactor outlet temperature and the heat exchanger
The heat
i n l e t temperature are both 145OOF.
exchanger outlet and the reactor i n l e t are both
a t 115OOF. It i s t o be noted that when the heat
exchanger outlet temperature drops below 115O"F,
a n alarm i s sounded.
The mean reactor temperature w i l l be elevated
from 1300 t o 1325°F by a slight withdrawal of t h e
shim rods.
The mean temperature of both fuel
and sodium w i l l be held at t h i s temperature. T h e
helium blower speed w i l l now be increased u n t i l
i
b
L
h
L
s
a-
i
I
= i
i
!
AT of nearly 350°F appears across the reactor.
With t h i s AT, the reactor outlet and heat exchanger
i n l e t are at the upper l i m i t of 1500"F, and the
a
d
af
i
i
i
?
21
-1
4
4
B
:
1
4
heat exchanger outlet and reactor i n l e t are at the
The power from the fuel
lower l i m i t of 1150°F.
w i l l be 1.3 Mw, A t t h i s time the pump speed can
be increased from 34 gpm t o about 40 gpm, caution
being taken that the reactor inlet pressure does
not exceed 50 psig. The maximum AT across the
reactor w i l l s t i l l be 35OoF, and the power extracted in the fuel circuit w i l l be 1.5 Mw. T h i s
i s the maximum power a t which the reactor can
be operated.
With the reactor on manual
Power Transients.
control and the regulating rod a t mid-position, the
helium blower speed w i l l be regulated u n t i l the
power in the fuel system i s 1 Mw, and the reactor
w i l l be allowed t o stabilize.
The helium flow
w i l l be suddenly increased t o i t s maximum and
the transient response of a l l temperature and
nuclear recordings noted. The experiment can be
repeated from successively decreasing i n i t i a l
powers of 500, 200, 100, 50, 20, and 10 kw. At
some level below 100 kw, the power cannot be
reduced further because the helium blowers w i l l
b e completely shut off. When t h i s lower l i m i t of
i n i t i a l power has been reached, the series o f
If a t any time
experiments w i l l be concluded.
the reactor period gets too fast or the upper or
lower temperature l i m i t s are approached, the experiments w i l l be concluded.
Sudden Changes i n Reactivity. The reactor w i l l
be brought t o a power o f 1 Mw and allowed t o
stabilize,
From t h e previous calibration of the
regulating rod, the rod w i l l be placed at a position
such that a complete withdrawal w i l l give 10 cents
o f excess reactivity.
T h e rod w i l l suddenly be
withdrawn and the transient response of the i n l e t
pressure and a l l nuclear and temperature recorders
noted.
The experiment can be repeated w i t h
sudden changes o f 25, 50, 75, and 100 cents of
reactivity. If the fuel temperature coefficient has
a value o f approximately -5 x
as expected,
a sudden change in reactivity of 100 cents should
b e safe. However, if the experiments w i t h smaller
reactivity changes indicate that such a large step
w i l l be unsafe, t h i s series o f experiments w i l l be
concluded.
Otherwise, the experiment w i l l be
repeated from i n i t i a l powers of 100 kw, 10 kw, and
successively lower power levels.
Reactor Startup Using
- the Temperature CoefThe reactor w i l l be brought t o 1 Mw o f
ficient,
power and t h e shims adjusted so that the regulating rod i s completely withdrawn. The helium
f l o w w i l l be cut o f f and the reactor power allowed
t o drop t o i t s normal power with no heat extraction
The
(estimated t o be between 10 and 100 kw).
regulating rod w i l l now be inserted by 0.1% of
reactivity and the reactor allowed t o go subcritical
for about 10 min.
After t h i s time, helium flow
w i l l gradually be increased.
If the temperature
a drop i n the mean
coefficient i s -5 x
temperature of the reactor of 2OoF w i l l bring the
As soon as a positive
reactor c r i t i c a l again.
period i s noted, the helium cooling flow w i l l be
held fixed and the reactor allowed t o come up t o
power and level out of i t s own accord.
loe5,
The experiment can be repeated by driving the
reactor subcritical by 0.2, 0.3, and 0.4% with the
regulating rod.
Care w i l l be exercised not t o
approach the upper or lower temperature limits.
E f f e c t of Xenon Buildup.
The change i n reactivity calculated from xenon buildup in the ARE
i s shown in Fig. D.7.
The reactivity as plotted\
i s nominal because of uncertainties in the xenon
cross sections and i s probably a maximum. The
shape o f the curves represents the change in
reactivity i f no xenon i s l o s t by off-gassing. The
reactor w i l l be operated for 5 hr a t f u l l power,
and then reduced t o 10% of f u l l power. If no xenon
i s lost, the reactivity should change as shown in
the lower curve of Fig. D.7.
T h e reactor w i l l
then be operated for 25 hr a t f u l l power, and
subsequently reduced t o 10% power. Again i f no
xenon i s lost, the reactivity change should be
If some
a s shown i n the upper curve of Fig. D.7.
xenon does off-gas, the shape o f the curves w i l l
b e changed, and by proper analysis, an estimate
o f the amount of off-gassing can be obtained.
A t f u l l power, the reactivity change w i l l be
calculated from the change i n reactor mean temperAt 10%
ature and the temperature coefficient.
power, the reactor w i l l be put on f l u x servo and
t h e movement o f the regulating rod w i l l measure
t h e reactivity change.
Since the moderator w i l l
contain considerable heat when t h e power i s
reduced, the mean reactor temperature w i l l slowly
hr after t h e
decrease during about the f i r s t
power i s reduced t o 10%. T h i s w i l l cause an
insertion o f the regulating rod by the f l u x servo.
Since the xenon buildup w i l l cause a withdrawal
o f the regulating rod, a correction must be applied
for the effect o f the drop in moderator temperature.
\
153
154
N
'0
M
N
c)
0
I
M
c
-
I -
N
0
P
-
'
I
i
i
k
1
6
e
L
t
1
Ek
b
e
b
k.
L
L
k
i
= k
P
To
obtain t h i s correction experimentally, a
control experiment w i l l be performed. The reactor
t o 1 hr and
w i l l be operated at f u l l power for
t h e moderator allowed to come up t o temperature.
During t h i s short time, no appreciable xenon w i l l
'/2
be formed. The reactor w i l l then be reduced to
10% power and the flux servo turned on. The
reactivity change from the drop in moderator
temperature w i l l be observed and used as a correction for the xenon buildup experiments.
t
i
d
I
i
I
i
155
I.
h
Appendix
E
MATHEMATICAL ANALYSIS OF APPROACH TO CRITICALITY
W. E.
When the ARE was brought to c r i t i c a l by successive fuel additions, i t was observed that the
( V m u l t i p l i c o t i o n constant)] vs
usual plot of [l
uran ium concentration increased, at first, very
rapidly, but, when the curve got close to 1, the
r i s e was very slow. Qualitatively, such behavior
i s observed i n many reactors, but the ARE exhibited the effect to an unusual degree. In order t o
explain this, the ORACLE three-group, three-region
code was modified so that f l u x shapes at successive fuel additions could be calculated.
Goup
constants were obtained by f l u x weighting with
fluxes from an Eyewash calculation on the ARE.
Figure E.l shows the space distribution of the
thermal flux for no fuel and for runs 2 through 6.
-
Kinney
f
The effect of the reflector as fission neutrons become more numerous can be seen,
Figure E.2
compares experimental and calculated startup
curves for the fission chambers which were located
i n the reflector as indicated i n Fig. E.l.
I n the
calcu lotion,
I
t
where CR i s the counting rate of the f i s s i o n
chamber, o f . i s the f i s s i o n cross section for group
i , and rPi i s f t h e group i flux. For the ARE startup
CRi
5
c
4
c
i.
t
b
L
3
X
3
LL
1
_I
2
LL
W
+
I
2
c
i
6"
I
t
i
L
0
0
IO
20
30
40
RADIUS (cm)
Fig.
156
E.1.
T h e r m 1 F l u x vs Radius.
50
60
70
1
ORNL-LR-DWG
4451
1 .c
0.E
c
-E
-2.
0.6
I
0.4
02
0
0
2
6
4
IO
8
!2
t4
16
U235 CONCENTRATION ( l b / f t 3 )
Fig.
F
E.2. 1
- (l/m)
where m i s the multiplication constant, C R . i s the
I
counting rate on run j , and C R , i s the counting
rate with no fuel.
It seems, then, that the unexpected, rapid i n i t i a l rise i n the counting rate o f
( l / m ) ] curve i s
the fission chamber and the [l
due not only to the general rise i n flux level but
-
vs
U235 Concentration.
also to the formation o f the thermal-flux maximum
near the fission chambers.
Once the shape of
spatial distribution of the thermal-neutron flux i s
set up, the fission chambers register only the
general increase in flux level and the count rate
increases slowly.
157
F
Appendix
COLD, CLEAN CRITICAL MASS
The value of the cold, clean c r i t i c a l mass i s o f
interest i n connection with any reactor, since it i s
the calculation of this elusive number that takes
so much time during the design of the reactor and
system.
The cold, clean c r i t i c a l mass may be
found by extrapolating the curve of uranium (Ib of
U235) i n the core vs Ak/k (%) i n the rods to the
On the
p o i n t where there i s no rod poisoning.
other hand, if the mass equivalent of the rod
poisoning i s determined and deducted from that
then i n the core, another curve may be obtained,
T h i s curve gives corrected mass vs Ak/k (%) i n
the rod and may also be extrapolated to 0% Ak/k
i n the rods.
Both these curves, as shown i n
Fig. F.l, extrapolate to the same mass value a t
0% AkJk in the rods, which indicates that the rod
calibration was quite accurate and the data were
reliable.
The value of the cold, cleon c r i t i c a l
mass extrapolated from Fig. F.l was 32.75 Ib o f
u235.
The data from which the curves i n Fig. F.l were
plotted are tabulated i n Table F.l. The data used
to determine the cold, clean c r i t i c a l mass were
those obtained during the rod calibration from fuel
addition (Exp. L-2). T h i s experiment and Exp. L-5,
rod calibration from reactor period, then provided
the information on the value of the regulating rod.
The values of the shim rods were taken from
Appendix J, “Calibration of the Shim Rods.” The
Ak/k values i n both regulating and shim rods were
then converted to their mass equivalents by using
the (Ak/k)/(AM/M) r a t i o o f 0.236, as determined
experimentally,
Three points listed in the table
and shown i n the figure were derived by using the
f i r s t regulating rod, which was subsequently reThe calibraplaced because it was too “light.”
158
tion for that rod wos not accurate, and, hence, the
three points were disregarded i n extrapolating the
curves.
OANL-LR-DWG
IC
-
I
t
c
64!9
36
l
i
!
l
:URVE
A U235 IN CORE vs Ak/k IN RODS
35
B
Ak/k IN RODS vs U235 IN CORE C(
3ECTED TO NO
t
34
b
-e
33
1
0
ro
N
3
32
31
I
30
29
0
02
0 4
06
08
( 0
12
Ak/k IN ALL RODS (’70)
Fig.
Mass.
F.l.
Extrapolation t o Cold, Clean C r i t i c a l
c
a
I
t:
TABLE F.l.
i
CRITICAL URANIUM CONCENTRATIONS FOR VARIOUS ROD INSERTIONS
(Exp.
?)
Run No.
Regulating
Rod Insertion
(in.)
Shim Rod Insertion
(in.)
N ~ 1. No. 2
No. 3
M/k
of
A k / k of Shim Rods
Regulating
Rod (%)
L-2)
No. 1
( %)
No. 2
No. 3
U235
Concentration
i n Core
Total A k / k
in Rods
(lb/ft3)
(Ib)
(%)
~ 2 3 5
Total
AM/M i n
Rods
U235 i n
Clean Core
(Ib)
0
0.92
3.3
1.7
0
0.016
0.19
0.08
0
23.94
32.80
0.286
1.212
31.59
1
3.35
3.3
1.7
0
0.058
0.19
0.08
0
23.98
32.85
0.328
1.390
31.46
2
0.93
3.3
2.5
0
0.016
0.19
0.13
0
24.12
33.04
0.326
1.381
31.66
3
4.25
0
0
3.5
0.140
0
0
0.20
24.17
33.11
0.340
1.441
31.67
4
6.35
3.5
0.210
0
0
0.20
24.22
33.18
0.410
1.737
31.44
7.5
0
0
0
5
0
3.5
0.248
0
0
0.20
24.25
33.22
0.448
1.898
31.32
1
6
8.5
0
0
3.5
0.281
0
0
0.20
24.27
33.25
0.481
2.038
31.21
a
Level
Trimmed
7.4
2.2
2.5
0
0.244
0.11
0.13
0
24.27
33.25
0.484
2.051
31.20
0
0
0
0.264
0.11
0.13
0
24.29
33.28
0.504
2.135
31.14
0.281
0.11
0.13
0
24.3 1
33.30
0.521
2.207
31.09
0.360
0.14
0.15
0
24.41
33.44
0.650
2.754
30.69
0.041
0.33
0.33
0
24.43
33.47
0.701
2.970
30.50
0.382
0.23
0.16
0
24.44
33.48
0.772
3.271
30.21
I
1
d
@
7
8.0
2.2
2.5
8
8.52
2.2
2.5
9
10.92
2.7
2.8
10
1.25
5
5
11
11.56
4
3
0
0
1
?
1
4
1
4
I
L
1
a
Appendix
G
FLUX AND POWER DISTRIBUTIONS
A. D. Callihan
There were, of course, no measurements of the
flux or power distribution i n the reactor during the
actual experiment.
These distributions were
measured, however, on a zero power, c r i t i c a l mockup of the experimental reactor over a year earlier.
A detailed description of the c r i t i c a l mockup, includ ing the reactor parameters obtained therefrom,
The neutron flux
may be found i n ORNL-1634.'
distributions obtained from indium and cadmiumcovered indium f o i l measurements, the fission
neutron flux distributions obtained by the catcher'A. D. Callihan ond D. Scott, Preliminary Critical
Assembly for the Aircraft Reactor Experiment, ORNL1634 (Nov. 28, 1953).
D. Scott
f o i l method, and the power distributions obtained
from aluminum catcher foils, which are of particular
interest, are presented here.
-
k
t
b
NEUTRON F L U X DISTRIBUTIONS
The neutron flux distributions were measured
with indium and cadmium-covered indium f o i l s i n a
number of runs i n which a remotely placed uranium
disk and an aluminum catcher f o i l were used t o
normalize the power from run to run, The results
of the bare indium and cadmium-covered indium
traverses made a t a point 12.06 in, from the center
of the reactor are shown in Fig, G.1. The zero of
the abscissa i s the bottom o f the beryllium oxide
DWG. 2 1 5 2 9
INDIUM TRAVERSE, LONGITUDINAL
AT 12 06 in RADIUS
I
k
c
I
c
.
t
t
x
k
0 8
&
e
-
ti
0.6
30
h0
a
z
0
F
a
0
i
(L
0.4
20
I
i
I
b
U.
2
L
0
a
P
L
0
0 2
i
,
0
i
0
0
42
6
18
24
30
36
DISTANCE FROM BOTTOM OF REACTOR (in.!
L
c
i
Fig.
160
G.1.
Longitudinal Neutron F l u x Distribution.
k
*
.
column where i t rests on the 1-in,-thick lnconel
support plate.
The scattering by the lnconel
probably accounted for the reduction of the cadmium fraction a t t h i s point.
The radial f l u x traverse a t the mid-plane of the
reactor with the regulating rod inserted i s given i n
Fig, G.2.
The dashed lines on the figure are
a c t i v i t i e s extrapolated from the data obtained i n
the fine-structure measurements made near the
11-in. position, The wide gap i n this traverse was
unexplored because of the importance which was
attached t o the study of a u n i t core c e l l i n the
time available. T h i s emphasis has been at least
partially contradicted by the f i s s i o n f l u x traverse
described i n the following section.
The strong
effect of the center regulating rod assembly i s
indicated i n these curves.
The comparatively
small flux and cadmium fraction depressions a t
the center of the regulating rod accentuated the
relative inefficiency of this type of rod and guide
tube arrangement for reactor control.
I n another
measurement of the radial flux, this time with the
regulating rod withdrawn, the traverses were very
similar to that shown i n Fig. G.2, but the neutron
f l u x was s l i g h t l y greater (-8%).
FISSION-NEUTRON F L U X D I S T R I B U T I O N
A measure of the distribution of neutrons that
caused fissions was obtained by using the catcherf o i l method i n which the aluminum foil, together
with some uranium, was both bare and cadmium
covered. A radial traverse taken a t the mid-plane
of the reactor i s shown i n Fig. G.3. The flux of
low-energy neutrons produced f i s s i o n peaks near
the reflector and was depressed by the lnconel
tube (regulating rod sleeve) at the center. The
DWG 215308
07
06
05
z
0
04
5a
LL
I
3
03
0
5
4
0
02
04
0
0
2
4
6
8
10
I?
14
46
18
20
22
24
D I S T A N C E FROM A X I S OF R E A C T O R (in.)
Fig, G.2.
Rodial Neutron F l u x Distribution.
16 1
been desirable to repeat the experiment w i t h
smaller f o i l s to improve the resolution, particularly
since the results were very sensitive t o f o i l location because of fuel self-shielding. Each catcher
f o i l was nominally located centrally on the a x i s
o f a fuel slug. One traverse extended from below
the lnconel support plate t o above the top of the
beryllium oxide; the other two covered the upper
The data obtained i n the fuel
h a l f of the core,
tubes a t the indicated r a d i i are shown i n Fig. G.5.
The data of Fig. G.5 are replotted i n Fig. G.6 t o
show the radial power distribution at several elevations i n the reactor, a l l normalized at the center.
It i s to be noted thatthe37.86-in. elevation traverse
i s 2 i n a b o v e the top of the beryllium oxide column.
cadmium fraction followed roughly the same pattern.
Similar half-length longitudinal traverses are shown
i n Fig, G.4 for a position 10.09 in, from the center.
POWER D I S T R I B U T l O N
Three longitudinal power distributions i n this
reactor assembly were measured by using aluminum
catcher f o i l s placed against the end surfaces of
the short, cast, fuel slugs contained i n an lnconel
The s i z e of these f o i l s was such that the
tube.
counting rate from each was higher than necessary
for the desired statistics.
I n the arbitrary units
reported, an a c t i v i t y of 50 represents approximately
lo5 counts. Had time permitted, it would have
I
..,B
c
c
I
OWG 2l533A
IO
08
z
06 0
+
0
LL
lL
I
2
H
0 4
2
0
0 2
0
0
3
6
'
9
12
.
15
$8
21
24
27
DISTANCE FROM AXIS OF REACTOR ( i n )
Fig, G.3. Radial F i s s i o n Neutron Flux Distribution.
162
c
DWG 21534A
J
1.4
c
1.2
/
1.0
z
0
0.8
5
a:
LL
E
0.6
3
5
n
J
Q
0
0.4
i
0.2
c
0
0
4
8
12
DISTANCE
F i g . G.4.
20
24
28
FROM BOTTOM OF REACTOR ( i n . )
16
32
36
40
Longitudinal F i s s i o n Neutron F l u x Distribution.
J
163
DWG 245354
90
80
70
I
-
60
lJl
t
z
3
-
2a
cm
50
E
E
a
> 40
t2
Iu
-
4
30
20
io
0
-4
0
4
t2
9
16
20
24
28
32
36
40
44
DISTANCE FROM BOTTOM OF REACTOR (in.)
Fig. G.5.
Longitudinal Power Traverses.
DWG 2l536A
fi
10
09
e
08
g 07
5a
06
w 0 5
5 04
_I
!$
0.3
02
O.'0
"
ic
I
0
4
2
4
6
8
10
12
DISTANCE FROM AXIS OF REACTOR ( i n )
Fig. G.6. Radial Power Traverses.
164
i
Y
h
14
i6
ir
i
i
Appendix H
3
I
t
POWER DETERMINATION FROM FUEL ACTIVATION’
E. B. Johnson
Perhaps the most basic value obtained from the
operation o f the ARE was the power level at which
it operated.
T h i s quantity was determined both
from activation o f the fuel and from a heat balance.
The determination of reactor power from activation
of the fuel was based on the measurement of the
relative activity of samples of the fuel exposed i n
the ARE and that exposed i n a known flux i n the
Bulk Shielding Reactor (BSR).
The activity of the aggregate of fission products
i s not easy t o predict because o f the large number
of isotopes for which calculations would be required. For identical conditions o f exposure and
decay, that is, for identical flux, exposure time,
and waiting period after exposure, the specific
activity should be the same, however. Furthermore, a t low fluxes the activity should be proportional t o the flux.
The ARE was operated a t a nominal power level
o f approximately 1 w for 1 hr, was then shut down,
and a sample of the irradiated fuel was withdrawn
for counting and analysis. Since the activity i n
t h i s fuel sample was too low t o measure accurately,
the test was repeated at an estimated power of
10 w. The ARE, experimental program then proceeded as scheduled. Decay curves were obtained
from the fuel samples from the ARE and were then
compared with the decay curve of a similar fuel
sample irradiated in the BSR i n a known neutron
flux of about the same magnitude. From the relative activity of the samples, a determination o f
the power of the ARE was made.
1i
-
i
I
’
e
t
1
-1
..
d
a
the fission cross section, energy release per
fission, and the necessary conversion factors.
Obviously, i f the thermal-neutron flux and the
amount of fuel present are known, it i s quite simple
t o obtain the power. The problem was t o obtain
the power without measuring the flux i n the ARE.
The quantity P / G obtained from the measurement
in the BSR w i l l be called (P/G)BsR.. The decay
curve for the two fuel samples irradiated i n
the BSR i s shown i n Fig. D.5 of Appendix D. The
power production in a given sample i s proportional
t o both the counting rate and the amount of fissionable material i n the sample. Thus,
( P / G )A R E
SR
-
(CR/G),
E
(CR/G),,,
where
CR = counting rate a t time t after shutdown,
G = weight of uranium i n the sample.
c
e
T h i s equation can be rewritten as
C
R
~
R ‘6
~
x - .
SR
G~~
t
E
The total power of the ARE i s then the product o f
the power per gram ( i n the sample) and the total
amount of uranium i n the fuel c i r c u i t (Gtot)ARE, or
i
THEORY
It has been shown2 that the power ( i n watts)
produced i n an enriched-fuel thermal reactor i s
P =
4.264
x
lo-’’
x nvth x G
-“.
4.264 x 10’
’ ’,
X
(G+ot)A R E
P
cr
ARE
I
where
i s the number O f grams Of u235 in the
volume over which the average thermal-neutron flux
i s nuth, and the constant,
GBSR
X-
contains
’This
appendix was o r i g i n a l l y issued as ORNL
CF-54-7-11, Fuel Acttuation Method for Power Detennination o / the A R E , E. B. Johnson ( J u l y 31, 1954), and
was r e v i s e d for t h i s report on A p r i l 22, 1955.
2J. L. Meern, L. B. H o l l a n d , and G. M. McCarnmon,
Determination o f the Power of the Bulk Shielding Reactor, Part Ill. Measurement of the Energy Released
per Fission, ORNL-1537 (Feb. 15, 1954).
The unperturbed thermal-neutron flux i n which the
7
samples were exposed i n the BSR was 1.895 x 10
neutrons/cm2.sec.
~ the self-depression
~
~
3 ~
H
of the flux by the uranium sample was
fore ( P / G )
becomes
(P/G)BSR = 4.264 X 10”’
=
6.464 x
X
1.895
X
0.80. There10’
X
0.80
loa4 w/g .
3W. K. Ergen, p r i v a t e communication.
4
165
~
Each of the fuel samples irradiated in the
contained 0.131 g o f U235. Thus
BSR
a t 1- and 10-w nominal power. The instruments
which recorded the neutron level (log N, etc.) were
then calibrated in terms o f reactor power.
EXPERIMENTAL PROCEDURE
0.131
X-
= 8.468
1 0 - ~x
C
(‘tot)
r
u
AR E
ARE
RE ~
~
C R S R~
(‘tot)
AR E
X
‘ARE
This equation was then used t o determine the
power o f the A R E as indicated by fuel activation
during the previously mentioned operation o f 1 hr
The BSR i s immersed i n a pool of water which
serves as moderator, coolant, reflector, and shield.
Therefore, any material which i s t o be activated
i n the BSR must be placed i n a watertight container
before immersion in the pool. Capsules t o contain
the ARE fuel were made o f 2s aluminum with screwtype caps and were sealed with a rubber gasket.
Each capsule contained approximately 1 g o f material; one of the capsules i s shown in Fig. H.l.
Since it was desirable t o irradiate the capsules
i n the reactor core, the reactor was loaded with a
partial element, containing only h a l f the usual
number o f plates, i n the interior o f the lattice, as
\
/
ORNL-LR-DWG
1947
c
t
Fig.
166
H.1.
Apparatus for F u e l Exposure in the Bulk
Shielding Reactor.
,
shown i n Fig. H.2. A l u c i t e holder was designed
which would fit inside t h i s partial element and
support the capsule t o be irradiated a t approximately the vertical centerline of the element. Small
gold f o i l s for measuring the thermal-neutron flux
adjacent t o the capwere mounted in “drawers”
sule, as shown i n Fig. H.l. A monitoring f o i l was
placed on the bottom o f the lucite block.
O R N L - L R - D W G 4948
SAFETY RODS
L A I TRAP , , , , , ,
\
PARTIAL ELEMENTS
Fig,
J
I
;
-
H.2.
/’
Bulk
ELEMENT FOR
CAPSULE ACTIVATION
‘CONTROL
Shielding
ROD
Reactor Loading
(No. 27).
Two samples of each of two ARE-type fuels of
different composition were irradiated. One (No.
44) contained 53.5-40.0-6.5 mole % of NaF-ZrF4UF, and was the anticipated fuel of the ARE. The
other was No, 45, which was the carrier t o which
the uranium-containing fuel concentrate was added
during the c r i t i c a l and low-power experiments; it
contained 50-50 mole % o f NaF-ZrF,.
It was necessary i n the BSR t o expose the
samples in the aluminum capsules; a t the ARE the
fuel was, o f course, f i r s t activated and then poured
i n t o the capsules. Therefore it was necessary t o
determine the contribution of the aluminum capsule
t o the activity observed. Two empty capsules were
exposed ( i n separate runs) for 1 hr a t a nominal
power level of 1 w. A t the same time, gold f o i l s
were exposed i n the positions adjacent t o the
capsule t o determine the thermal flux. Similarly,
two carrier-filled capsules and two f u e l - f i l l e d capsules were irradiated, each separately but each
for 1 hr a t 1 w. The capsules each contained 1 g
o f material.
The decay curves taken on each capsule were
made with two counters because of the difference
i n disintegration rates between the empty and the
A scintillation counter
fluoride-filled capsules.
was used to obtain the curves for the empty capsules and for the carrier-filled capsules, and the
high-pressure ion chamber was used to measure
the activity i n the f u e l - f i l l e d capsules. The data
for the carrier-filled capsules were converted, on
the assumption that a l l the activity was attributable
t o sodium, to equivalent counting rates i n the highpressure ion chamber.
The assumption that the
activity was due t o sodium was apparently valid,
since the decay curve indicated a h a l f - l i f e of about
15 hr a t approximately 6 hr after shutdown.
It was found from the decay curves that the
aluminum activity was a factor of approximately
10 less than that o f the carrier-filled capsules 3 hr
after shutdown and a factor o f 450 less than that
o f the f u e l - f i l l e d capsules a t the corresponding
time, and was therefore negligible. The activity
in the carrier-filled capsule was about 4% of that
in the f u e l - f i l l e d capsule 6 hr after shutdown.
There was essentially no difference between the
two different f u e l - f i l l e d capsules and the two different carrier-filled capsules.
For t h i s reason,
only one curve i s shown for each i n Fig. D.5.
The nuclear power o f the Aircraft Reactor Experiment was then set on the basis of the activation
in fuel samples irradiated for 1 hr a t 10 w. T h i s
power calibration was the basis of a l l power levels
indicated i n the nuclear log up t o the l a s t day of
the experiment, by which time i t had become obvious from the various heat balances that the actual
reactor power was considerably higher (app. L).
A comparison o f the two methods o f power c a l i bration i s given i n Appendix N.
I
167
I
Appendix
I
1
INHOUR FORMULA FOR A CIRCULATING-FUEL REACTOR WITH SLUG FLOW'
W. K. Ergen
A s pointed out in a previous paper,2 the circulating-fuel reactor differs in i t s dynamic behavior
from a reactor with stationary fuel, because fuel
circulation sweeps some o f the delayed-neutron
precursors out of the reacting zone, and some
delayed neutrons are given o f f i n locations where
they do not contribute t o the chain reactions. One
o f the consequences of these circumstances i s the
fact that the inhour formula usually derived for
stationary-fuel reactor^,^ requires some modificat i o n before i t becomes applicable to the circulatingfuel reactor. The inhour formula gives the relation
between an excess multiplication factor, introduced
into the reactor, and the time constant T o f the
If the inhour
resulting rise i n reactor power.
formula i s known, then the easily measured time
constant can be used to determine the excess
multiplication factor, a procedure frequently used
in the quantitative evaluation of the various
arrangements causing excess reactivity. Furthermore, the proper design of control rods and their
drive mechanisms depends on the inhour formula.
Frequently, the experiments evaluating smal I
excess multiplication factors are carried out a t
low reactor power, and the reactor power w i l l then
not cause an increase in the reactor temperature.
T h i s case w i l l be considered here. In this case,
the time dependence of the reactor power P can be
described by the following e q ~ a t i o n : ~
(1) d P / d t
7 i s the average lifetime o f the prompt neutrons
and k e x the excess multiplication factor (or excess
reactivity).
The meaning o f p and D ( s ) for a
' T h i s appendix was issued earlier a s ORNL CF-5312-108,
The Inhour Formula for a Circulating-Fuel
Nuclear Reactor with Slug Flow, W. K . Ergen (Dec. 22,
1953).
*William Krasny Ergen,
702-71 1 (1954).
J. Appl. P h y s . ,
Vol.
25, NO. 6,
for instance, S. Glasstone and M. C. Edlund,
The Elements of Nuclear Reactor Theory, D. Van
Nostrand Co., Inc., 1952, p. 294 ff.
3See,
authors, for instance Glasstone and Edlund,
loc. cit.. write the equations corresponding to (1) in a
s l i g h t l y different form. The difference consists i n terms
of the order k e x ( 7 / T ) or kex,B, which are negligibly
small.
4Sorne
168
circulating-fuel reactor has been discussed in some
detail i n ref. 2. We approximate in the following
the actual arrangement by a reactor for which the
power distribution and the importance of a neutron
are constant and for which a l l fuel elements have
8 through the outside
the same transit time 8,
D ( s ) i s simply the
loop. In t h i s approximation,
probability that a fission neutron, caused by a
power burst a t time zero, i s a delayed neutron,
given o f f inside the reactor a t a time between s
and s + ds, D ( s ) i s normalized so that
-
*;i
e
-
p
Jo%(s)ds = 1
(2)
k
PW)
(3)
t
L
a
i
.
c
t.
k
t
I f the fuel i s stationary, p D ( s ) i s the familiar curve
obtained by the superposition of 5 exponentials:
5
I"
t
P
-Xis
Pihf
=
t
i= 1
i
The X i are the decay constants of the 5 groups of
are the probabilities
delayed neutrons, and the
that a given fission neutron i s a delayed neutron
of the i t h group.
For the circulating-fuel reactor we f i r s t consider
the fuel which was present i n the reactor a t time
At any time s, only a fraction o f t h i s fuel
zero.
T h i s fraction i s
w i l l be found in the reactor.
denoted by F ( s ) , and by multiplying the right side
o f (3) by F ( s ) , we obtain the function PO(.) for
the circulating-fuel reactor.
Since 8, i s the t o t a l time required by the fuel
t o pass through a complete cycle, consisting o f
the reactor and the outside loop, it i s clear that
pi
t
4
*
a t s = ne,
(n = 0, 1, 2 . . ,),
at s = ne,
+
F ( s ) i s equal to 1;
8
.
(n = 0, 1, 2 ..),
(We assume 8, 2 20 s o that the fuel under consideration has not started t o re-enter the reactor
when the l a s t o f i t s elements leaves the reacting
zone.)
Between s = ne, and s = ne, + 8 , F ( s )
decreases linearly, and hence has the value
At =
= 1, 2, 3
(ne, + e - s)/e.
t
F ( s ) i s equal to zero.
ne, - e(n
...I,
-
i
t
c-
i
I
i
a'
F ( s ) i s zero, but since the fuel under consideration re-enters the reactor between t h i s moment and
F ( s ) increases linearly: F ( s ) = ( s - n e , +
For P D ( s ) we thus obtain:
@/e.
s = nel,
PD(S) = 0
Equation
for
nel + e =< s =<
(4) i s now substituted into Eq. ( l ) , and for P we set P
The common factor Poet'= cancels out.
+ l)el -
(n
+
=
0, 1, 2,
... .
Then
=
The substitution o = n e ,
e, n
8
- s transforms
into
+
exp(-nOIXi
+
(l/T)]l exp(-B[Ai
(l/T)llj:c
+
exp{[Xi
( l / T ) I o l do
I
\
I
.
and the substitution
-
v
- ne, + 8
= s
transforms
into
exp(-nOl[Ai
+
(l/T)]I explO[Ai
+
+
(l/T)II J;oexpI-[Ai
(l/T)Iol do
.
T h e geometric series exp (-n[hi + (l/T)]O1] can now be summed, and the integrals over o evaluated
b y elementary methods. After performing a l l these operations, one obtains the following inhour formula:
(5)
I
1
pi
(6)
=
xi + (1/T) .
i s obtained by integrating Eq. (3) from 0 t o a. Expressions are obtained which are of the same type
as the ones just discussed, and which can be evaluated by the same methods. The r e s u l t i s
Xid-l+e
5
(7)
~
=
x
i= 1
P
i
-Aie
- e
-hie1
exi(i
(X,O
+
-A
.8
-e
1)
+
e
+e,
- e)
'1
1
169
I n spite of the formidable appearance o f Eqs. (5),
it i s easy to find, for any given I9 and
the value of k e x which produces a given time
constant T .
sum i n Eq.
(5) are
essentially of the form
1-
(6), and (7),
e,,
Furthermore, the following reasoning describes
the general features of the equations. Consider
f i r s t the dependence of k e x on T . If T = a,p , = Xz,
and the sum on the right of (5) i s equal to p, k e x
i s equal t o zero. This corresponds to the state i n
which the reactor i s just c r i t i c a l . If T becomes
very small, the p i become very large and i n the
fraction on the right of (5) the numerator i s dominated by pZ8 and the bracket i n the denominator by
1. Hence the fraction tends to zero l i k e 19/pz, as
T goes to zero, If 7 i s very small, as it i s i n
practice, T w i l l be small as soon as k e x exceeds
p by a small amount. Then the complicated sum on
the right o f (5) i s o f l i t t l e importance, and T i s
determined by r / T = k e x
p, that is, the reactor
period i s inversely proportional t o the excess of
the reactivity over the reactivity corresponding to
the “prompt critical” condition.
-
I n the stationary-fuel reactor, the k e x which
makes the reactor prompt c r i t i c a l i s given by
With circulating fuel, the reactor i s prompt c r i t i c a l
i f k e x = p, which i s less than Z p , , that is, i t
takes less excess reactivity t o make the cirqulatingfuel reactor prompt c r i t i c a l than t o do the same
t h i n g to a stationary-fuel reactor. T h i s i s physically
evident because the fuel circulation renders some
o f the delayed neutrons ineffective.
That p i s
less than Z p , can also be verified mathematically.
E@,.
For T intermediate between very small positive
values and +cot we consider again the analogy to
the stationary-fuel reactor. Here the inhour formula
reads :
’
x
-
1+e-%-
e - a x (x +
1)
+ e-(a-l)x
and an expression of this form can be shown to be
a monotonic decreasing function of positive x ; ~
the expression i s thus a monotonic increasing
function o f T [see Eq. (6)], and as T increases k e x
decreases monotonically, according to Eq. (5);
for every positive k e x there i s one, and only one,
positive T,and vice versa. This, of course, does
not preclude that there exists for a given k e x
several negative T values, i n addition t o the one
positive value. Negative T correspond, however,
t o decaying exponentials, which are of no importance i f the rise i n power i s observed for a
sufficiently long time.
Consider now the behavior of Eq. (5) with variation o f 8, the transit time o f the fuel through the
reactor, and of 8 , ) the transit time o f the fuel
through the whole loop. If 8 (and hence also 6,)
i s very large compared to a l l ’ / A i and l/pi, Eq.
(5) reduces to Eq. (8), the inhour formula for the
stationary-fuel reactor. The circulation i s so slow
that the reactor behaves as i f the fuel were stationary, inasmuch as a l l delayed neutrons, even
the ones with the long-lived precursors, are given
o f f inside the reacting zone, before much fuel
reaches the outside. On the other hand, i f 8 , and,
hence, also 8, i s small compared to a l l l/Xi, that
is, i f the transit time o f the fuel through the complete loop i s small compared t o the mean l i f e o f
even the short-lived delayed-neutron precursors,
then for T >> 8
1
b
c
i
k
k
it;
k
F
For every positive k e x there i s one, and only one,
positive time constant T , and vice versa. T h i s i s
a consequence of the fact that k e x i s a monotonic
decreasing function of T , for i f t h i s monotony did
not exist, there could be several positive T values
corresponding t o a given value of k e x , or v i c e
versa. In the circulating-fuel reactor the situation
i s qualitatively the same; the fractions under the
T h i s i s the same as the inhour formula for the
stationary-fuel reactor, except that a l l the fission
are decreased by the factor I9/0,. This
yields
i s physically easy t o understand, since 6/19, i s
just the probability that a given delayed neutron
i s born inside the reactor.
Of interest i s the intermediate case, i n which I9
301, Eq.
6 W . K. Ergen, The Behavior of Certain Functions Related to the lnhour Formula o Czrculating Fuel Reactors,
O R N L CF-54-1-1 (Jan. 15, 1654);
’See
10.29.1.
170
Glosstone and Edlund, loc. cit.,
See a l s o preceding footnote.
p.
pi
L
i
c
- 6
i;
h
i s smaller than the mean l i f e of the long-lived
delayed-neutron precursors and larger than the
In that
mean l i f e of the short-lived precursors.
case, the long-delayed neutrons act approximately
according to Eq. ( 9 ) and are reduced by the factor
1
8/8,.
On the other hand, the neutrons with the
short-lived precursors behave approximately l i k e
Eq. (8) and are not appreciably reduced. Hence,
a small excess reactivity enables the reactor t o
;
increase i t s power without “waiting” for the not
very abundant long-delayed neutrons. The reactor
goes t o f a i r l y short time constants with surprisingly
small excess reactivities. However, t o make the
reactor prompt critical, that is, to enable i t to
exponentiate without even the I ittle-delayed neutrons, takes a substantial excess reactivity because of the almost undiminished amount of the
latter neutrons.
I
i
I
1
j
I
l
,
1
i
i
?
1
171
Appendix
J
CALIBRATION OF THE SHIM RODS
Three essentially independent methods were
used t o find the value o f the shim rods in terms
of reactivity.
The f i r s t method involved an
analysis of the counting rate data taken during the
c r i t i c a l experiment, and the second was a c a l i bration o f shim rod No. 3 in terms of the regulating
rod.
The agreement between these two methods
was very good. A s a check on the general shape
of the reactivity curves as a function of rod
position, the rods were also calibrated by using
the fission chambers.
T h e various methods are
discussed i n detail below.
C A L I B R A T I O N FROM C R I T I C A L E X P E R I M E N T
DATA
After about one-half the c r i t i c a l mass of uranium
had been added t o the system, the counting rates
o f the two fission chambers and the BF, counter
were taken as a function of shim rod position for
each subsequent fuel addition u n t i l c r i t i c a l i t y was
reached (as described i n the main body of t h i s
Because the fission chamber
report, chap. 4).
counting rates were subject t o the phenomenon
noted i n Appendix H, only the BF, counter data
were used i n the rod calibration.
From the BF, counting rates for each rod position taken after a given fuel injection, the
multiplication hl was determined from the relationship
N
h l = 'V 0
Fig. J . l .
Then, with t h i s plot, a value o f
( & % ) / i n . was obtained for every 1-in. movement
of the shim rods from no insertion t o 16 in. o f
insertion of the rods. A plot of (Ak/k)/in. as a
function of rod position i s given in Fig. J.2, and
the data are tabulated in Table J.l.
I n order t o find the integrated reactivity, or t o t a l
worth of the rods, in terms of ( A k / k ) as a function
of the number o f inches of insertion, the curve
o f Fig. J.2 was divided into three sections. Over
each section the curves were f i t t e d t o a formula
of the form
a
=
Ad2
+
Bd
+
C
t
r
t
t
a
b
,
where a = (Ak/k)/in.
The formula and i t s integral,
which were applied t o the three sections of t h e
curve, are given i n Table 5.2. For each section,
then, the integrated reactivity was found by integrating over the u curves. The integrated curves
were of the form
L
hk
A'd3
P=k=
+
B'd'
+
C'd
.
The three sections into which the curve of Fig.
ORNL-LR-DWG
6420
OF SHIM ROD ( i n )
-r-------INSERTION
10
15
J.2
5
1 0
,
099
where
0 98
N = counting rate for a given fuel concentration and shim rod setting,
*
097
N o = counting rate before start of enrichment.
For each value of M the value of the multiplication
factor k was determined:
P?
0 96
I
0 95
t
7
k = 1
- -MI .
Figure 4.9 of Chapter 4 shows k plotted as a
function o f the uranium i n the systeln for rod
positions of 20, 25, 30, and 35 in.
T h e shim rod calibration was obtained from
Fig. 4.9 by f i r s t making a cross plot of k, against
rod position at the c r i t i c a l mass, as shown i n
172
t
0 94
SHIM ROD POSITION ( i n )-
Fig. J.1. Shim
k at Criticality.
Rod Position a s a Function of
I
T A B L E J.l,
I
CALIBRATION OF SHIM RODS FROM F U E L ADDITION DURING CRITICAL EXPERIMENT
i
I
Rod Position
Average Insertion
Interval T a k e n
(in.)
c
Average
Rod
of Shim
P o s i t i on
Rods
(in.)
Movement of
Average
-&
k
Shim
Rods,
.Ad (in.)
3 Rods
[(Ak/k)/in.]
Average per
[(&/k)/i
Rod
n.1
(in.)
36 to 35
35 to 34
34 to 33
32 t o 33
31 to 32
30 t o 31
29 t o 30
28 to 29
27 to 28
26 to 27
25 to 26
24 t o 25
23 t o 2 4
22 to 23
21 t o 2 2
20 t o 21
35.5
34.5
33.5
32.5
31.5
0.5
1.5
2.5
3.5
4.5
30.5
5.5
29.5
28.5
27.5
26.5
25.5
24.5
23.5
22.5
21.5
20.5
6.5
7.5
8.5
9.5
10.5
11.5
12.5
13.5
14.5
15.5
T A B L E 5.2.
1.o
0.9992
0.9975
0.9953
0.9926
0.9895
0.9862
0.9825
0.9785
0.9743
0.9695
0.9643
0.9585
0.9523
0.9458
0.9386
0.001
0.0015
0.0020
0.0025
0.0028
0.0032
0.0035
0.0038
0.004 1
0.0045
0.0050
0.0055
0.0060
0.0064
0.0069
0.0075
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
0.00033
0.00050
0.00067
0.00083
0.00094
0.00108
0.00118
0.00129
0.00140
0.00154
0.00172
0.00 190
0.00209
0.00224
0.00243
0.00266
b
t
1
t
i
FORMULA (AND INTEGRAND) F I T T E D TO SHIM ROD SENSITIVITY CURVE
Range of Use
Formulo No.
0.001
0.00150
0.00200
0.0025 1
0.00282
0.00 323
0.00355
0.00387
0 .O 04 19
0.00462
0.00516
0.00570
0.00626
0.00672
0.00730
0.00799
(in.)
Formula*
Type
i
i
t
d
*a =
1
0 to 8
D i f f eren ti a1
2
8 to 16
Differential
3
16 to 18
Di fferenti a1
4
0 to 8
I nte gr a /
5
8 to 16
Integral
6
16 to 18
Integral
+ 0.0187d + 0.0239
a = 0.0005d2 + 0.0055d + 0.059
a = -0.0045d2 + 0.16554’ - 1.221
p = -0.0002d3 + 0.00935~2’~+ 0.0239d
p = 0.000167d3 + 0.00275d2 + 0.059d
p = -0.0015d3 + 0.0875d2 - 1.221d
a
=
-0.0006d2
I
1
i
(hk/k)/in.
p = &/k
I
was divided and the corresponding formulas which
were f i t t e d to the curve are given i n Table J.2.
I n the case of the integrated curves, the A k / k
found by integration in the range of the curve was
added t o the t o t a l A k / k of the previous curve t o
give the t o t a l A k / k t o t h e point of integration.
The resulting worth, ( A k / k ) , of the shim rods over
the f i r s t 18 in. of their movement i s shown i n
Fig. J.3 and the data are tabulated i n Table J.3.
I t should be noted that the last 2 in. of the curve
o f Fig. J.2 (16 t o 18 in.) was extrapolated. The
basis for the extrapolation i s the curve shown i n
Fig. D.4 of Appendix D, which i s a calibration of
an ARE regulating rod made during some preliminary experiments done i n the Critical Experiments F a c i l i t y . The shim rods were assumed t o
give the same type of curve.
If it were assumed that the l a s t 18 in. of the
shim rods gave a shape of A k / k vs rod insertion
which was essentially the image of the f i r s t
B
1
i
i
I
173
t
t
i
6421
ORNL-LR-DWG
0 32
030
0 28
-
'
0 26
C
<024
.2"
a 0 22
0
u
r
0 2 0
I
m
+
o 018
__
'CALIBRATION
F R O M FUEL A D D I T I O N
DURING CRITICAL E X P E R I M E N T 1
( A V E R A G E OF A L L R O D S )
5
046
-*5.
-a
044
>
k
1
012
a
oio
+
0
w
0 C R I T I C A L E X P E R I M E N T DATA
0 LOW-POWER E X P E R I M E N T DATA
'+!
[r
008
0 06
0 04
002
0
0
I
2
3
4
5
6
Fig.
J.2.
7
8
9
40
1i
42
43
I N S E R T I O N OF S H I M RODS (In.)
Differential Shim
18
in. (see Appendix D, Fig. D.3), then the total
worth of a shim rod was 5.8% ( A k / k ) and the total
worth of a l l three rods was about 17% ( A k / k ) .
CALIBRATION AGAINST T H E REGULATING
ROD
I n Exp. L-6, w i t h the reactor a t a power of
1 watt (nominal), shim rods Nos. 1 and 2 were
set at predetermined positions.
Then shim rod
No. 3 was moved over various small increments
o f i t s travel and thus calibrated against the regulating rod. With the reactor on servo the regulating
rod then moved automatical l y a compensating
distance, t h i s distance being roughly 10 in. of
i t s travel. From the previous measurement of the
of
worth of the regulating rod of (Ak/k)/in.
174
Rod
.14
15
16
i7
48
49
20
Sensitivity.
0.033Qin. and the ratio of the movement of shim
rod No. 3 and the regulating rod, a calibration of
No. 3 shim rod was obtained over the f i r s t 14
inches of i t s travel (starting from the out position).
F i g u r e J.2 shows the results of t h i s method of
calibration (dashed curve), and the data are l i s t e d
The agreement between the two
i n Table J.4.
methods of calibration was surprisingly good.
Since the increments of shim rod movement were
larger in this method than in the preceding method
and since there were other sources of error, such
as the error i n the movement o f the shim rod
position indicators on the reactor console, no
attempt was made t o use t h i s curve (dashed curve,
Fig. J.2) t o find the integrated value of A k / k over
the rod.
i
?
I
1
3
i
I
i
f
4
i
1
1
1
I
i
1
%
.
r
z
0
IP
P
4
3
w
J
L
'tr
N
03
L
b
i
-0
9
(3
7
'tr
z
U
C
.c
11
L
0
0
Di
>.
->c
I-
U
4
W
n
Di
a:
0
i
?5
L
0
Z
0
I4
3
U
V
U
J.
1
7
w
J
rn
4
c
0 0 0 0 0 0 0 00 0 0 0 0 0 0 0 0 - c
.5
N N r s
Ll
+
+
ccl
"CC
-u
+
L,
i
i
176
?
-I
c
C
L
.-E
X
a
0
W
V
m
lY
.-c
0
3
m
a
lY
0
m
Z
V
.-E
v)
2=
v)
?
P
8
v)
rg
Q:
0
2
d
F
d
0:
v)
OI
9
cy
m
0:
03
2
5
=:
2
2
d
9
W
8
Y
0,
x
0
e
e
4
x
e
c
c
Y
rn
c
'?
9
M
v)
e
h!
m
9
m
F?
:
9
h
cy
9
rr
e
0:
W
Y
P
W
8
7
VI
9
F?
e
9
x
v)
m
9
v)
v)
9
cy
4
cy
9
c
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9
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ai
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2m
2m
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m
9
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9
m,
c
-4
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t
t
I
i
b
t
ORNL-LR-DWG
6423
ORNL-LR-DWG
6422
07
3 4
3 2
30
06
2 8
-
E
26
c
g 0 5
24
0
c-
04
16
2
0
+
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W
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14
$
n
12
03
?
F i g . J.4. Calibration of Shim Rods from F i s s i o n
Chamber Data.
I I O
m
08
0.2
0 6
,
--
0.1:
0
I
0
2
4
6
8
42
10
14
16
18
20
1
TOTAL INSERTION OF SHIM ROD (in.)
I
1
Integral Calibration of Shim
F i g , J.3.
a Function of Position.
Rods a s
5
-s
-
0.'
IC)
C A L I B R A T I O N BY USING T H E FISSION
CHAMBERS
A third calibration of the shim rods was attempted during the c r i t i c a l experiments i n which
the counting rate of the neutron detectors was
taken as a function of shim rod position for two
different uranium concentrations.
The r e a c t i v i t y was then obtained from the
counting rates by using the relationship
1
,i
I
-
-
ORNL-LR
NG 6424
r"=
04
02
'
->
SHIM ROD POSITION ( i n )
0.05
FISSION
1
AMBEF
U235 IN SYSTEM I9 7 Ib
I
1
0
I
~~
5
10
(5
20
:
SHIM ROD POSITION (in.)
Fig.
J.5.
R e a c t i v i t y as a Function of Shim
Rod
Position.
k = 1 --,
M
and a p l o t of k as a function of rod position then
gave a check on the general shape of the curve.
Figure 5.4 shows the k v s rod position for fission
chambers 1 and 2. Because the uranium concentration i n the system was low for both the runs
the f i s s i o n chambers were showing a subcritical
multiplication M greater than the actual value, as
discussed i n Appendix E. Therefore, the absolute
values of k, as calculated from the counting rates,
are in error. Nevertheless, the general shape of
the curve i s experimental verification of the curve
Figure J.5
shown i n Fig. D.3 of Appendix D.
shows the value of (&/%)/in.
v s rod position,
177
as obtained from one of the curves of Fig. 4.10.
Again, the general shape of t h i s curve i s about
the same as shown i n Fig. D.3 of Appendix D,
TABLE 5.5.
U235
Rod
in System
Position
(Ib)
(in*)
16.6
0
5
10
15
20
25
30
30
35
35
0
5
10
15
20
25
27.5
30
32.5
35
36
19.7
178
REACTIVITY OF THE SHIM RODS
Fission Chamber No.
Counting Rate
but i t s absolute magnitude i s not meaningful.
The data from which Figs. 4.10 and 4.11 were
plotted are given in Tables J.5 and J.6.
1
vs
ROD POSITION
Fission Chamber No.
Counting Rate
NO/N
2
BF3 Counter
k
Counting Rate
NO/N
k
NO/N
k
17.52
17.60
18.85
20.93
23.44
25.23
26 .OO
25.87
26.83
26.85
0.5457
0.5375
0.501 9
0.4520
0.4036
0.3750
0.3538
0.3557
0.3526
0.3523
0.4543
0.4625
0.498 1
0.5480
0.5964
0.6250
0.6362
0.6343
0.6474
0.6477
25.00
25.52
26.69
31.33
35.63
38.91
40.93
41.12
41.76
41.65
0.6512
0.6379
0.6100
0.5196
0.4569
0.41 84
0.3978
0.3952
0.3898
0.3909
0.3488
0.3621
0.3900
0.4804
0.5431
0.5816
0.6022
0.6048
0.601 1
0.6091
6.00
6.05
6.51
6.40
6.67
7.01
0.7483
0.742
0.690
0.702
0.673
0.640
0.2517
0.258
0.310
0.298
0.327
0.360
6.72
6.80
6.77
0.668
0.660
0.663
0.332
0.340
0.337
18.99
19.31
21.07
23.63
26.21
28.08
29.23
29.55
29.76
29.76
29.90
0.4982
0.4899
0.5018
0.5101
0.551 0
0.5997
0.6391
0.6631
0.6764
0.6793
0.682 1
0.6821
0.6 836
26.83
26.88
29.28
34.43
38.99
42.67
44.59
45.97
46.19
46.8 3
46.81
0.6068
0.6057
0.5560
0.4728
0.4175
0.3815
0.365 1
0.3541
0.3525
0.3476
0.3478
0.3932
0.3943
0.4 44
0.5272
0.5825
0.6185
0.6349
0.6459
0.6475
0.6524
0.6522
6.27
6.43
6.56
6.91
7.04
7.28
7.15
7.01
7.04
7.09
7.16
0.716
0.698
0.684
0.650
0.638
0.617
0.628
0.640
0.638
0.633
0.627
0.284
0.302
0.316
0.350
0.362
0.383
0.372
0.360
0.362
0.367
0.373
~(counts/sec)
0.4490
0.4003
0.3609
0.3369
0.3236
0.3201
0.3179
0.3 179
0.3164
N (covnts/sec)
N (counts/sec)
~
i
T A B L E J.6.
!
u~~~in
System
(Ib)
R E A C T I V I T Y AS A FUNCTION O F SHIM ROD POSITION (FROM FISSION CHAMBER DATA)
L i m i t s of
d
Ak
da V
t
19.7
I
4
1
1
0 to 2
2 to 4
4 to 6
6 to 8
8 t o 10
10 t o 12
12 t o 14
14 t o 16
16 to 18
18 to 20
20 t o 22
22 t o 24
24 t o 26
26 to 28
28 to 30
30 t o 32
32 to 34
34 to 36
1
3
5
7
9
11
13
15
17
19
21
23
25
27
29
31
33
35
0 SO2
0.5065
0.5 125
0.523
0.5 32
0.555
0 -574
0.596
0.6 155
0.6 32
0.6465
0.6575
0.6655
0.6715
0.676
0.6795
0.6815
0.6825
0.04
0.05
0.08
0.12
0.16
0.20
0.21
0.22
0.18
0.16
0.13
0.09
0.07
0.05
0.04
0.03
0.0 1
0.01
2
2
2
2
2
2
2
2
2
2
2
2
2
2
2
2
2
2
0.0398
0.0494
0.07 80
0.1 147
0.1503
0.1801
0.1829
0.1853
0.1462
0.1266
0.1005
0.0684
0.0526
0.0372
0.0296
0.022 1
0.0073
0.0073
c
k
L
i
i
Appendix
f
K
i
CORRELATION OF REACTOR AND LINE TEMPERATURES
An effect which was noted late i n the operation
of the ARE was that fuel and sodium line temperatures indicated i n the basement disagreed quite
radically with those recorded i n the control room.
It s o happened that a l l the basement indicators
gave line temperatures measured by thermocouples
outside the reactor thermal shield, while the
temperatures recorded i n the control room were
measured by thermocouples a l l located within the
thermal shield, When t h i s was f i r s t discovered it
was thought that helium from the rod-cooling system
blowing on the thermocouples inside the thermal
shield was making them read low. Turning the rod
cooling blowers on and off, however, was demonstrated t o have no effect on the thermocouples, and
therefore a t the end of the experiment no positive
explanation for these temperature discrepancies
had been found.
This situation has created serious difficulties i n
trying t o analyze the data for this report. There
was evidence that the discrepancies were intensified during operation at high power. Nevertheless,
when nearly isothermal conditions existed, i t was
found that the absolute magnitude of the measurements made by the thermocouple within the thermal
shield were incorrect, but the rates o f change
obtained from the data were correct. The temperatures read on the basement instruments were
considered to be correct for several reasons. First,
the temperature indications available i n the basement were much more numerous than those i n the
control room, and the thermocouple sensing elements
for these indicators were a l l located along lines
outside the thermal shield; the temperatures agreed
with each other t o k10 deg from the average. Equilibrium conditions prevailed across the pipes because of the insulation,
Also, there were many
reasons for the thermocouples to indicate higher
than actual temperatures, but none for lower than
act ua I indications.
I n order to use the temperature data from the
experiment i t was necessary to correlate the temperature data obtained i n the control room with
those obtained i n the basement. The correlations
were then used to correct temperature data obtained i n the control room.
The results o f the temperature correlations for
the fuel system are shown i n Figs. K.l, K.2, and
K.3.
The agreement between the low-power l i n e
180
temperatures (Fig. K . l ) reflects the attention which
was given t o calibrating these thermocouples i n
the isothermal condition. The curves in Figs. K.2
ORNL-LR-DWG
P
b-
6563
1350
F
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,I
OI
O
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J
4
5
1300
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I
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D
II)
z
W
rn
3
$
1250
a
5
I
w
3
OUTLET TEMPERATURE
8 INLET TEMPERATURE 1
1200
1200
13 0 0
4250
1350
P
FUEL LINE TEMPERATURES OUTSIDE THERMAL SHIELD (OF)
i.
h
Fig. K.1. Correlation Between F u e l L i n e Temperatures Measured Inside and Outside the Reactor
Thermal Shield During Subcritical and Low-Power
Experiments.
F
k
ORNL-LR-DWG
6564
?i
t
I
I
I
w
J Wn
:z
b
z
i
1200
I300
I400
1'500
1600
FUEL LINE TEMPERATURES OUTSIDE THERMAL SHIELD (OF)
h
k
Fig. K.2. Correlation Between F u e l L i n e Temperatures Measured Inside and Outside the Reactor
Thermal Shield During High-Power Experiments.
,r
i
.-lL
ORNL-LR-DWG 6567
n
i
1350
w
I
t
v,
A
a
I
1
i
W
I
(300
P
v,
!
z
In
LY
W
2+ 1250
W
a.
z
kW
k
W
z
I 1200
8
1200
Fig. K.3. Correlation Between Fuel Temperature
Differentials Measured Inside and Outside the
Thermal Shield.
ORNL-LR-DWG
-
LL
I400
c
ORNL-LR-
150
D W G 6568
lL
e
n
n
A
A
w
r
wI
v)
v)
A
i
A
a
4
I
a
(300
k-
IO0
kI
W
P
W
v)
kI
W
W
v)
P
cc
v)
z
k3
5
I350
Fig. K.5.
Correlation Between Sodium L i n e
Temperatures Measured Inside and Outside the
Thermof Shield During High-Power Experiments.
6566
e
E
r
I300
SODIUM LINE TEMPERATURES OUTSIDE THERMAL SHIELD (OF)
FUEL LINE AT OUTSIDE THE THERMAL SHIELD (OF1
- 4350
1250
IC
1250
a
a
W
+
z
W
I
50
W
I
6
2
_1
n
a
t
v)
0
n
2
0
(200
1300
1350
I400
SODIUM LINE TEMPERATURES OUTSIDE THERMAL SHIELD (OF)
(250
Fig. K.4.
correlation Between Sodium L i n e
Temperatures Measured Inside and Outside the
Thermal Shield During Low-Power Experiments.
-
and K.3, both of which were obtained from data
taken during operation at high (>200 kw) power,
show the anomaly in question.
The analogous data for the sodium system are
It i s evident
shown i n Figs. K.4, K.5, and K.6.
from Fig. K.4 that an isothermal condition was
never attained i n the sodium system, since both
0
0
50
IO0
(50
SODIUM LIME A T OUTSIDE THE THERMAL SHIELD (OF1
Fig. K.6. C o r b l a t i o n Between Sodium Temperature Differentials Measured Inside and Outside
the Thermal Shield.
the inlet and the outlet line temperature correlations have slopes of 1 but intercepts of -25 and
-40°F, respectively. T h i s was probably the result of the long sodium i n l e t and outlet lines being
used as a convenient means of adding heat to the
system to maintain a thermal equilibrium i n the
Good data for the high-power
reactor of -130OOF.
181
a
I
h
h
1
1
correlations, Figs. K.5 and K.6, were rather scarce,
but the curves shown are believed to be reasonable
approximations t o the actual correlation.
The
data i n Fig. K.6 have been corrected for known
error in the temperature differential a t zero (or low)
power. The correction, as applied, was t o reduce
the value of the temperature differential determined
from the thermocouples outside the thermal shield
by 1OOF.
A l l the correlations were based upon data which
were obtained during runs long enough for the
establishment of equi librium conditions i n the
system or at times when isothermal conditions
prevailed. The most important conclusion that can
be made from these data i s that the temperature
differentials across the reactor i n both the fuel
and sodium systems were about a factor of 2 low.
In the fuel system the outlet l i n e temperature
changes read in the control room (inside pressure
shell temperatures) were only one-half as great as
the corresponding temperature changes read i n the
T A B L E K.l.
F U E L LINE T E M P E R A T U R E S MEASURED BY THERMOCOUPLES INSIDE A N D
OUTSIDE T H E REACTOR THERMAL SHIELD
F u e l Inlet L i n e
Date
Time
E xper i me nt
Temperatures (OF)
(Line
No.
2115
Before
F u e l Outlet L i n e
(OF)
Temperatures
120)
(Line
111)
Temperature
D i f f e r e n t i a l (OF)'=
Control Roomb
Basementn
Control R o o m b
1290
1285
1290
1287
0
2
1293
1293
1296
1293
3
0
Basementn
10/26
basement (outside pressure shell temperatures). In
the sodium system both outlet and i n l e t line temperature curves had different intercepts but the
same slopes (as during the low-power runs).
All temperature data used i n t h i s report that
pertained to operation at high power were corrected
t o agree with line temperatures obtained outside
the thermal shield (basement readings) by using
Data from runs made
curves K.l through K.6.
under isothermal or equilibrium conditions where
equal temperature differencss were obtained on
instruments either i n the control room or i n the
basement needed no temperature corrections. The
temperature correlation data from which the curves
were obtained are presented i n Tables K.l and
K.2. Because of the lack of coordination between
the control room and basement operations during
the experiment, much potentially useful data had
t o be discarded because the exact times of the
readings were not known or the data were taken
before equilibrium conditions were established.
Basement
Control Room
operation
10/27
0630
Before
operation
11/4
1100
L- 1
1300
1299
1297
1295
-3
-4
11/6
0340
L-4
1305
1310
1304
1300
-1
-10
11/11
0300
H-11
1207
1212
1522
1418
315
206
1215
H-11
1210
1214
1524
1418
314
2 04
0001
H-13
1316
1308
1323
1308
7
0
0630
H-13
1315
1307
1320
1306
5
1003
H-14
1246d
1243
1587
1475
341
232
1264
1260
1260
1250
-4
-10
11/12
11/13
0915
After
-1
operation
'Fuel
line temperatures outside thermal shield.
b F u e l line temperatures inside thermal shield.
cThermocouple differences for no-power runs indicate extent of thermocouple errors (highest i s
the accuracy of the readings.
d E s t imated.
182
f l 0 d e g ) and therefore
T A B L E K.2.
?
SODIUM L I N E TEMPERATURES MEASURED BY THERMOCOUPLES INSIDE AND
OUTSIDE THE REACTOR THERMAL SHIELD
i
~~~~
Date
Time
Experiment
Sodium I n l e t L i n e
Sodium O u t l e t L i n e
(OF)
Temperatures (OF)
Temperatures
~~
-
Temperature
Differential
(OF)
F(0.
Basement
Control Room
Basement
C o n t r o l Room
Basement
Control Room
i
10/26
2115
J
L
Before
1305
1278
1320
1278
15
0
i
operation
10/27
0630
Before
1312
1292
1325
1290
13
2
operation
11/4
1100
L-1
1320
1295
1330
1288
10
7
11/6
0400
L-4
1325
1300
1335
1295
10
5
11/11
0900
H-11
1223
1224
1332
1271
109
47
11/12
1003
H-14(1)
1246
1258
1369
1307
123
49
11/9
1907
H-3(7)
1313
1290
1373
1313
60
23
11/13
0915
1282
1258
1292
1252
10
-6
After
operation
1
3
3
3 -
4
1
A
i
T h i s points out the need in future operations of the
scale of the ARE for some sort o f timing device
that would automatically stamp the date and time
every few minutes on the charts o f a l l recording
instruments, and a more systematic method of
manually recording data for each experiment performed i n which conditions are held constant long
enough for a l l pertinent data to be recorded. These
and other similar recommendations w i l l be found
in Chapter 7 of t h i s report.
f
E
L
POWER DETERMINATION FROM HEAT EXTRACTION
Appendix
The power level o f the reactor was determined
from the total heat extracted by the fuel and the
sodium, Data for the heat extraction determination
were obtained from continuous records of the reactor fuel inlet, outlet, and mean temperatures,
the temperature differential across the reactor,
and the flow rates o f the fuel and the sodium.
The flow rates could also be calculated from the
pump speed, and there was independent experimental evidence from which a plot of speed v s
flow rate was obtained.
The power level of the
reactor was calculated by use of heat capacities,
flow rates, and temperature differentials for both
sodium and fuel. It was found that the fuel and the
sodium accounted for 99% of the extraction of
power generated i n the reactor, The data referred
to above were a l l obtainable i n the control room.
The controls for preheating and maintaining heat
on the system were located i n the basement, and
along with these controls were temperature recorders and indicators for the whole system, exclusive of the reactor, The temperatures o f both
fuel and sodium lines to and from the reactor were
also recorded i n the basement.
These basement
data were used for a separate power extraction
determination.
An independent source of power level information
was also available, since the fuel and sodium were
cooled by a helium stream flowing over a heat exchanger and the helium in turn was cooled by a
No flow or temhelium-to-water heat exchanger.
perature measurements were made of the helium,
but since the water exchangers were very close t o
the fuel or the sodium exchangers, a l l the heat that
was taken out of the fuel and sodium should have
appeared i n the water.
The water flow through
these exchangers was metered by orifice-type
flowmeters, and the outlet temperatures were measured by thermocouples i n wells.
The fuel loop
exchangers and the two sodium loop exchangers
were metered separately.
Up to the time of the 25-hr xenon run the reactor
was operated a t high power for very short periods
o f time, and the extracted power was determined
from the control room data. During the 25-hr xenon
run a comparison was made of the extracted power
determined from the control room data, that determined from the basement data, and that from
Large discrepancies
the water data (Table L.l).
184
3
were found, and the indications were that the
control room data were low.
The disagreement o f the various data led t o the
examination of the i n l e t and outlet l i n e temperatures, as described i n Appendix K. During the
xenon run the fuel outlet l i n e temperatures were
about 100°F higher than the fuel outlet manifold
temperatures (control room data), and the fuel
inlet
line temperatures (basement data) were
essentially the same as the fuel i n l e t manifold
temperatures (control room data).
The sodium
outlet line temperature (basement data) was 6OoF
higher than the sodium outlet temperature a t the
reactor (control room data).
(The only check on
power extraction by the rod cooling system, which
was only 1% of total power, was the water data
for the rod cooling helium-to-water heat exchanger.
In most discussions t h i s 1% i s neglected.)
The power extracted i n both the fuel and the
sodium systems has been calculated by using
temperature data from several experiments, and the
There were
results are presented i n Table L.2.
three separate measurements of the fuel temperature
differential available i n the control room which
were usually i n fair agreement, The largest discrepancy was, as mentioned, between the control
room data and the basement data, The temperature
differential obtained from the basement data was
an average result from a number o f l i n e thermocouples, while the control room data came from
thermocouples located on the fuel headers inside
the reactor thermal shield.
The so-called secondary heat balance was obtained by determining the heat dumped by the water
from the various heat exchangers. These data are
tabulated i n Table L.3 for the high-power experiments.
The various estimates of reactor power from
both primary and secondary heat extraction are
then listed i n Table L.4, together w i t h the power
estimated from the calibration of the nuclear instruments, that is, the log N recorder and the micromicroammeter.
These latter instruments were
normalized to agree w i t h the primary extracted
power as determined by the l i n e temperature
(basement data) during the xenon experiment (H-11).
Equilibrium conditions were certainly attained
during t h i s 25-hr experiment and such error as there
may be i n using these data as the criteria for
b
.
f
P
b
-
i
t.
t
i
i
L
Ir
b
4
k
c
E
b
L
_rr
4
3
d
TABLE L.l.
i
COMPARISON OF POWER EXTRACTION DETERMlNATlONS MADE FROM DATA
OBTAINED DURING 25-hr XENON RUN
Temperatures
4
a
f
‘
e F)
Flow
Power Extracted
(gpm)
(Mw)
Inlet
Outlet
AT
1212
1209
1418
1522
206
313
44
44
1.02
1.52
1225
1226
1271
1335
46
109
153
153
0.244
0.577
T o t a l Power‘
Extracted
(Mw)
F u e l System
Control room data
Basement dato
Sodium System
Control room data
Basement data
d
Total Power
Control room data
1.28
2.12
Basement data
Water Data
61
61
I n fuel loop
In sodium loop
56
53
117
114b
205
38.36
1.68
0.588=
2.28
T o t a l Power from
Water D a t a
‘Includes
J
i
-1% for rod cooling.
bAverage for both loops.
‘-Sum
of both loops.
TABLE L.2. PRIMARY POWER EXTRACTION
4
153
1
44%
40
33.5
0.196
0.163
0
0.199
#
2
44%
66
52
0.323
0.254
0
0.308
4
3
44%
115
116
0.562
0.567
0
4
44%
120
114
124
0.588
0.560
0.606
153
14
5
44b
230
231
239
1.128
1.129
1.17
153
10.5 0.0548
6
44$
237
234
243
1.16
1.145
7
44$
13
19
20
H-6
44$
236
240
245
H-8
45
266
225
H-ll
44
206
213
182
245
H-3
.i
H-13
44
H-14
45
21 1
0.584
0.0734
0.678
1.19
152
IO
0.0523
1.24
0.0653 0.0946
0.0975
152
22
0.115
0.221
1.158
1.175
1.198
152
1.31
1.111
0.984
1.017
IO
0.900
1.018
198 0.970
45
0.236
1.23
138 0.677
60
0.315
1.012
1.21
0.0182
1.22
320
1.572
153
21
0.110
1.34
293
1.452
153
46.4
0.244
1.27
313
1.520
0.0757
312
1.530
19 0.099
1.65
1.45
355
1.760
127 0.667
2.45
0.0484
153
1.212
153
3.5
1.4 0.0073
50
0.262
75 0.394
1.87
0.577
2.12
110
185
TABLE L.3.
SECONDARY POWER EXTRACTION
No. 2 Sodium
Heot Exchanger
No. 1 Sodium
Heat Exchanger
Fuel Heat Exchanger
Rod-Cooling
Heat Exchanger
Run
H-3‘
1
204
13
0.389
38.4
0
0
38.4
0
0
17.3
3
0.0076
0.397
2
204
13
0.389
38.4
0
0
38.4
0
0
17.3
3
0.0076
0.397
0
38.4
0
0
17.3
3
0.0076
0.905
3
204
30
0.897
38.4
0
4
204
29
0.866
38.4
26
0.146
38.4
21
0.1188
17.3
3
0.0076
1 .14
5
204
59
1.765
38.4
26
0.146
38.4
25
0.141
17.3
3
0.0076
2.06
6
204
59
1.765
38.4
27
0.152
38.4
24
0.135
17.3
3
0.0076
2.06
7
204
3
0.0903
38.4
26
0.146
38.4
25
0.141
17.3
2
0.0051
0.382
1.901
38.0
37
0.207
38.4
40
0.226
17.3
3
0.0076
2.34
0.228
17.6
2
0.0052
2.28
0.0056
17.1
0.316
16.9
6
0.0150
2.53
H-6
204
H-8
206
63
H-11
205
56
1.685
38.2
55
0.300
38.4
51
H-13
205
3
0.0902
38
10
0.0561
38.4
1
H-14
204
63
1.883
38
57
0.318
38.4
56
T A B L E L.4.
Primary Power (Mw)
Experiment
Run
No.
N 0.
N
Secondary
Basement
Control Room D a t a
p A ~ p~
Micromicroammeter
~
VAT
Water
H e a t Exchanger,
‘av
pB
pW
~
1
0.449
0.425
0.216
0.183
0.199
0.3966
2
0.449
0.463
0.343
0.274
0.308
0.3966
3
0.848
0.910
0.582
0.587
0.584
0.9046
4
0.923
0.988
0.681
0.653
0.699
0.678
5
2.47
2.05
1.20
1.20
1.24
1.21
6
2.12
2.34
1.23
1.22
1.26
1.24
7
0.281
0.288
0.200
0.230
0.232
0.221
H-6
2.21
2.16
1.20
1.21
1.25
1.22
H-8
2.12
2.38
1.44
1.24
1.34
1.87
2.342
1.28
1.27
2.12
2.278
H-3
H-11
2.12
2.12
H-13
0.125
0.151
H-14
2.41
2.53
P
P o w e r (Mw);
(Mw)
Log
- 1
REACTOR POWER SUMMARY
N u c l e a r Power
1.25
1.28
0.0757
1.42
amount of extracted power i s conservative, since
the water heat balance at the same time showed
the power to be 7% higher.
T h e sources of error in the various measurements
of extracted power were associated with the tem-
186
Total
Extracted Power
Exwrirnent
1.45
1.23
2.0596
1.012
2.0596
0.3824
1.93
0.1562
0.0757
1.43
1.1384
2.45
2.53
perature measurements on the fuel and sodium lines
a t the reactor, which gave the reactor AT. These
thermocouples were unique in several respects;
t h a t is, they were located within the reactor thermal
shield; they were exposed to the reactor pressure
i
i
i
shell; they were exposed to high nuclear radiation
fluxes, etc.
These unique aspects immediately
suggest several possible explanations for the
erroneous temperature measurements.
However,
upon further examination, each of these aspects,
w i t h the dubious exception of nuclear radiation,
has been shown t o be incapable of producing the
observed anomaly.
f
i
1
i
The control and shim rods and the fission
chambers were cooled by forced helium circulation
i n the rod-cooling system. Part of the helium that
was blown down through the rod tubes was deflected back up across the outlet manifold and
between the pressure shell and the thermal shield.
While i t w a s thought that perhaps this was cooling
the fuel outlet manifold thermocouples and making
them read low, this was disproved when changing
the speed of the blower or stopping it had no effect
on the fuel outlet temperature.
It was thought that possibly the heat radiation
from the fuel outlet manifold to the colder bottom
o f the pressure shell was great enough to actually
lower the manifold w a l l temperatures 100°F below
the fluid temperature. T h i s has been disproved by
heat transfer calculations. I t was also thought that
since the outlet fuel lines passed through the
2-in. plenum chamber that the resulting surface
cooling of the fuel might account for the lower
wal I temperatures, although the mixed mean fuel
temperature was considerably higher.
T h i s was
also disproved when no increase i n w a l l temperature was observed for thermocouples w i t h i n the
thermal shield but located progressively farther
away from the pressure shell bottom.
I t i s of interest that after the final shutdown o f
the reactor, the fuel outlet line and fuel outlet
manifold temperatures agreed; i n other words, w i t h
no power generation the basement data and control
room data agreed.
f
4
187
i.
I
t
Appendix
f
t
M
t
THERMODYNAMIC ANALYSES
Some two years before the Aircraft Reactor Experiment was placed in operation a comprehensive
report was written on the “Thermodynamic and
Heat Transfer Analysis o f the Aircraft Reactor
Experiment.”’
Not only was the design of the
reactor system based, in part, on the studies culminating in that report, but the material therein
served as a guide during the acceptance test o f
the experiment.
Therefore a comparison of the
calculated and the actual thermodynamic performance of the reactor system was attempted, It
soon became apparent that the experimental data
o f the type needed for thermodynamic analyses
were inadequate to permit a meaningful comparison
with any of the calculated situations. Perhaps the
most valid comparison may be made between the
calculated insulation losses, the heater power
input a t equilibrium, and the heat removed by the
space coolers. Even here the agreement i s less
An illustration of the inefficacy o f
than 50%.
attempting to calculate such thermodynamic constants as the heat transfer coefficients of the fuel
and the sodium i s presented below.
and not the actual insulation which had cracks,
clips, etc, Second, the electrical heat load was
high, since it did not allow for transformer, variac,
and l i n e losses. Also, the space cooler heat load
may not have represented an equilibrium condition
if the p i t walls were s t i l l heating up and/or radiating heat,
i
e,
k
.-
c-
t
k
EXPERIMENTAL VALUES O F HEAT
TRANSFER C O E F F I C I E N T S ~
6
f
A n attempt was made tocalculate both the sodium
and the fuel heat transfer coefficients from the
measured values o f temperatures and flows in the
various heat transfer media, that is, fuel, sodium,
helium, water, These data are given i n Table M.1
for a typical operating condition.
The desired heat transfer coefficient should be
determinable from the calculated values o f the
various thermal resistances i n the heat exchangers.
T o obtain the thermal resistance on the fuel side
or the sodium side o f the respective heat exchangers, the helium side and metal resistances
had t o be subtracted from the over-all resistance.
The expression for the over-all resistance i s
I N S U L A T I O N LOSSES, H E A T E R POWER I N P U T ,
A N D S P A C E COOLER PERFORMANCE
A t equilibrium the electrical heat required t o
maintain the system a t a mean temperature o f
1325OF should have been equivalent t o the heat
removed i n the eight p i t space coolers. The maximum amount of heat ever removed by the space
coolers was the250 kw attained with 66.5-gpm-total
water flow with a temperature rise of 3OOF. AIthough this value agrees very well with calculated’
heat loss of 220 kw for the entire system (the
reported value of 240 kw was calculated for an
earlier design system and was reduced by about
20 kw for the actual system), the actual electrical
power input t o the heaters averaged around 375 kw
after the system reached equilibrium.
There are several possible explanations for the
discrepancies between the calculated heat loss,
the electrical power input, and the heat removed
by the space coolers. First, the calculated loss
was low because it assumed idealized conditions
.
’
B. Lubarsky and B. L. Greenstreet, Thermodynamic
and Heat Transfer Analysis of the Aircraft Reactor E x periment, ORNL-1535 (Aug. 10, 1953).
188
where
U = over-all heat transfer coefficient,
A = area across which heat i s transferred,
L
e#i?
q = rate of heat transfer,
AT
= over-all temperature difference.
p
A knowledge o f the i n l e t and outlet temperatures
on the helium side of the heat exchanger i s required by the above expression, and lack o f t h i s
information immediately introduces a large unOf greater importance,
certainty i n the results.
however, i s the relative magnitude of the resistances involved. The calculated r a t i o for the fuelto-helium exchanger i s
c:
*
RHe
-= 10.3 ,
Rf
where R H e i s the resistance on the helium side
r
i
b
t
orL
2H. H.
Hoffman, P h y s i c a l Properties Group, Reactor
E x per irnen to I Engineering D i v i si on.
EXPERIMENTAL HEAT TRANSFER DATA IN F U E L AND SODIUM LOOPS
TABLE M.1.
Primary Coolant
Helium
Temperature
Flow
s'
4
Water
Temperature
Flow*
(OF)
(cfm)
(gpm)
(OF)
Temperature
Flow
(OF)
(gpd
Inlet
Outlet
44 (?5%)
1209 f 10
1522 ? 10
8000 (f30%)
No data
194 (*5%)
61
f2
117 k 2
152 ( f l % )
1225 f 10
1335 f 10
1700 (f20%)
No data
76.6 ( k l % )
61 f 2
114 f 2
.
Inlet
Outlet
Inlet
Outlet
4
F u e l loop
i
Sodium loop
~
* E s t i mate d
.
and R i s the resistance on the fuel side, and for
f
the sodium-to-helium exchanger i s
--
He
- 65
RNa
The liquid-side resistance i s
.
where R m i s the w a l l resistance,
Hence, the
liquid-side thermal resistance i s the small difference between two much larger numbers. A small
error i n R T or R H O i s greatly magnified in R l .
Unfortunately, the error i n R H O i s not small, being
perhaps as great as 60%. Thus, except for over-all
heat balances, l i t t l e useful heat transfer information can be extracted from the available data on
the ARE heat exchangers.
Appendix
COMPARISON
N
OF REACTOR POWER DETERMINATIONS
W. K. Ergen
In order to get an estimate of the reactor power
and t o be able to calibrate the instruments, the
ARE was run for 1 hr a t a power which was e s t i mated, a t the time, to be 10 watts (exp. L-4). A
sample of the fuel was then withdrawn and the
gamma activity was compared with that of a sample
which had been irradiated at a known power level
in the Bulk Shielding Reactor (BSR). This method
o f power determination i s discussed i n Appendix H.
A curve showing the gamma counting rate of a
BSR-irradiated sample as a function of time after
shutdown had been obtained. The irradiation time
was 1 hr a t a constant power of 1 w. From this
curve a decay curve corresponding to the actual
power history of the ARE was synthesized by
taking into account not only the “10-watt” run,
but also the previous lower power operation.
When this synthetic curve was compared with the
one obtained from measurements on the ARE sample, the shapes of the curves did not agree, the
ARE curve having a smaller slope than the synthetic curve, Also, the power determiped by t h i s
method was lower than the power ultimately determined by- the heat balance. Both th9se effects
can be explained i n a qualitative manner by the
loss of some of the radioactive fission fragments
from the ARE sample. This would reduce the total
radioactivity of the sample and hence the apparent
power level. Furthermore, the loss of radioactive
fission fragments would reduce the counting rate
a t short times after irradiation more than a t long
times, because among the most volatile fission
fragments are the strong gamma emitters that have
relatively short half 1 ives (notably, 2.77-hr Kr88).
T h i s would flatten the slope of the decay curve
o f the ARE sample.
It has not been possible so far t o treat t h i s
matter i n a quantitative way, but i n order to eliminate some computational complications a sample
was irradiated i n the BSR under conditions exactly
duplicating the power history of the ARE, except
for a proportionality factor. A comparison of the
decay curve of this sample and that of the ARE
sample is shown in Fig. N.1.
The BSR sample
contained 0.1166 g of U235; the fission cross
section at the temperature of the BSR i s 509 barns.
The BSR power at the final 1-hr irradiation was
’
10 w, corresponding t o a flux2 of 1 . 7 ~l o 8 n/cm2-sec,
and a self-shielding factor, estimated t o be 0.8,
had to be applied t o this sample. Hence, during
the 1-hr irradiation there were
0.1166 g x 0.6 x
atoms/g-atom x
509
fissions/(atoms*n/cm2) x 1.7
x
x 108 n/cm2.sec x 0.8/235 g/g-atom
=
.
20.6 x I O 6 fissions/sec
The ARE sample contained 0.1177 g o f U235and
the total U 2 3 5 i n the ARE a t the time o f the run
The reactor power, as determined
was 59.1 kg.
later by the heat balance, was actually 27 w,
instead of the expected 10 w. Hence there were
27 w x 3.1 x 10” fissions/w.sec
:
c
t
x 0.1177 g
59,100 g
=
1.66 x
l o 6 fissions/sec
in the ARE sample.
The r a t i o of the radioactivities of the BSR
sample and the ARE sample should thus have been
20.6/1.66 = 12.4. As may be seen from Fig. N.1,
the measured ratio i s 27 a t 1 hr 40 min, and 17.5
a t 38% hr. As pointed out above, the discrepancy
can be explained by the loss of radioactive fission
fragments from the ARE sample.
A small amount of the radioactivity o f either
sample was contributed by the capsule and the fuel
carrier,
especially the sodium content of the
carrier.
This was measured by irradiating a nonuranium-bearing capsule with carrier, and countIng
i t s radioactivity.
However, since this correction
proved to be small and since both samples contained about the same amount o f uranium i n the
same amount of carrier, the results would not be
appreciably affected.
P
li
i%
k
1
!
- a
’ J . L. M e e m , L. 6. H o l l a n d , a n d G. M. M c C a m m o n ,
Determination of the Power o / the sulk Shielding R e actor, Part 111. Measurement o/ the Energy Released
per Fission, ORNL-1537 (Feb. 15, 1954).
2 E . 8. J o h n s o n ,
b
private c o m m u n i c a t i o n .
190
f
I
4
In
d
t
0
In
rn
L
.c
B
3
Z
In-
N
II>
I
m
w
K
4
"o Ew
In
c
Q
In
0
Appendix
0
ANALYSIS 0 F TEMP ERATUR E COE F FICI ENT MEASUREMENTS
IMPORTANCE OF T H E F U E L TEMPERATURE
COE F F l C l E N T
One of the most desirable features o f a circulating-fuel reactor i s i t s inherent s t a b i l i t y because
of the strong negative temperature coefficient of
reactivity.
T h i s was conclusively demonstrated
by the ARE. It had been predicted, on the basis
of the temperature dependence of the density,
which was
p (g/cm3)
3.98
=
-
0.00093T
(OF)
for the final fuel concentration i n the reactor, that
the A R E fuel would have a negative temperature
coefficient o f sizable magnitude. A t the operating
temperature of 13OOOF the fuel density was 3.33
g/cm3.
The resulting mass reactivity coefficient
was
(AM/M)PF =
-
0.00052
3.33
=
-1.56 x 10'4PF
,
and, for Ak/k = 0.236 AM/M, the predicted temperature coefficient that would result from the changing
density of the fuel alone was
(Ak/k)/'F
=
-3.68
x 10-5/0F
.
Actually, as was stated i n the body of t h i s report
(cf., Fig. 6.4), the fuel temperature coefficient was
loop helium blower speed was increased there was
always a sudden marked increase in reactivity. I n
attempting to determine t h i s instantaneous temperature coefficient by observing the rod movement
by the servo as a function of the fuel temperature,
values of the fuel temperature coefficient of a
could be
magnitude much larger than -9.8 x
obtained.
The change i n the apparent temperature coeff i c i e n t as a function of time during experiment H-4
i s shown i n Fig. 0.1. The time intervals chosen
were of 15-sec duration. The apparent peak temperature coefficient o f r e a c t i v i t y was i n excess of
-3.5 x
(Ak/k)/OF, and after 5 min the coefficient leveled out t o a value of about -6 x loa5.
A more complete discussion of t h i s experiment i s
given below i n the section on "High-Power Measurements of Temperature Coefficients of Reactivity," i n which the time lag of the thermocouples
i s taken into account. Two coefficients were obtained: an i n i t i a l fuel temperature coefficient and
an over-all coefficient, which was the asymptotic
given by Fig. 0.1.
value (-6 x
The effect shown i n Fig. 0.1 was due partly t o
geometrical considerations. By reference to Figs.
2.2 and 2.3, i t can be seen that as the fuel entered
the reactor i t passed through the tubes closest to
-9.8
10-~PF.
A s long as the over-all temperature coefficient
o f a reactor i s negative, the reactor w i l l be a slave
t o the load demand. However, the fuel temperature
coefficient was the important factor for reactor
control i n the ARE because i t took a significant
time for the bulk of the material of the reactor to
change temperature and, therefore, for the over-al I
caefficient to be felt. In experiment H-5 (cf., Fig.
6.4), it was found that the fuel temperature coefficient predominated for 6 min.
E F F E C T O F GEOMETRY I N T H E ARE
When the reactor was allowed t o heat up by
nuclear power, as i n experiment H-5, the measurements o f the temperature coefficient were quite
definitive.
The experiments i n which the heat
extraction by the helium blower was suddenly
increased, i.e., experiments E-2, L-8, and H-4, did
not give clear-cut results. I n fact, it was noticed
throughout the operation that whenever the fuel
192
2122
2123
2124
2125
2426
TIME OF DAY, NOV. 9, 1 9 5 4
Fig. 0.1.
Variation i n Apparent Temperature
Experiment
Coefficient of R e a c t i v i t y w i t h Time.
H-4.
1
-
i
1
1
J
1
the center where the reactivity effect was greatest.
With a sudden increase i n heat extraction, a slug
of cooled fuel i n i t i a l l y entered the center of the
core and caused a large reactivity change. Since
the mean reactor temperature was the average
temperature of a l l the fuel tubes, a comparison of
the i n i t i a l rate of rod insertion by the servo with
the i n i t i a l rate of change of the mean reactor
temperature could give an apparent temperature
coefficient much larger than i t s actual value. For
example, i n experiment L-8 (cf., Figs. 5.13 and
5.14) a comparison of the slopes of the rod position
curve and the mean temperature curve near the
beginning of the run was made. If it i s assumed
that the temperature response of the thermocouples
lagged behind the response o f the regulating rod
movement, it i s interesting t o compare the points
of steepest slope for the two curves.
By using
the slope from Fig. 5.13 at time 0218:30, and the
slope from the mean temperature curve (Fig. 5.14)
a t 0220, a temperature coefficient of -1.6 x
( A k / k ) P F i s obtained, as opposed to the actual
A
fuel temperature coefficient of -9.8 x
detailed discussion of t h i s experiment i s given
i n the section on “Low-Power Measurements o f
This
Temperature Coefficients of Reactivity.”
effect was characteristic of the geometry of the
ARE but not necessarily of circulating-fuel reactors in general.
TIME L A G CONSIDERATIONS
i
A s mentioned above and i n the “Reactor Kinetics”
section of Chapter 6, one of the most consistently
noted phenomena of the ARE operation was the
time lag i n reactor temperature response during
every phase of the experiment.
These time lag
effects can be partly explained by the geometry
effect just described; and, i n this sense, the time
lag i s not a true time lag but only an apparent lag
due to the differences in location and response of
the thermocouples and the neutron detectors (and,
1
hence, regulating rod movements ). Other possible
causes for the time lags were the fuel transit time
around the system, the heat transfer phenomenon
within the reactor, the design and location of
thermocouples,
and a mass-temperature inertia
effect. These effects were a l l discussed briefly
i n the section on “Reactor Kinetics,” Chapter 6.
i
’ T h e regulating rod was controlled
mechanism which received i t s error
neutron detector (cf., App. C).
b y a flux servo
signal
from o
Whether or not the time lag was a l l or partly real
i s academic. The fact that it did give an observed
effect during many phases of the ARE operation
made i t mandatory to take the lag into account in
interpreting much of the data. The manner i n which
this was actually done was t o assume that the
response of the thermocouples lagged behind the
nuclear response of the reactor (by as much as
2 4 min for low-power operation and by about 1 min
during the high-power regime) in those experiments
i n which the equilibrium between the reactor and
i t s load was upset (i.e., rapid fuel cooling rates).
The thermocouple readings were then “moved up”
by that amount, and the new readings were compared with the appropriate nuclear instrument
observation. For those experiments i n which equilibrium prevailed, but i n which cooling was taking
place, it was only necessary t o correct for temperature readings (cf., app. K).
The temperature coefficient measurements were
probably more affected by the temperature-time
lags than any other single type o f measurement
made on the ARE, mainly because of the short
duration o f the experiments and their great dependence on time correlations (for example, correlations between regulating rod motion and mean
temperature changes). The results of temperature
coefficient measurements which contained time lag
corrections were not included i n the main body o f
the report because such corrections needed to be
discussed i n detail inappropriate to the context of
the report. These experiments are described i n
the following sections of this appendix.
SUBCRITICAL MEASUREMENT OF
TEMPERATURE COEFFICIENT
The subcritical measurement o f the temperature
coefficient (exp. E-2) was described i n Chapter 4.
Briefly, the procedure followed was t o cool the
fuel by raising the heat barriers on the fuel heat
exchangers, turning on the fuel helium blowers,
and then observing the increase in multiplication
w i t h the two fission chambers and the BF3 counter.
The BF, counting rate was so low that the stat i s t i c s were poor; therefore, the fission chamber
data had to be used. The subcritical multiplication
o f the fission chambers was then subject t o the
I n this
phenomenon discussed i n Appendix E.
experiment the cooling was rapid and equilibrium
conditions were not attained.
Consequently, i n
order to find a value o f the temperature coefficient
from these measurements, i t was necessary t o
193
f
eB
k
apply three corrections: a correction for the fission
chamber multiplication error, a time lag correction,
and a temperature correction of control room observed temperatures (app. K).
Throughout the subcritical experiments ample
data were taken simultaneously on the BF, counter
and the two fission chambers for a correlation p l o t
between the counting rates of the BF, counter and
the fission chamber to be easily obtained, as
shown i n Fig. 0.2. A plot o f the raw data obtained
from the fission chambers before corrections were
applied i s shown i n Fig. 0.3,2 which shows the
counting rates of the chambers plotted as a function
o f the reactor mean temperature. The hysteresis
effect i s the result of the time lag. The progress
o f the experiment can be read from the curves by
starting on the right side at 1004 and proceeding
counterclockwise around the loops.
The fuel
blower was turned on at 1004 and allowed to cool
t h e fuel for 5 min, after which the blower was turned
off and the system then slowly returned t o i t s i n i t i a l
condition.
The f i s s i o n chamber counting rates
increased while the blower was on and decreased
after the blower was turned o f f again i n immediate
response t o the cooling. The reactor mean temperature change, on the other hand, lagged behind
both when the blower was turned on and when i t
was turned off. The blower was turned off shortly
after 1009, but even though the counting rate
started to decrease immediately, the reactor mean
temperature continued to fa1 I for approximately
2Y2 min before it began t o show a warming trend.
Undoubtedly the fuel temperature did actually
follow closely the changes introduced by the
blower (otherwise the counting rate changes would
not have been observed as promptly as they were),
but because of the various effects noted in t h i s
appendix and i n the "Reactor Kinetics" section
of Chapter 4 the thermocouples were slow i n
responding. If the thermocouples had shown instant
response there would have been no hysteresis
effect observed, and the p l o t of k vs mean temperature would have been a straight l i n e with a negat i v e slope proportional to the temperature coefficient.
The p l o t of k v s mean temperature was obtained
by applying the three corrections noted above i n
2 T h i s plot i s a c t u a l l y a cross plot of the curves o f
Fig. 4.5, which show both the counting rate and reoctor
mean temperature plotted as a function of time.
194
the following way. The f i r s t correction was applied
t
by changing the fission chamber data to BF,
counter data by using the curves of Fig. 0.2. The
resulting points obtained from each fission chamber
were averaged and then plotted on a time scale
along with the reactor mean temperature t o produce
a p l o t similar to the curves of Fig. 4.5. The maximum of the counting rate curve and the minimum
of the temperature curve were then matched up (the
1
temperature curve was effectively moved up 2/2
min i n time), and new temperatures were read from
the temperature curve corresponding to the time
that the counts were taken.
From the counting
(1/M) was
rates the multiplication factor k = 1
determined for each point, and then a p l o t of k as
a function of mean temperature was drawn up, as
shown i n Fig. 0.4. A straight l i n e could reasonably
be drawn through the points.
The third and final correction to be applied to
the mean temperature was obtained from Appendix
K. Since t h i s experiment was one i n which the
fuel was cooled rapidly and equilibrium conditions
were not met, the curves of Fig. K.2 are applicable.
From Fig. K.2 it can be shown that a change of
1°F in the true mean temperature corresponds to
a change of 1.40"F i n the mean temperature read
from the control room instruments. Thus the mean
temperature change observed had to be increased
by a factor of 1.4. A measurement of the slope o f
the curve of Fig. 0 . 4 gives a ( A k / k ) / A T of 1.65 x
The average k over the p l o t i s 0.922. By
applying the factor 1.4 to the observed mean temperature change, the fuel temperature coefficient
was calculated t o be
-
a= (Ak/k)/AT =
-1.65 10-4
0.922 x 1.4
= -1.28
lo-*
=
-(1.28 5 0.24)
x
L,
L
t
L
I
f
.
T h i s value i s about 30% higher than that given
by the results of experiment H-5, but i t i s i n fair
agreement i n consideration of a l l the necessary
corrections. A consideration o f the errors involved
showed that the maximum error was of the order of
Therefore,
magnitude o f 2.4 x lo-?
a
L
.
If the lower l i m i t of t h i s value i s taken, the
agreement between t h i s value and the accepted
value i s fairly good. T h i s experiment did not y i e l d
an over-a I I temperature coefficient.
Ei
t
C
L
. !
4
E
o4
9"
5
i
2
:
4 o3
a
5
-"
j'
P
+
c
3
5 2
W
t
(L
z
53
8
40'
W
(L
m
$
I
5
0
z
0
v)
v,
U.
2
5
2
i
,
4
2
5
40
BF,
Fig,
0.2.
2
5
4 0'
2
5
CHAMBER COUNTING RATE (countslsec)
Correlation Between the Counting Rates of the BF, Counter and the Two F i s s i o n Chambers.
ORNL-LR-
DWG 38848
8OC
550
t
\,
790
4007
i
540
I
L
78C
530
770
520
I
L
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0)
0
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+
c
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0
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5
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760
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500
750
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LL
I?
LI
$
-
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740
490
730
400
720
470
.
*
i
740
700
4240
4250
4260
4270
4280
4290
T
4300
460
450
4340
REACTOR M E A N T E M P E R A T U R E (OF)
Subcritical Measurement of Temperature Coefficient of R e a c t i v i t y (Uncorrected F i s s i o n
Fig. 0.3.
Chamber Data).
196
L
t
ORNL-LR-DWG
0 927
65
0 926
The low-power measurements o f the temperature
coefficients of reactivity were similar t o the subc r i t i c a l measurements, except that w i t h the reactor
c r i t i c a l and on servo a t 1-w power, r e a c t i v i t y
introduced by cooling the fuel was observed by a
change in the regulating rod position. The experiment i s described i n Chapter 5.
0 925
0 924
A s shown i n Fig. 0.5, which i s a p l o t of the
regulating rod position vs the observed mean
temperature during the experiment, a hysteresis
phenomenon was obtained. It i s significant that
no time lags needed to be taken into account i n
the interpretation of this data, because the experiment proceeded slowly enough for equilibrium
conditions to prevail. However, since t h i s was a
cooling experiment, a temperature correction had
t o be applied. Cooling took place along the lower
half of the figure and heating occurred along the
upper portion. Each of the curves had an i n i t i a l
steep slope corresponding t o an i n i t i a l fuel temperature coefficient and a l e s s steep slope from which
an over-all r e a c t i v i t y coefficient was found.
0 923
k
0922
0 921
0 920
0 919
I
0 918
I
09f7
1230
LOW-POWER MEASUREMENTS O F
TEMPERATURE COEFFICIENTS
O F REACTIVITY
After correction for the mean temperature read1240
I250
1260
I270
I280
1290
1300
REACTOR MEAN TEMPERATURE (OF)
1
I
J
R e a c t i v i t y as a Function of Reactor
Fig. 0.4.
Mean Temperature as Determined from the Subc r i t i c a l Temperature Coefficient Measurement. Experiment E-2.
ings, the average of the two i n i t i a l slopes gave
a fuel temperature coefficient of reactivity of
-9.9 x lo-', and the other slopes gave an average
over-all reactivity coefficient of about -5.8 x
These values agree well w i t h the accepted values
and -6.1 x lo-' for the fuel and
of -9.8 x lo"
over-a1 I temperature coefficients o f reactivity,
respectively.
197
P
4
ORNL-LR-DWG
25
6573
1
EFFECT
h
20
1
REACTOR BEING COOLED
D REACTOR HEATIF
-._
C
-z
45
0
c
cn
a
a
0
0-
0
E
W
zt-
Q
/I
_J
3
W
E
40
7
J'
/
7
T E M P E R ATUR E COE F F I
FOR CURVE 4 , - 9 4 6 )
FOR CURVE 2, -5.70 x
5
A V E R A G E CURVES i AND 3,-9.9
AVERAGE CURVES 2 AND4,-5.75
~
x ~ O - ~
0
1230
4240
Fig. 0.5. Regulating
Experiment L-5.
198
1250
1260
i2 7 0
4280
4290
REACTOR M E A N F U E L TEMPERATURE ( O F )
Rod Position a s
ai *unction
1300
13.10
1320
I
of Observed Reactor Mean Temperature During
r
i
f
i
HIGH-POWER M E A S U R E M E N T S OF
TEMPERATURE COEFFICIENTS
OF REACTIVITY
i
1
t
I
1
i
4
.
During the high-power operations an experiment
was conducted at a power of 100 k w which was
similar to the low-power experiment just described.
The fuel helium blower was turned on w i t h the
reactor on servo, and the fuel was cooled. However, the action took place so rapidly that a l - m i n
time lag3 of the fuel mean temperature had to be
accounted for in plotting the data i n addition to
the temperature correction. Figure 0.6, i n which
the regulating rod movement i s plotted as a function
o f the mean fuel temperature, shows the experimental measurements. Two d i s t i n c t slopes were
observed that corresponded to an i n i t i a l fuel
temperature coefficient and t o an over-a1 I temperature coefficient of reactivity.
From the steep slope (curve No. 1) a fuel temper(AK/k)/OF was
ature coefficient of -1.17 x
obtained, and from the other slope an over-all coef( A k / k ) P F was found. These
f i c i e n t of -5.9 x
two values are i n fair agreement w i t h the accepted
and -6.1 x
for these
values of -9.8 x
coefficients.
high power operation were observed to
be shorter than those at low or no power. F o r a discussion, see Chapter 4.
'Time
l a g s at
__
20
ORNL-LR-DWG-6574
I
1
,
i
I
(5
i
-
It
1
-2
z
0
L
<
t
u)
0
n 10
$
0
z
i
4
J
0
n
5
ATURE
1.17
FOR CURVE 2,
5.9
1
0
1240
iNT
FOR CURVE I,
i
I
1250
1260
I270
1280
1290
1300
1310
1320
REACTOR MEAN FUEL TEMPERATURE 1°F)
F i g . 0.6. Regulating Rod P o s i t i o n a s a Function of Reactor Mean F u e l Temperature for Experiment H-4.
199
P
Appendix P
THEORETICAL XENON POISONING
The xenon poisoning which existed i n the ARE
was determined experimentally to be an almost
negligible amount, It was therefore o f interest to
compute the amount o f such poisoning that would
have been present i f no xenon had been l o s t due
to off-gassing o f the fuel so that a measure o f the
effectiveness o f the off-gassing process could be
obtained.
The poisoning by Xe135, when the xenon content
has reached equilibrium, i s given by'
4
P o = Q2(Yl
(1)
+
I;(
increasing neutron energy,
R. R. Bate, R. R.
Coveyou, and R. W. Osborn investigated the neutron
energy distribution in an absorbing i n f i n i t e moderator b y using the Monte Carlo method and the
Oracle. B y assuming a constant scattering cross
section and a constant l / v absorption cross
section, they found that the neutron energy d i s t r i bution i s well represented by a Maxwellion d i s t r i bution corresponding t o an effective temperature,
T e , provided
12
y
y 2 ) $0
K
=
I
-
P o = the ratio o f the number o f thermal
neutrons adsorbed i n the xenon to
those adsorbed i n the fuel,
c2 = microscopic xenon cross section for
t hermal -neutron absorption,
(y,
+ yz)
= the total fractional yield o f Xe13'
from fission, both from iodine decay
and direct Xe formation = 0.059,
A,
= the
decay
constant
for
Xe135 =
2.1 x I ~ - ~ / s e c ,
XJru
= the r a t i o of the macroscopic thermal-
neutron cross section for fission t o
that for absorption, for U235 Z f / c u =
0.84,
+o
= the average thermal-neutron flux i n
the fuel (since the entire fuel volume, 5.33 ft3, i s equally exposed to
t h i s flux, although there i s only
1.37 ft3 o f fuel i n the core a t one
time, the average flux in the fuel i s
1.37h.33 or 0.26 times the average
flux i n the reactor, which i s 0.7 x
10'~).
The Xe135 absorption cross section, c2, i s
smaller i n the ARE than a t room temperature, because the average neutron energy i n the ARE
exceeds the average neutron energy corresponding
to room temperature and because the Xe135 absorption cross section drops o f f rapidly w i t h
'S. Glasstone and M. C. Edlund, The E l e m e n f s o/
Nuclear Reactor Theory, D. V a n Nostrand Co., Inc.,
N e w York (1952). p 333, 11.57.2.
200
i
k
.
!
t
a
2
&a
-<
(A2 + ff2950)
where
-
,
0.06
P
2s
where the mocroscopi c absorption cross section
Za i s measured a t the moderator temperature and
i s the macroscopic scattering cross section.
The effective temperature
Zs
Te
=
+
T,(1
UAK)
5.5 x
Beryl I i um
5.5 x
"235
7.8
k
I
,
where T m i s the moderator temperature, a i s a
constant approximately equal to 0.9, and A i s the
atomic weight o f the moderator.
The present state of the theory does not permit
consideration of the inhomogeneous distribution o f
the various constituents o f the ARE core, and
therefore the main constituents were considered to
be evenly distributed over the core, The nuclei
per cubic centimeter were thus
Oxygen
L
t
-
i
1019
Other elements made o n l y a negligible contribution
to the cross section o f the core. The following
cross sections were used:
Oxygen, scattering
Beryllium, scattering
Uranium, absorption
4
7
360
barns
barns
barns2
20b+ained b y converting the room temperature value
of the cross section to the value a t the reactor operating
temperature b y multiplying b y the square root o f the
ratio o f the temperatures:
630 (barns) x
1033
= 360
(OK)
(barns)
.
e
8
Thus
360
x
7.8 x 1019
=
+ 7)
d
x
0,038
5.55 x
TABLE
P.1.
Xe135 ABSORPTION CROSS SECTION
IN T H E REACTOR
E i (ev)
f
n(EJ
hE
.
=
=
1033[1 + (0.9 x 12.5
1474OK
= 220OOF
i
4
#
1
The Xe13' absorption cross section i n the reactor
was then determined, as shown i n Table P.1,
which gives the energy intervals o f the neutrons,
Ei, the fraction o f neutrons i n these energy intervals according t o the Maxwell-Boltzman d i s t r i bution, n ( E i ) / n , and the total Xe13' cross section,
which the
~
ut, for each energy i n t e ~ a l , from
average Xe13' cross section i s obtained, In the
energy range i n question the xenon adsorption
cross section i s approximately equal to the total
A s shown i n Table P.l,
xenon cross section,
the value o f t h i s cross section i s 1.335 x
barns or 1.335 x
cm2.
The anticipated xenon poisoning during the 25-hr
xenon run (exp. H-11) may then be computed from
Eq. 1 by using the value determined above for t h e
Xe13' absorption cross section:
i
I
I
(Ei)
(barns)
0.060
2.50 x
0.04
0,073
2.75 x
0.06
0.076
3.25 x
0.08
0.075
3.30 x
0.10
0.072
2.82 x
0.12
0.067
1.92 x
0.14
0,062
1.27 x
0.16
0.057
0.75
0.18
0.051
0.45 x
0.20
0.046
0.32 x
.
a;.
f
t
0.02
0.038)l
x
(7
n
For the atomic weight, A , the average of the
values for beryllium and oxygen, 12.5, was assumed. Thus
Te
=
X
f x o t x e(Ei)
(barns)
lo6
lo6
lo6
lo6
lo6
lo6
lo6
lo6
lo6
lo6
Average OLXe
0.079 x
lo6
lo6
lo6
lo6
lo6
lo6
lo6
0.042
lo6
0.150 x
0.200 x
0.248 x
0,248 x
0.201 x
0.129 x
X
0.023 x lo6
0.015 x
lo6
1.335
lo6
X
(such as lnconel, etc.), and during the 25 hr of
operation, i f the xenon had a l l stayed i n the fuel
it would have reached 69% of i t s equilibrium
c ~ n c e n t r a t i o n . ~ By applying these various corrections, it i s found that the xenon poisoning i n
the ARE at the end of the 25-hr run should have
lo6
4
9
1.3
P0 =
(2.1 x lo")
B
-
0.0117
(2.1
1
3
d
1
4
x
x
+
x
0.24)
0.059
x
0.7 x 1013 x 0.26
+ (1.3 x
lo"
=-
x
0.01 17
10-5
T h i s value has to be corrected because about
one-third of the fissions occur a t energies above
thermal and therefore have only l i t t l e competition
from xenon absorption. Furthermore, the r e a c t i v i t y
loss due to poison was about 89% o f that computed
above because o f the absorption i n other poisons
x
0.84
0.7 x 1013 x 0.26)
= 0.005 ,
2-34
been 0.2% i n
gassed.
hk/k if
no xenon had been off-
BNL -1 70.
4Glosstone and
Edlund,
op. cit., p
333.
f
I
4
20 1
Appendix Q
b
OPERATIONAL DIFFICULTIES
The operational difficulties described here are
only those which occurred during the nuclear phase
of the operation, that is, from October 30 t o November 12, the period of time covered by t h i s
report. With a system as large, as complex, and
as unique as that which constituted the ARE, it i s
amazing that so few difficulties developed during
the crucial stages o f the operation. Furthermore,
such troubles were, without exception, not of a
serious nature. T h i s i s in large measure attributable t o the long period o f installation and testing'
which preceded the nuclear operation, the safety
features inherent in the system, and the quality
o f workmanship which went i n t o i t s construction.
All major difficulties and impediments which arose
are discussed below and are grouped by systems,
with special regard for chronology o f occurrence.
double tracing of the l i n e with calrod heaters
staggered so that successive pairs of calrods did
not meet a t the same point. Even with t h i s arrangement, thermocouples were necessary a t every
heater junction and each calrod or each pair o f
calrods should have had a separate control.
i
After it became an established part o f the enrichment procedure t o continuously bleed gas through
the transfer l i n e into the pump ( i n order t o keep
the l i n e clear), the gas was vented through an
extra l i n e a t the pump. The extra l i n e was not
properly heated and soon plugged with vapor
condensate. It then became necessary t o use the
primary pump vent system which had a vapor trap.
B y reducing the bleed gas flow to a minimum, t h i s
vent system could be used without becoming
plugged with vapor condensate.
t
ENRICHMENT S Y S T E M
A s mentioned previously,
the fuel enrichment
system was changed (shortly before the c r i t i c a l
experiment was t o begin) from a remotely operated
two-stage system, i n which the transfer wos to
start with a l l the fuel concentrate i n a single container, t o a manually operated two-stage system,
i n which small batches o f the available concentrate
were transferred, one a t a time. A portion o f the
Although
equipment used i s shown i n Fig. Q.l.
t h i s change resulted i n an improvement both i n
safety and control, the temperature control o f the
manually operated system was persistently difficult. Furthermore, i n order t o avoid plugged lines
because o f the concentrate freezing a t cold spots,
the lines had t o be continuously purged with gas
and the e x i t gas lines then plugged as a result o f
concentrate-vapor condensation. Both these difficulties could have been avoided with proper design.
The temperature control was a greater problem
here than anywhere else i n the system because o f
and
in.) tubing used i n the transthe small
fer lines and an inferior technique o f heater installation. T h i s problem was aggravated by the virtual
inaccessibility o f the connection between the
transfer l i n e and the pump at the time the system
was revised. The final heater arrangement used,
which proved to be satisfactory, consisted o f
(t6
'Design and Installntion of the Aircraft Reactor E x periment, ORNL-1844 (to be issued).
202
The transfer l i n e t o the pump served adequately
throughout the c r i t i c a l experiment, although it had
t o be reworked four times either because of the
formation o f plugs or the development o f leaks ( i n
Swagelok connections where the l i n e was cut and
replaced). During the l a s t injection for rod c a l i bration the transfer l i n e again became plugged a t
the f i t t i n g through the pump flange. T h i s f i t t i n g
was a r e s i stance-heated concentric-tubing arrangement which provided an entrance for the 1300OF
transfer l i n e through the 7OOOF pump flange. The
oxidized fuel from previous leaks had shorted the
heating circuit, and the f i t t i n g had therefore cooled
and plugged. Attempts t o clear the f i t t i n g caused
it t o leak; the leak was sealed but the f i t t i n g was
then inoperable.
The final injection o f concentrate, which was
required for burnup at power, xenon poison, etc.,
was therefore made through the fuel sampling line.
A special batch o f concentrate was prepared and
pressurized into the pump through the sample line.
T h i s technique worked satisfactorily but had previously been avoided because the l i n e had not been
designed to attain the high temperatures > 120OOF
required by the melting point o f the concentrate.
Furthermore, the sample l i n e was attached t o the
pump below the liquid level in the pump, and i f
a leak had developed i n the l i n e a sizeable s p i l l
would have resulted.
b
I
{
1
.t
c
e
a
t
F
b
I
?
,
VI
al
.-c
A
f
0
C
VI
Y
P
0
c
m
b
ti
P
Y
u)
U
C
E
i:
x
u)
c
al
4-
Y,
C
E
al
5
c
L
.-u
W
.-
.-c;,
u
LL
203
i
PROCESS I N S T R U M E N T A T I O N
For an appreciation o f the generally excellent
performance of the instrumentation, know1edge o f
the number and complexity of the instruments i s
required. There were at least 27 strip recorders
(mostly multipoint), 5 circular recorders, 7 indicating controllers, 9 temperature indicators (with
from 48 to 96 points apiece), about 50 spark plugs,
20 ammeters, 40 pressure gages, 16 pressure regulators, 20 pressure transmitters, and numerous flow
recorder indicators and alarms, voltmeters, tachometers, and assorted miscellaneous instruments.
Furthermore, many of these were employed i n systems circulating fuel and sodium at temperatures
up to 1600OF. Since a l l instruments were subject
to routine inspection and service, petty difficulties
were kept to a minimum.
I n the course o f nuclear operation, only the
following instruments gave cause for particular
concern:
fuel flowmeter, main fuel pump level
indicator, several sodium system spark plugs, fuel
pressure transmitter, and several pump tachometer
generators. Of these only the tachometer generator
“failures”
were not caused by the materials and
temperatures being instrumented. The tachometers,
as installed, were belt driven rather than directcoupled and were not designed to withstand the
side bearing loads to which they were subiected.
The tachometers were replaced, however, before
the p i t s were sealed, and they performed satisfactori l y during the high-power operation.
The fuel flowmeter and the fuel pump level indicator were similar instruments i n which a float or
bob was attached to a long tapered iron core suspended in a “dead leg.” C o i l s were mounted outside the dead leg which located the position of the
core. The position o f the core could be interpreted
as a measure either of fuel level or of fuel flow up
past the bob. The c o i l current, however, was very
sensitive to the fuel temperature, which had to be
maintained above the fuel melting point. In addition to the temperature sensitivity, several c o i l s
(spare c o i l s were provided on each instrument)
opened up during the experiment, presumably due
to oxidation o f the coil-to-lead wire connection.
The fuel flowmeter oscillated rapidly over a 10gpm range throughout the later stages o f the experiment, although the electronics of the instrument
appeared to be i n order. On the other hand, operation of the main fuel pump level indicator was
satisfactory up to the last day o f operation, at
which time the spare c o i l opened up (the main c o i l
204
had previously opened). It i s f e l t that these instruments would have performed satisfactorily i f the
iron core were designed t o move in a trapped gas
leg above the float rather than i n a dead leg below
the float so that the operating temperature could
be reduced. Furthermore, the c o i l reading should
be “balanced out” to eliminate the temperature
sen s it iv ity
Of the numerous spark plugs which were employed
i n the various fuel and sodium tanks as the measure
or check on the l i q u i d level, only three o f those i n
the sodium system showed a persistent tendency
to short. These shorts could not always be cleared
by “short-burning,”
but they frequently cleared
themselves as the l i q u i d level dropped. Although
the probes were located i n a riser above the tank
top t o minimize shorts, the shorts could probably
have been eliminated by using larger clearances
than were afforded by the use of &-in.-OD probes in
a \-in.-IPS pipe riser.
.
The high-temperature fuel system pressure transmitters suffered a zero s h i f t during the course of
the experiment.
These transmitters employed
bellows through which the l i q u i d pressure was
transmitted t o gas. It i s probable that the bellows
were distorted a t times when the gas and l i q u i d
pressure were not balanced as a result of aperational errors or plugged gas supply lines. I n any
case, the gas ports i n the transmitter occasionally
plugged, and a zero shift i n the instrument was
observed.
In view of the large number of thermocouples i n
use throughout the experiment ( i n the neighborhood
of lOOO), i t i s not surprising that a small number
were in error. However, those which gave incorrect
indications included some o f the most important
ones associated with the entire experiment. A s
discussed i n Appendix K there was serious disagreement between line thermocouples inside and
outside the reactor thermal shield, although it
would appear that they both should have given the
same indication.
Other misleading temperature
readings were obtained from the thermocouples on
the tubes i n the fuel-to-helium heat exchangers.
These thermocouples were not properly shielded
from the helium flow and read low.
Although most o f the thermocouple installations
were designed t o measure equilibrium temperatures
and did so satisfactorily for a number o f experiments involving fast transients, it was important
that the response o f the thermocouples be > 10°F
per sec i n order to correlate the changes i n fluid
h
i
k
L.
L
i
C
L
g:
b
h
F
i
p.
.
.1
t
1
i
temperatures with nuclear changes. Unfortunately
it i s not certain that t h i s was the case, and, in
addition, i n certain instances, as with one of the
thermocouples on each o f the reactor AT and reactor mean temperature instruments, the thermocouples were mounted on electrical insulators
which increased the thermal lags.
The above discussion covers most o f the instrumentation d i f f i c u l t i e s that arose.
T h i s i s not
meant t o imply, for example, that all the 800-odd
thermocouples lasted throughout the experiment;
there were open thermocouples scattered throughout the system. Furthermore, the instrument mechanics were kept busy; when not doing installation
work, they were usually involved i n routine service
and maintenance work.
N U C L E A R I N S T R U M E N T A T I O N AND C O N T R O L S
1
i
i
1
r
5
It i s d i f f i c u l f i n the case of the nuclear instruments t o separate operation problems from those
inherent i n routine installation and debugging,
since these latter operations were continued right
up to the time the instruments were needed. However, during actual operation, a l l nuclear instrumentation performed satisfactorily.
The control
mechanism operated as designed, except that one
shim rod had a higher hold current (it was supported by an electromagnet) than the other two.
T h i s situation was improved by cleaning the magnet
face and filtering the gas surrounding the magqet.
1
ANNUNCIATORS
i
t
i
The control system included an annunciator
panel which anticipated potential troubles and
indicated off-design conditions by a l i g h t and an
alarm. During the course o f the experiment, certain annunciators consistently gave false indications, and therefore the bells (but not the lights)
o f these annunciators were finally disconnected.
Included were the standby sodium pump lubrication
system flow, the standby fuel pump cooling water
flow, the fuel heat exchanger water flow, and the
fuel heat exchanger low-temperature alarm. The
f i r s t three annunciators had mercury switches
which were either improperly mounted or set too
close to the design condition, but the fuel heat
exchanger low temperature alarm error was due t o
faulty thermocouple indication; that is, instead of
indicating fuel temperature, the thermocouple,
which was located i n the cooling gas stream, read
low.
H E A T E R S A N D HEATER CONTROLS
During the time the system was being heated, the
heater system power was over 500 kw. T h i s heat
was transferred to the piping and other components
by the assorted ceramic heaters, calrods, and strip
heaters that covered every square inch o f fuel and
sodium piping, as well as a l l system components
Although there
which contained these 1 iquids.
were numerous heater failures resulting from
mechanical abuse up until the time the p i t s were
sealed, a l l known failures were repaired before
the p i t was sealed. Only four heater circuits were
known t o be inoperable a t the time the reactor was
scrammed two heaters showed open, two shorted.
Except for the fuel enrichment system i n which
the heating situation was aggravated by the higher
temperatures as well as the small lines, the
available heat was adequate everywhere.
The
control of the various heater c i r c u i t s was, however,
i n i t i a l l y very poor; it consisted of four voltage
buses to which the various loads could be connected plus variacs for valve and instrument
heaters. However, i n order t o obtain satisfactory
heater control for a l l elements of the system, 13
additional regulators were installed i n addition t o
numerous additional variacs. With the additional
regulators it was possible to s p l i t up the heater
load t o get the proper temperatures throughout the
system without overloading any distribution panel.
Even with the helium annulus, one function of
which was t o distribute the external heat uniformly,
i t must be concluded as extremely desirable, if
not an absolute necessity, that a l l components o f
any such complex high-temperature system be
provided with independently control led heater units
in order to achieve the desired system equilibrium
temperature.
-
SYSTEM C O M P O N E N T S
The only major components of either the sodium
or fuel system which caused any concern during
the nuclear operation were the sodium valves
(several of which leaked) and the fuel pump (from
which emanated a noise originally believed to
originate i n the pump bearings). In addition, there
were problems associated with plugged gas valves
and overloaded motor relays.
The sodium valves that leaked were the two
pairs that isolated the main and standby pumps
and a t least one of the fill valves i n the lines t o
the three sodium fill tanks. The leakage across
the pump isolation valves was eliminated from
concern by maintaining the pressure in the inoperative pump at the value required to balance with
the system pressures. The leak (or leaks) i n the
f i l l valves was of the order of
to 1 ft3 of
sodium per day, and was periodically made up by
r e f i l l i n g the system from one of the f i l l tanks. In
the course of these operations, two tanks eventually
became empty, and i t was then apparent that most
of the leakage had been through the valve to the
third tank. This leakage was i n i t i a l l y abetted by
the high (-50 psi) pressure drop across the valve,
but even though the pressure difference was reduced to 5 to 10 psi the leakage rate appeared t o
increase during the run.
It was of interest, as well as fortunate for the
ultimate success of the project, that the fuel carrier
f i l l valves were tight. However, in the fuel system,
only two valves were opened during the f i l l i n g
operation and only one of these had t o seal i n
order t o prevent leakage. T h i s valve did seal, and
i t was opened only one other time, i.e., when the
system was dumped.
Each sodium pump (main and standby) and each
fuel pump (main and standby) was provided with
four microphone pickups to detect bearing noises.
Shortly after the fuel system had been f i l l e d with
the fuel carrier, the noise level detected on one
of the main fuel pump pickups jumped an order of
magnitude, while that on the other increased
substantially. A t the time the noise was believed
t o be due to a f l a t spot on a bearing, and operation
was therefore watched very closely.
When the
noise level did not increase further (in fact, i t
tended to decrease), i t was decided to continue
the experiment without replacing the pump (a very
d i f f i c u l t job which could conceivably have resulted
i n contamination of the fuel system). The pump
operated satisfactorily throughout the experiment.
Subsequent review of the pump design and behavior
of the noise level indicated that the noise probably
originated at the pump discharge where a sleeve
was welded inside the system to effect a s l i p
connection between the pump d i scharge duct from
the impeller housing and the e x i t pipe, which was
welded to the pump casing. Vibration of the sleeve
i n the s l i p joint could account for a l l the noise.
Although the vent header was heated from the
fuel and sodium systems to the vapor trap f i l l e d
with NaK (which was provided t o remove certain
fission products but also removed sodiuin vapor),
$
20 6
the temperature control of the header and the
individual vent lines connected to it was not
adequate t o maintain the l i n e above the sodium
melting point and yet not exceed the maximum
temperature l i m i t (4OOOF) of the solenoid and
diaphragm gas valves. Consequently, these vent
lines became restricted by the condensation 04
sodium vapor. It was apparent that a higher temperature valve would have been desirable, that the
gas line connecting the tank t o the header should
have been installed so that i t could drain back into
the tank, and, also, that good temperature control
of the line and valve should have been provided.
A s it was, exceptionally long times were required
t o vent the sodium tanks through the normal vent
valves. It would have been necessary t o use the
emergency vent system, which was s t i l l operable
at the end of the experiment, i f a fast dump had
been required.
In addition t o the above failures or shortcomings,
there were numerous problems of less serious
nature i n connection with auxiliary equipment,
motor overloads, water pipes which froze and split,
and air dryers which burned out. However, nothing
occurred in such a manner or at such a time as t o
have any significant bearing on the conduct of the
experiment.
LEAKS
The only sodium or fluoride leaks that occurred
have already been discussed. These included one
minor sodium leak i n the sodium purification system (discussed in chap. 3, “Prenuclear Operation”)
and two fuel concentrate leaks from the enrichment
system. That neither the reactor fuel system nor
sodium system leaked i s a tribute t o the quality
of the workmanship in both welding and inspection
that went into the fabrication of these systems. I n
all, there were over 266 welded joints exposed to
the fluoride mixture, ond over 225 welded joints
were exposed to the sodium.
In contrast to the l i q u i d systems, there were
several leaks i n gas systems. It i s f e l t that these
leaks would not have occurred i f the gas systems
had been fabricated according to the standards
used for the liquid systems. The notable gas leaks
were from the fuel pumps into the pits, from the
helium ducts i n the heat exchangers into the pits,
and from the p i t s into the building through the
vorious p i t bulkheads, as well as the chamber
which housed the reactor controls.
”
c
i
c
c
=
%
c
B
t
P
B
r:
i
The combination o f the leak out of the fuel pump
and that out of the p i t s required that the p i t s be
maintained a t subatmospheric pressures i n order
t o prevent gaseous activity from contaminating the
building. Accordingly, the p i t pressure was lowered
,
by using portable compressors
by about 6 in. HO
which discharged the gaseous activity some 1000 ft
south o f the ARE building. The activity was o f
such a low magnitude that, coupled with favorable
meteorological conditions, it was possible to
d
4
J
operate in t h i s manner for the l a s t four days o f the
experiment.
The leaks out o f the helium ducts in the fuel and
sodium heat exchangers resulted in a maximum
helium concentration in the ducts o f the order of
50%, and t o maintain even t h i s low concentration,
i t was necessary to use excessive helium supply
rates, i.e., 15 cfm to the ducts alone and another
10 cfm to the instruments.
Appendix
R
L;
INTEGRATED POWER
The total integrated power obtained from the
operation of the Aircraft Reactor Experiment does
not have any particulur significance in terms of
the operational l i f e of the system.
However, a
total integrated power of 100 Mw was more or less
arbitrarily specified as one o f the nominal objectives of the experiment. While a t the time the
experiment was concluded it was estimated that
this, as w e l l as a l l other objectives of the experiment, had been met, the estimate was based on a
crude evaluation of reactor power. Since subsequent
analyses of the data have permitted a reasonably
accurate determination of the reactor power, it is
of interest to reappraise the estimated value of the
total integrated power.
The total integrated power could be determined
from either the nuclear power or the extracted
power (cf., section on "Reactor Kinetics" in
chap. 6).
The power curves, which should have
equivalent integrals, could be obtained from any
o f a number of continuously recording instruments,
i.e., the nuclear power from either the micromicroammeter or log N meter, and the extracted power
from any of the several temperature differential
recorders i n the fuel and sodium circuits.
The
total integrated power has been determined both
from integration of a nuclear power curve (log N )
and from the sum of the extracted power i n both
the sodium and fuel circuits, as determined by the
temperature differentials i n each system together
w i t h their respective flows (which were held
constant). For both power determinations a calculation of the associated error was made.
EXTRACTED POWER
T h e integrated extracted power was determined
from the charts which continuously recorded the
temperature differential ( A T ) i n each system. As
a matter of convenience the sodium system AT
was taken from a 24-hr circular chart, while the
fuel AT was taken from one of the s i x Brown s t r i p
charts which recorded, in the control room, the
AT across each o f the s i x parallel fuel circuits
through the reactor.
The individual c i r c u i t
recorders had a much slower chart speed than
that of the over-all AT recorder and were therefore
much easier to read,
T o keep the results w e l l
w i t h i n the accuracy o f the whole experiment, tube
No. 4 was selected for the determination because
208
t
the chart trace very closely corresponded t o that
of the over-all fuel AT across the reactor. From
Fig. K.3, Appendix K, the control-room-recorded
AT was converted t o what was accepted as the
correct AT. The power extraction was then calculated and plotted against time in Fig. R.l.
t
The sodium AT across the reactor was obtained
from the circular charts of the recorders located
in the control room.
These AT'S were a l s o
corrected by using Fig, K.6, Appendix K, and
the plot of the power extracted by the sodium as a
function of time i s a l s o shown in Fig. R.l on the
same abscissa as that of the fuel plot. The total
bf
t
integrated (extracted) power, i.e.,
the sum of the
area under both the fuel and sodium system power
curves, was then determined by using a planimeter;
it was found t o be 97 Mw-hr.
A calculation was also made of the magnitude
o f the "maximum" possible error i n the determination of the reactor power. The power equation,
which was calculated for both the f u e l and sodium
systems, was
P = kfAT,
where
P =
k
reactor power,
= a constant containing heat capacity and
conversion units,
f = flow rate of fuel or sodium,
AT = temperature difference across reactor.
The consequent error equation i s
AP
=
kf A ( A T )
+
k AT A/
+
f AT Ak
,
where
Ak = maximum error in the heat capacity,
Af = maximum error i n the flow rate,
A ( A T ) = maximum error i n the temperature
difference.
Nominal average values of these factors for the
fuel system were
f = 46
AT
k
A/
A(AT)
Ak
L
I
gpm
= 350OF
= 0.11 kw/day gpm
= 2 g p m 2' 5%
= 10°F
= 0.011
2' 3%
2' 10%
Therefore
,.
b
i
k
AP (for fuel system) = 20%
,
2.4
-
2.0
3
I
5
46
2
0
p
12
0
LT
5
08
0.4
0
4
NOVEMBER 8
NOVEMBER 9
i
NOVEMBER40
i
f
d
NOVEMBER I4
NOVEMBER !I
NOVEMBER 12
Fig.
R.1.
Power-Time Curve.
by adding up a l l the incremental areas which were
i n terms of average log N units times time,
From the 25-hr constant-power xenon run a
relation between the log N reading and the
extracted power was obtained, It was found that
22.5 log N units = 2.12 Mw, or 10.6 log N units = 1
Mw; therefore
The values for the sodium system were
f
= 152 gpm
A T = 50°F
k = 0.0343 kw/day-gpm
Af = 4 gpm 2' 3%
A(AT) = 10°F
Ak = 0
2 20%
log N
Therefore
A P (for sodium system)
The weighted,
theref ore
over-all
=
10.6
11%
percentage
error
was
where Pf and P, are the power extracted i n the
fuel and sodium systems. The errors made i n the
power integration by using the A T chart are only
those associated with the determination of power
extraction. Since Fig. R.l i s a reproduction of the
A T trace and i t was integrated by using the
planimeter, any integrating errors were assumed
t o have been averaged out. Therefore, the error
that applies to the integrated extracted power i s
the 17% that i s applicable t o the power level
determination; therefore the integrated extracted
power was
97 f 16.5 Mw-hr
.
NUCLEAR POWER
During most of the time the reactor was operating
a t any appreciable power level, the nuclear power
level was kept fairly constant and was recorded i n
the log book. During the times the power level
was either not constant or not recorded i n the log
book, the nuclear power was determined from the
log N recorder chart by integrating the area under
the power trace.
(The log N trace was used
rather than the micromicroammeter because no
record was kept of the micromicroammeter shunt
value as a multiplying factor,)
Since the log N
chart gave a log of power vs time plot, exact
integration under the curve would have been an
extremely d i f f i c u l t task; accordingly, the area
under the curve was integrated graphically. The
curved portions of the trace were approximated by
straight lines, and an average log N value was
determined for each line segment, It was assumed
that there were as many positive as negative
errors i n this method and that the errors cancelled
out. The total integrated power was then obtained
210
t
f
t
7
;
b
k
x time (hr) = Mw-hr
for any segment under the curve. By using this
correlation between actual power and log N
reading, the total integrated nuclear power was
calculated to be 96.6 Mw-hr. This value of the
integrated nuclear power, Q, as determined from
the log N chart, was, i n effect, found from the
following relation
Q = MCAt
t
6.
B
r
c
,i
t
,
I
where
average log N recorded value over an
increment of time At,
At = increment of time i n hours,
C = a constant which converts the log N value
to Mw-hr.
The constant, C, i n the expression i s equal to the
percentage error i n the extracted power level
determination, which was
calculated i n the
preceding section t o be 17%. Therefore the error
i n C i s (0.17) (1/10.6) = 0.016. The 74-hr period
of high-power operation was divided into two
parts, one part being the 25-hr period of constant
power during the xenon run and the remainder
being the 49-hr period of variable power.
The
errors were of different magnitude for each part,
since during the xenon run there was no error i n
time or i n the determination of the average value
of M.
For the 25-hr xenon run the integrated power was
53 Mw-hr. Thus
M
F
3
M = 22.5 log N units
At
C
AM
A(At)
=
=
=
=
AC =
AQ, = MC A(At)
+ M A C At + AM C At = 9 Mw-hr
c
f
&f
t
6
L
i
L
i
e
h
25 hr
0.094 Mw/Iog N
0
0
0.016 2 17%
Therefore
k
k
B
.
For the balance of the operating time (49 hr) and
integrated power (43.6 Mw-hr), the average M must
be determined
-
I
A(&)
AC
as
Q
43.6
Ct
(0.094) (49)
M = -=
=
9.46
and for AQ2
-
1
M = 9.46 log N units
v
49 hr
C = 0.094 Mw/log N
AM = 1
At
=
=
=
AQ,=MC A(At) + M A C At + AM C At
The total error was therefore
AQ
=
12.8
+
9.0
!
1 hr
0.016 2 17%
=
i
=
12.8Mw-hr.
21.8 Mw-hr or 22.5%
i
.
tI
The total extracted nuclear power then was
96.6 2 21.8 Mw-hr
.
i
i
k
A
21 1
Appendix 5
k
b
INTERPRETATION OF OBSERVED REACTOR PERIODS DURlNG TRANSIENTS
When excess reactivity is being introduced into
a reactor at a given rate, the period meter w i l l
show a period which is not a true period but a
combination of the true period and its time rate of
change. I f the time rate of change of the period i s
known, then the true period may be found by the
following means.
First, i t i s assumed that whenever any excess
reuctivity is introduced into the reactor the power
w i l l rise according to the following equation:
and
p.
P0
r=
(”-
dA).
dt
’
~.
.
L
dt
I n experiment H-8 (cf., chap. 6) it was noted
that when the regulating rod was withdrawn the
period observed was not a true period and that it
kept changing over the 35 sec it took the regulating
rod t o travel i t s ful l distance. For a calculation
o f the true period, the following values were
taken from Fig. 6.10 and Table 6.4:
I
where
I’,
--
x
7
=
50 sec (final period observed),
1’
=
T,
72
reactor period,
time power i s increasing.
=
=
1,’7,
=
I
i n i t i a l power,
2.34 Mw,
3.94 Mw,
25 sec ( i n i t i a l period observed),
Po
C
t = 35 sec.
k
f
From these values, it was found that
Then, t o introduce a time rote of change of the
period, the power i s expressed i n the form of an
infinite series
dP
-=
dt
0.0457 Mw/sec
and
If the higher order terms are neglected, the rate
of change of power i s
ill’
- =
rlt
and therefore
Po ( A
___
dX
t - dt
=
35 [ 5 0 35 25)
=
.
-0.02
Therefore
i
f$)
,
%
r=
2.34
0.0457 + 0.02
=
36 sec
.
T h i s calculated value for the true period corresponds fairly closely t o the period of 42 sec
measured from the slope of the l o g N recorder
trace.
P
212
r
I:
Appendix
i
T
NUCLEARLOG
The following information was copied verbatim from the nuclear log book,
The only material omitted
was the running uranium inventory, which was kept during the c r i t i c a l experiment, and several calculations and graphs, which were inserted in the log book in an attempt t o interpret the data,
All
the
omitted information i s presented i n much better form elsewhere in t h i s report.
The reactor power data referred t o i n the log are not consistent, because the power was estimated
f i r s t from a calculation of flux a t the chambers, subsequently from the activation o f a fuel sample, and
finally from the extracted power.
estimate was a factor of
If it i s assumed that the extracted power value i s correct, the original
2.7 low and that calculated from the fuel activation was a factor o f 2 low. In
the log book, a l l power levels mentioned through experiment
from experiment
L-8
were based on the original estimate;
L-9 t o H-14, a l l power levels were based on the fuel activation; and it was not until
after high-power runs
1
and
2 o f experiment H-14 that correct power levels were listed.
t
P
i
213
EXPERIMENT E-1
Objective:
Run
October 30, 1954
Bring t h e R e a c t o r C r i t i c a l
Time
1
1425
1500
1507
Zero fuel flow.
Main fuel pump started to bleed down t o minimum prime level.
Main fuel pump a t minimum prime level. System volume i s now
1539
1545
4.82 ft3 (calculated).
Started adding l o t from can 6.
F i r s t 5-lb lot going into pump.
It was noticed that pips on the fission chamber occurred i n about 2-min
4.64
2
1554
1558
1603
1610
1614
1623
-
-
1625
1720
1730
+ 0.18
=
intervals (pump rprn, 520).
Second 5-lb batch going in.
Third 5-lb batch going in. Interval is exactly 2 min. T h i s gives a flow o f
18.8 gpm for a pump speed of 520 rpm.
Fourth 5-lb batch going in.
Fifth 5-lb batch going in.
L a s t of f i r s t 30-lb l o t going in.
Speeded up pump. Roughly 26 counts/sec on fission chamber 1 and 40
counts/sec on fission chamber 2. 500 rpm o n pump.
Pump speed 1100 rpm.
Sample 6 taken from pump for analysis.
Shim rods up,
control rod down,
fission chambers down.
i
fission chamber
f i s s i o n chamber
1
2
26.74 counts/sec
4 1.86 counts/sec
8,
6.93 courris/sec
BF,
1750
Shim rods down,
control rod down,
fission chambers down.
1812
1827
1
2
BF,
f i s s i o n chamber 1
f i s s i o n chamber 2
27.1 counts/sec
42 .O c ou nts/s ec
6.97 counts/sec
Source out of core
fission chamber 1
fission chamber 2
1.49 counts/sec
2.24 counts/sec
13.44 counts/sec
Source i n core
1
2
27.01 counts/sec
42.59 counts/sec
6.82 counts/sec
1845
1855
1912
1320
1930
2015
18.74 counts/sec
24.18 counts/sec
6.03 counts/sec
L
t
Rods started up. Control rod back to &in. out.
Rods out and count started prior to taking source out of core.
BF,
Storted adding lot from can 7.
Started transferring first 5-lb.
Plugged line from transfer tank t o pump.
Chemists report that uranium concentration of sample 6 taken a t
2.47 t 0.01 wt %.
214
I
These count rates checked very w e l l w i t h count rate meters.
Source i n core
3
t
1720 i s
t
t
3
E X P E R I M E N T Eo1 (continued)
f
Run
(3)
I
J
]
I
October
31, 1954
Time
0035
0230
2000
2006
2230
Attempt to clear line from transfer tank t o pump.
S t i l l unsuccessful.
New transfer line into main fuel pump installed.
Sample 7 drawn off for analysis.
2300
2307
2315
Starting transfer o f first 5 Ib of 25 Ib in can 7.
0.78 min between pips = 43 gpm, indicator on rotameter shows
Transfer line plugged again,
r
J
1
B
Analysis of sample 7 taken a t 2006 shows 1.84 t 0.04 w t %
total uranium.
Grimes estimates that no appreciable concentrate was transferred
last night (i.e., Saturday night) a t the time the line plugged, Therefore,
this sample should be representative of run 2 and supersedes the previous
sample which showed 2.47% (see above).
46 gpm.
November
0008
1315
1320
1335
9
1349
1403
1413
1415
1421
1505
i
f
I
1
1615
1617
4
1707
1720
1731
1742
1753
1804
1838
1908
1940
1
1
2028
Rods inserted.
From data on rod position vs count rate, rods appear t o be ineffective
above 30 in.
Plug removed from transfer line. Can 7 i s s t i l l attached t o the system.
It is estimated that 5 of the 25 Ib i n can 7 were injected l a s t night. The
other 20 Ib are now to be injected i n 5-lb batches.
Started transfer of second 5 lb of 25 I b i n can 7.
Interval between pips on fission chamber 2 was 30 sec. Pips are
becoming hard to see.
Started transfer of third 5 Ib. Interval between pips *0.86 min.
Started transfer of fourth 5 Ib.
Started transfer of last batch from can 7.
Transfer of 25 Ib from can 7 completed. Interval between pips -0.79 min.
Withdrew rods from 30 t o 35 in. Three IO-min counts taken.
Rods inserted,
Set log 4 to come on scale a t an estimated 2 x
w.
Sample 8 taken for analysis.
Ib of 30 Ib from can 8 transferred.
5 Ib transferred.
Third 5 Ib transferred.
Fourth 5 Ib transferred.
Fifth 5 Ib transferred.
Sixth (final) 5 Ib from can 8 transferred.
Analysis of sample 8 taken at 1617 shows 3.45
First 5
Second
*
0.01 w t %
total uranium,
Rods inserted. Check on noise on p i l e period recorder started.
P i l e period meter shows microphonics, Epler put log N back on normal
w.
setting. L o g N should come on scale ~ 1 / 2
Taking sample of fuel system, sample 9.
1, 1954
E X P E R I M E N T E-1 (continued)
Run
5
November
1, 1954
F i r s t 5 Ib from can 9 transferred.
Second 5 Ib transferred.
Third 5 Ib transferred.
Sixth (final) 5 Ib transferred.
Chemists report 5.43 w t % total uranium for sample
21 12
2123
2133
2142
2153
2203
2255
I
t
I
lL
7
8
1
9 taken a t 2028.
1954
* I
f
Second 5 lb transferred.
Third 5 Ib transferred, Two 5-min counts taken.
Fourth 5 Ib transferred,
Sixth (final) 5 Ib trans'ferred,
0104
01 17
0128
0153
0202
0213
0506
0521
0533
055 1
0600
0610
0831
C
c
November 2,
6
tB
Time
I
Second 5 Ib transferred.
Third 5 Ib transferred, Two 5 m i n counts taken.
Fifth 5 Ib transferred,
Sixth (final) 5 Ib transferred.
First
5 Ib from can 5 iniected. Leak i n line occurred.
E X P E R I M E N T E-2
Objective:
1003
1005
1005
1005:lO
1010
1615
1820
2015
2020
Preliminary Meosurement of Temperature Coefficient
Heat barriers started up,
Heat barriers up,
Blower started.
Blower up t o 275 rpm.
Blower off.
Water heat exchanger rose 2 5 O F .
Manometer reading 5.65 in.; corresponds t o
t
170 gpm.
Sample from fuel system taken. This i s sample 10.
Results from sample 10 show 9.58 f 0.08% uranium.
Sample 1 1 taken for analysis,
Referring back t o run 8, experiment E-1:
It i s estimated that 5.5 Ib from can 5 went into transfer tank. Approximately 0.2 Ib was lost i n the leak and 5.3 Ib went into the system.
+
I
Counts vs shim rod position were taken a t 5-in. intervals on shim rods.
Time of day was 1442 to 1512.
2215
Sample
11
showed
9.54 f 0.08%
total uranium.
November 3,
0136
0152
0205
216
Second 5 Ib batch from can 5 iniected. Rods at
intervals for each 5 in. of rod withdrawal,
Rods inserted to 20 in.
Injection nozzle shorted.
20 in,
Counts taken a t
1954
I
f-
t
i
!
E X P E R I M E N T E-2 ( c o n t i n u e d )
November
Run
(8)
Time
0226
0256
J
#
r
1
9
0459
0523
10
0836
12
F i r s t batch consisting of 5.5 Ib from can 11 injected. Rods at 20 in.
Counts taken as rods withdrawn,
Remainder of can 11 iniected. Rods at 20 in, Counts taken as rods
withdrawn.
5.5 Ib from can 22 injected. Rods a t 20 in. Counts taken as rods
1018
1110
5.5 Ib from can 31 injected. Rods at 20 in. Counts taken as rods
1147
1230
1245
withdrawn.
Second 5.5 Ib from can 31 iniected. Rods a t 20 in.
F i s s i o n chamber 2 withdrawn 2 orders of magnitude,
Balance of can 31 injected. Rods at 20 in. Counts taken as rods
w ithdrawn.
1453
5.5 Ib from can 20 injected. Rods at 20 in, Counts taken as rods
1
d
Third 5 Ib botch from can 5 injected. Rods a t 20 in. Counts taken as
rods withdrawn.
F i n a l batch from can 5 injected. Chemists estimate about 2% Ib. Counts
taken as rods withdrawn.
withdrawn.
Second 5.5 Ib from can 22 injected. Counts taken as rods withdrawn.
Completing injection from can 22. Can 22 empty, Counts taken as rods
withdrawn.
Reactor monitored. Reads - 6 mr/hr.
0911
094 1
11
3, 1954
w i t hdrawn.
Source withdrawn
reactor subcritical.
Second 5.5 Ib from can 20 iniected. Rods a t 20 in, Counts taken as rods
withdrawn.
Reactor c r i t ical.
Control given t o servo.
10 mr/hr on
Radiation survey made. Reactor reads 750 mr/hr at side;
g r i l l above main fuel pump.
Reactor shut down.
Shim rods brought out to about 18 in, with source i n core.
-
1536
1545
1547
1603
1604
1626
-
November 4,
0830
0840
Sample
Sample
12 taken.
13 taken.
d
EXPERIMENT L-1
Objective:
1107
1118:40
1
1218:40
1250
1300
1400
i '
I
'Actual
1 w1 ( E s t i m a t e d ) t o D e t e r m i n e P o w e r L e v e l ;
Level C h e c k
One-Hour R u n a t
Radiation
Shim rods coming out.
Reactor up and leveled out. Estimated 1 w0tt.l On servo.
Micromicroammeter 1 x
48.6 on Brown recorder.
Reactor scrammed.
Sample 14 taken.
Sample 15 taken.
Estimated power from sample 15, 1.6 w . ~Count was low. Must be
repeated at 10 w.
power subsequently determined t o be
'See Appendix
H, "Power
2.7
w.
Determinotion from F u e l Activation."
i
1954
E X P E R I M E N T L - 1 (continued)
Run
November
4, 1954
Time
1525
k
Four samples taken from the fuel system since going critical.
Analyzed as follows:
Time
Sample No.
0830
0840
1250
1300
12
13
14
15
t
Uranium (total) (wt %)
12.11
12.21
12.27
12.24
5
f
5
5
t
i
0.10
0.12
0.08
0.12
p.
EXPERIMENT L-2
Obiective: Rod Calibration
0
1651
1657
1707
1
1710
1746
1758
1804
1814
2
2009
201a
2024
2030
VI
k
F u e l Addition
Reactor brought c r i t i c a l before injecting f i r s t penguin.
Rods coming out.
Reactor up t o -1 w. Regulating rod at 13.1 in.
Reactor subcritical.
Penguin 11 iniected.
Reactor up to - 1 w. Regulating rod at 10.6 in. Rod moved
from 13.1 to 10.6 in.
Readiusted rods, Regulating rod at 13.0 in.
Penguin 19 put i n furnace.
Penguin 19 iniected. Time between pips, 0.75 min.
Reactor up t o -1 w. Rod a t 5.7 in.
Reactor scrammed.
Started changing from rod 4 (19.2 g/cm) to rod 5 (36 g/cm).
on f i s s i o n chambers and BF, taken before changing rod.
h
I
L
i
2.5 in.,
L
t
Counts
E X P E R I M E N T L-3
2230
Fuel system characteristics were obtained.
EXPERIMENT L-2
f
(continued)
November
3
0142
0158
0210
0235
0240
0245
0250
0252
0346
0350
0354
0357
0401
218
Started withdrawing rods,
u p t o - 1 w.
1 w again. Temperature had drifted. Regulating rod
Up to
position, 12.1 in.
Penguin 4 injected.
Reactor a t -1 w again. Regulating rod position, 12.1 in.
Reactor subcritical. Apparently penguin didn’t come over
(no dip line).
Reactor at -1 w. Regulating rod position, 11.5 in.
Reactor su bcri t i c a l ,
Reactor a t -1 w. Regulating rod position, 9.9 in.
Reactor subcritical. Regulating rod movement, 1.6 in.
5, 1954
-
L
c.
i
i?
b
i
:
L
1
t
(continued)
November
I
Run
4
5
6
1
i
5, 1954
I
Ti me
0429
0431
0439
0442
0445
Reactor a t -1 w. Regulating rod position,
Reactor su bcr it ica 1.
9.7 in.
0519
0524
0529
0536
0542
Reactor a t -1 w. Regulating rod position, 7.65 in.
Reactor su bcr it ica I.
Reactor c r i t i c a l a t -1 w. Regulating rod position,
Reactor subcritical,
0607
0612
0619
0626
0631
Reactor c r i t i c a l a t -1 w. Regulating rod position,
Reactor su bcrit ica 1.
Penguin 5 injected.
Reactor critical. Regulating rod position, 5.50 in.
Reactor subcr it ica I.
0715
Shim rods 1 and 2 inserted. Reactor shut down due to increase i n
7.8 in. Rod moved 1.9 in.
6.80 in.
6.5
in.
radiation level above fuel pump tank upon adding concentrate t o pump.
During fuel addition for run 9 the background picked up t o -50 mr/hr
and, on run 5 addition, increased to 55 mr/hr. The normal background a t
the point of measurement being -1 mr/hr.
1015
1040
1105
f
i
i
Trimming pump level t o normal operating probe level.
in, below normal operating probe,
Trimming pump level t o estimated
Took fuel sample 16.
'/4
E X P E R l M E N T L-2-A
=
Objective:
j
- -
T e s t on A c t i v i t y o f V e n t L i n e s b y O p e r a t i n g R e a c t o r a t
1 w for
1350
1403
1413
1414
10 min and t h e n v e n t i n g a s Though A d d i n g Fuel
Preliminary data recorded,
Rods 1 ond 2 being withdrawn.
Rods a t 30 in., rod 2 being withdrawn, F i s s i o n chamber
Period meter shows s l i g h t period, -400 sec.
a t f u l l scale,
2 a t upper limit,
1 being withdrawn.
1416
Rod
Rod
1417
Period meter reaches 100 t o 50 sec.
L o g N reading, 2 x
F i s s i o n chamber 2 at 500 counts/sec.
Rod 1 at upper l i m i t
Regu lot ing rod being withdrawn.
1418
1
i
t
Micromicroammeter reading, 10.
F i s s i o n chamber 2 at 1000 counts/sec,
L o g N , 3 x 10-4.
Period, -400 sec.
2 19
4
d
s
E X P E R I M E N T L-2-A (continued)
Run
( 6)
November
5, 1954
i
Time
1420
Log N, 5 x 10-4
F i s s i o n chamber 2, 2 x
Micromicroammeter, 20.
Period, -400 t o 100 sec.
1421
lo3.
P
S
On servo. Micromicroommeter, 38.
L o g N, lo3. Fission chamber 2, 3.5 x
t
k
l o 3 counts/sec.
2 and 1
& - f a
1432
Regulating rod inserted. Shim rods
1515
L o g N and period channel normal and operating correctly
except 5-sec period reverse i s disconnected for duration o f
calibration run.
1940
The activity observed a t 0715 was explained during the run
t o 1432 as activity i n the vent l i n e from the pump.
1945
Analysis of sample
Total Uranium
Chromium
i
Iron
Nickel
L
L
inserted.
n
c
B
1420
c
16 taken at 1105 after the 70-lb removal gave:
12.54 wt % uranium or 11.7% U235
372 ppm
E
5 PPm
<5 PPm
E X P E R I M E N T L-2 (continued)
7
1947
2002
2009
2014
201 6
2020
8
9
10
2 20
4
2025
2032
2050
2148
2156
2204
221 1
2220
2238
2242
2250
2258
2305
2312
2332
2340
2345
2352
2400
-
Started pulling shims.
u p t o 1 w.
Move shim rods - must reset, Started pulling shims.
U p t o -1 w. Regulating rod position, 6.9 in.
Shim rods going in.
Penguin 15 injected.
Up t o - 1 w. Regulating rod position, 6.2 in.
Shim rods going in. Regulating rod movement, 0.7 in.
Up t o power of -1 w. Regulating rod position, 6.8 in.
Reactor subcritical. Shim rods going in.
Up t o -1 w. Regulating rod position, 5.7 in.
Shim rods going in. Reactor subcritical. Regulating rod
movement, 0.6 in,
Started up t o -1 w.
Reactor a t “1 w. Regulating rod position,
Shim rods going in. Reactor subcritical.
6.9
in.
3.3
in,
Reactor a t power of -1 w. Regulating rod position, 13.5 in.
Shim rods going in.
Reactor c r i t i c a l a t -1 w. Regulating rod position, 12.95 t o 13.0.
b
.
i
I
k
Run
11
6, 1954
I
Time
Reactor critical. Adiusted regulating rod t o -3.5 in. w i t h shims 1 and 2.
Shim rods inserted, Reactor subcritical.
Developed gas leak in injection system during injection of penguin 10.
Leak repaired; apparently penguin 10 was injected a t 0100. Rods coming
out,
Reactor critical. Regulating rod position, -2d4 in.
Reactor scrammed.
0020
0026
0100
0212
0223
0228
4
November
Obiective:
One-Hour Run a t 10 w to Make Radiation Survey
and Observe P i t e Period
Shim rods coming out,
0305
Leveled off manually at -0.1 w. Regulating rod position, 7 in.
0315
Pulled
regulating rod to 8 in.
0318
10
w
estimated
power.3 Rod moved from 6.8 to 7.9 on chart paper.
0320: 1
40-sec period according t o the chart, F l u x servo demand,
Micromicroammeter 51 on range 1 x
500.
lo-*.
0420: 1
Reactor scrammed after -1 hr a t
from slope of log N , 51 sec.
0505
051 6
0524
Sample
Sample
Sample
10 w estimated power.
P i l e period
16 taken.
17 taken,
18 taken,
Objective:
1
0651
Leveled out a t -1 w.
Shim 3 a t 30 in. Regulating rod a t 13 in.
Shim 3 a t 33.3 in, Regulating rod a t 3 in.
Shim 3 a t 30 in. Regulating rod a t 6.5 in.
Shim 3 a t 28.1 in. Regulating rod a t 13d in.
2
3
0705
0925
0934
0939
0940
0942
0945
0947
0948
Pulled regulating rod from
d
k
i
i
Calibration of Regulating Rod from Reactor Periods
’Actual
6.53
to
8.61
in.
Shim rods coming out.
Reactor a t 1 w.
Reactor subcri t ica I.
Pump stopped.
Reactor a t 1 w.
Pulled regulating rod.
Reactor scrammed a t -50 w.
Pump started, From log N, 21.3-sec period.
regulating rod moved from 6.61 to 12.66.
power subsequently determined
to
be 27
W.
Period,
22.1
sec.
From recorder:
i
t
E X P E R I M E N T L-2 ( c o n t i n u e d )
Run
November
7, 1954
Time
i
R
Rod Position (in.)
Shim Rods
Regulating Rod
12
-1 w.
0136
Reactor up t o
0138
023 1
0237
0245
0249
0423
Start injection from cans
Injection completed.
Reactor up t o -1 w.
8
1
2
3
32.8
32.4
32.1
- i
- i
b
t
120 and 125.
“
8
28.2
28.2
+
h
27.8
If
Reactor shutdown.
Sample 19 taken for chemical analysis. All fuel injection and sampling
lines removed from pump. It i s estimated that the l i q u i d level i s about
i
0.1 ft3 below the normal operating level probe; therefore final systemvolume4
i s 5.35 ft3.
Results on sample 19:
Uranium (total)
Chromium
13.59 f 0.08 wt %
445 ppm
E X P E R I M E N T L-5 ( c o n t i n u e d )
4
5
6
7
1605
1611
1613
u p t o 1 w.
Scrammed at 100 w. Wrong paper on log N. By super-imposing proper
paper got 27.2 sec.
Regulating rod withdrawal from 2.80 to 4.80 in., from Brown.
1804
1808
1810
u p t o 1 w.
Scrammed at 100 w. From log
2020
2022
2024
2026
2034
2036
2037
Up to 1 w. Pump at 48 gpm.
Reactor subcritica I.
Pump stopped, zero flow.
u p to 1 w.
Reactor scrammed at 60 w.
Started pump. Period, 23.0 sec from slope on log N. Rod moved from
2201
2208
2210
&
N, period i s 22.4 sec; A k / k , 0.06%.
.
2.60 t o 8.70 in., or 6.1 in.
Reactor up t o 1 w.
Scrammed at 100 w. Period, 20.6 sec from slope on log N . Rod moved from
4.19 t o 6.27 in., or 2.08 in. (Brown!
Obiective:
1
2
Calibration of Shim Rods v s Regulating Rod
One set of data taken before final fuel addition.
2307
2353
Up to 1 w. Data recorded.
End of run.
~
‘Inventory
gave 5.33 f t 3 for the volume of the flvoride i n the system.
T
Run
8
Ti me
2353
2358
2400
Obiective:
i "
1
d
Effect of Fuel F l o w Rate
0147
0050
Up t o 1-w power.
0057
0107
0110
0113
0115
0117
0119
Fuel Flow rate,
Fuel Flow rate, 48 gpm.
Fuel Flow rate,
Fuel Flow rate,
Fuel Flow rate,
Fuel Flow rate,
Fuel Flow rate,
Fuel Flow rate,
on
L
November
Delayed Neutrons
12.2 in.
Regulating rod position,
41 gpm. Regulating rod position, 11.5 in.
30.5 gpm. Regulating rod position, 10.5 in.
0 gpm. Regulating rod position, <2 (off-scale).
I
Reset shims.
9
0122
Fuel Flow rate,
10
0125
Fuel Flow rate,
11
0127
Fuel Flow rate,
0 gpm.
12 gpm.
12
13
14
0131
0136
0139
Fuel Flow rate,
Fuel Flow rate,
Fuel Flow rate,
3.6
Reset shims.
in.
12.2 in.
Stopped sodium flow.
15
0141
0142
Fuel Flow rate, 37.5 gpm.
End of experiment.
7.9
in.
E X P E R I M E N T La8
Objective:
1
2.
3
0142
0157
0214
0215
0218
022 1
0222
7, 1954
Start of run, reactor already a t 1 w.
Scrammed a t 200 w. Period, 23.0 sec from slope on log N.
Regulating rod moved from 8.32 to 10.38 in., or 2.06 in.
5
4
November
Measure Reactor Temperature Coefficient
Reactor already up to
Took data.
1 w.
0230
0231
Start prime mover.
Started raising barrier doors.
Helium blower, 255 rpm.
Readjust shims to put regulating rod to top of scale.
Heat exchanger water flow, 195 gpm. Temperature rise,
Steady state.
Low heat exchanger reverse. Prime mover off.
Barrier doors down,
0233
Took data.
62 t o 76OF.
8, 1954
E X P E R I M E N T Lo8 (continued)
Run
(3)
November
8, 1954
Time
0242
Readjusted shims.
3.15
in.
h
L
Reactor mean,
1257' F.
0700
0950
Regulating rod position now 9.55 in,, moved 6.40 in. since 0242
when final rod adjustment was made, Reactor mean temperature,
1286OF; 29OF r i s e since 0242.
End of experiment.
Set Reactor a t
Obiectivet
1
2
3
4
5
6
7
8
0950
0954
1004
1010
1018
1028
1033
1101
1106
1113
1116
1128
1132
1150
1155
1159
1 kw and Adlust Chambers
Reactor already a t 1 W.
Took data.
Pulled safety chamber 1.
Raised power t o -20 w.
Raised power t o -40 w.
Started pulling log N chamber.
Reactor scrammed,
40.w nominal power.
Pulled log
N
chamber a l l the way w i t h i n shield.
Pulled log N out of shield about
Data recorded,
Raised power t o
-400
20 in.
w.
On manual. P u l l i n g micromicroammeter chamber.
Set micromicroammeter t o exactly
Micromicroammeter
50 on 1
x
k
i
500 W.
IO-*;
corresponds t o
500 W.
Put reactor on 30.sec period and l e t safety chambers scram reactor.
E X P E R I M E N T H-1
id
Obiective: Approach to Power (IO-kw run)
1
2
3
1445
1448
1452
1505
1525
1611
1619
instruments on scale
Reactor up t o 10 kw.
Took data.
Pulled safety chambers.
Took data.
Obiactive:
1
2
1125
1127
1138
1208
1224
t
Took data.
Scrammed, F i s s i o n gases i n basement.
lO-kw Run to
November
9, 1954
Monitor Gas Activity
Instruments on scale.
Up t o 10 kw.
Data taken.
Data taken.
Decreased power t o
5 kw.
224
te
c
F
i
i
E X P E R I M E N T H - 2 (continued)
Run
November
9, 1954
Time
3
1227
4
1245
1246
Data taken.
Data taken,
Reactor scrammed.
During run, the estimated p i t activity from
RIA-4 was 0.032 pc/cm3.
EXPERIMENT H-3
Objective:
1
i
i
2
c
1520
1523
1526
Instruments on scale, Period,
17 sec.
Reactor at fi 100 kw.
Fuel and sodium system barrier doors raised,
1529
1530
1551
1600
1621
1625
1626
1643
started.
Blower at -300 rpm.
Safety chambers scrammed reactor at 130 kw.
Instruments on scale. Period, -30 sec.
Reactor scrammed by safety level (set too low).
Instruments on scale. Period, -27 sec.
Reactor on servo a t 50 kw. Safety chambers reset.
Reactor off servo, Power raised.
Power leveled out a t -250 kw. Data taken.
1656
Fuel system prime mover
-
1704
Elevated mean temperature (withdrew shim rods). Changed scales on
reactor AT,
Data taken.
3
1715
1727
Power raised t o
Data taken,
4
1740
1749
1801
1810
1820
Front sodium system blower brought t o -500 rpm.
Back sodium system blower brought to -500 rpm.
Data taken,
?
3
-
1 00-kw Run
j
1
5
1
6
7
-500
kw.
1825
Increased power.
I-Mw nominal power,
Data taken.
1845
Data taken.
1854
1900
1907
1915
1919
1938
1945
2027
Reduced power,
Fuel system blower off.
115 kw.
Data taken;
Rods going in.
Sodium system blowers off.
Started increasing power.
Reactor power, 50 kw.
Reactor power, * 100 kw.
-
Obiective:
1
2045
2100
2102
2103
Fuel and Reactor Temperature Coefficient Determinations
Reactor power level a t about 100 kw.
Turned on fuel system helium blower,
Readiusted shims.
Blower off.
350 rpm.
I
Run
2
November
9, 1954
Time
2121
2124
2126:30
2129
2156
Fuel system blower on, 340 rpm.
Readiusted shim.
Prime mover off. Servo off.
End of run.
Reactor at 10 kw.
Objective:
Reactor a t
1
2
3
2225
2243
2300
Over-all Reactor Temperature Coefficient Determination
100 kw
on flux servo. Allowed t o heat up from
1260 to 1313’F.
Data taken.
Data taken.
Data taken.
Obiectiver Startup on Temperature Coefficient
1
2307
2315
2327
2339
2348
2353
2400
2405
t
-
lncreased power,
Power,
1.3 Mw (fuel system blower speed a t maximum).
Took data.
Reduced power t o -200 kw. Pressure shell temperature getting high.
Inserted shims t o drop reactor mean temperature,
Ran blower up from 0 t o 1700 rpm (maximum speed).
Mark charts,
Ran regulating rod i n from 7 in,; then pulled out a l l the way.
November
2
001 6
0019
0026
0028
0033
0037
01 06
01 15
0137
0204
Reduced fuel system helium blower speed t o zero.
Inserted regulating rod from 14 t o 2 in. Reactor subcritical.
Ran fuel system helium blower speed from 0 t o 1700 rpm.
Low heat exchanger temperature reverse.
Stopped fuel system helium blower.
Inserted shim rods and reduced power. Stopped sodium system
helium blowers.
Reactor left a t 10 t o 100 kw t o maintain temperature on fuel lines.
Reactor scrammed due t o calibration of fuel flowmeter.
Reactor power a t * 100 kw.
Reactor mean temperature a t -1300°F; power a t -75 kw and f a l l i n g
slowly due t o heating of fuel. 5.25 in, of regulating rod movement E
temperature r i s e = 5.5 x
(Ak/k)PF.
10,1954
Y
c
t
L
t
c
b
b
h
-3OOF
L
h
k
PRACTICE OPERATION
Obiectiver T o F a m i l i a r i z e Crews with Reactor Operation
0503
0538
0545
0548
226
4
;1
Reactor power, w 250 kw.
Reactor power, -500 kw.
Stopped fuel and sodium system blowers.
A l l blowers back on a t about 500 rpm.
h
Power leveled a t
“50
kw.
t
hir
f
P R A C T I C E O P E R A T I O N (continued)
J
Run
4
(2)
d
*
November
10, 1954
Time
0552
0554
0600
0602
0615
0625
0630
0844
0852
0855
0858
0905
0909
i
All blowers off.
Blowers on. Power, “750 kw.
Blowers off.
Reactor power, -500 w.
Reactor power, -50 kw. Took data.
Started fuel system blower prime mover.
Started fuel system blower.
Leveled off power a t -200 kw.
K. Z. Morgan reports short-lived air activity in crane bay.
Blower off. Prime mover off. End of experiment. Reactor leveled out
a t “50 kw.
c
Oblective: Sodium Temperature Coefficient Determination
0914
0917
0924
I
Ran sodium system helium blowers from 0 t o -2000 rpm.
Stopped sodium system blowers, Peak power, -., 150 kw.
Leveled out. Charts pulled. Sodium temperature coefficient i s
negative.
t
1
J r ;
?
?
1
b
Obiective:
Measure Effect of
o
-
-50 kw. Started sodium system blowers.
1 Mw.
1040
1045
1057
Reactor power,
Reactor power,
Took data.
1110
1114
1128
1150
1200
1206
1229
1241
Pulled regulating rod one dollar in 0.61 min.
Inserted regulating rod one dollar. Charts pulled.
Scrammed while pulling charts.
150 kw.
Reactor power,
1 Mw.
Reactor power,
Sodium system blowers up t o 1000 rpm each.
Reduced blower speed t o zero,
100 kw.
Reactor power,
i
i
Dollar of Reactivity
--
ah
P R A C T I C E O P E R A T I O N (continued)
I
I
i
:
1300
1305
1310
1313
1315
Blowers started. Blowers a t 500 rpm.
Reactor power, -500 kw. Extracted power, -570 kw.
Blowers t o 1000 rpm.
Regulating rod withdrawn 2 in. Extracted power, 920 kw (fuel)
+ 96 kw (sodium) = 1.016 Mw.
E
227
1
21
d
i
i
E X P E R I M E M T k-1-9
Objective:
Run
Time
(2)
1400
1422
November
10, I954
Reactor power, 300 kw. Fuel mean temperature raised t o 135OOF.
Reactor on servo a t 36 kw. Front sodium system blower on, 600 rpm.
Reactor cooling.
Regulating rod reached lower limit.
Reactor a t 100 kw on servo.
Going up t o 1 Mw.
Mean temperature was raised t o 1350OF; now lowered t o 1325OF t o avoid
hot places on lines.
Mean temperature lowered t o 1315'F a t which point low heat exchanger
temperature interlock reversed he1ium blowers.
1452
1458
1505
1515
1523
Objective:
1
1550
1600
1605
1606
1611
1613
1625
1630
2
1735
1737:30
174s
1747
1752
1755
1802
1805
1808
I
I
228
L
Over-all Reactor Temperature C o e f f i c i e n t Determination
t
P
c"
b
f
Moderator Temperature C o e f f i c i e n t Determination
Lowered reactor t o -400 kw.
Fuel temperature, 1317OF; regulating rod position, 7.2 in.; reactor outlet
sodium temperature, 128OOF; reactor power, 500 kw.
Front sodium system blower, 870 rpm. Back sodium system blower, 1260 rprn.
Reduced back sodium system blower speed t o 870 rpm. Regulating rod
position, 7.5 in.; fuel temperature, 1317OF; sodium temperature, 128OoFe
Temperatures same as a t 1606. Regulating rod position, 8.3 in.
Front sodium system blower speed reduced t o 670 rpm; back sodium system
blower speed t o 850 rpm. Regulating rod position, 8.7 in.; fuel temperature, 131pF; sodium temperature, 1278OF.
Front blower a t 650'rpm; back blower a t 640 rpm. Regulating rod position,
7.6 in,; sodium temperature, 1278OF.
Reactor power a t 1 Mw for demonstration t o Air Force personnel.
Regulating r c d position, 8.0 in,; reactor mean temperature 1313OF; front
blower a t 960 rpm; back blower a t 1050 rpm; sodium inlet temperature,
1265OF (thermocouple 3-CR-1) sodium outlet temperature, 1282OF
(thermocouple 3-CR-5).
Front blower a t 1180 rpm.
Reactor mean temperature, 1314OF; regulating rod position, 8.0 in. Front
blower a t 1160 rprn; back blower a t 1050 rpm. Sodium i n l e t temperature,
1255OF; sodium outlet temperature, 128OOF.
Back blower a t 1280 rpm.
Reactor mean temperature, 1313OF. Regulating rod position, 7.6 in.
Front blower a t 1170 rpm; back blower a t 1260 rpm. Sodium i n l e t
temperature, 1248OF; sodium outlet temperature, 1278OF.
Front blower raised t o 1350 rpm.
Regulating rod position, 6.6 in. Reactor mean temperature, 1308OF.
temperature, 1238OF; sodium outlet temperature, 1272OF.
Front blower raised t o 1490 rprn.
Regulating rod withdrawn 1.3 in. t o regain 5OF l o s t in reactor mean
temperature. Rod position = 7.3.
b
c
h
i
f
k
L
f
E X P E R I M E N T W-110 ( c o n t i n u e d )
November
10, 1954
I
Time
8812
1816
9825
temperature, 1230°F; sodium outlet temperature, 1268OF.
Bask blower speed changed t o 1500 rpm.
Front blower at 1470 rpm; back sodium blower a t 1480 rpm.
Sodium i n l e t temperature, 11225OF; sodium outlet temperature, 1268OF.
t
i
Objectives
1935
!
25-Hour Run a t F u l l P o w e r to O b s e r v e X e n o n P o i s o n i n g
Since 1825, the reactor has been steady a t 90% f u l l power on servo; i t
i s now o f f servo. The 25-hr xenon run i s considered t o have started
at
1825.
November
0304
1145
1206
1435
Cooling water flow t o back sodium heat exchanger decreased. Sodium
AT f e l l -3°F. No explanation as yet.
Drop i n sodium A T due t o heaters for annulus helium being shut off.
Withdrew regulating rod from 10.3 t o 10.4 in. Reactor mean temperature
dropped to 1309OF from an i n i t i a l temperature of 1311OF.
Regulating rod withdrawn from 10.4 t o 10.5 in.
Micromicroammeter took a small dip (53 t o 52) for no apparent reason.
Adding helium t o sodium and fuel system ducts.
Fuel flowmeter coils calibrate OK. There i s a possibility of something
binding in the Rotameter t o give the erratic readings. Switched t o
operation on No. 3 coil.
Regulating rod withdrawn -0.05 in. Indicator now reading 9.25 in.
P i l e period shows unexplained excursion t o 400 sec.
Noticed fluctuations in the fuel pump ammeter recorder corresponding
t o fluctuations in the Rotameter chart. These Fluctuations appear t o
be between 9.0 and 9% amp.
Noticed same pump current and flow level fluctuations as a t 1130.
Switched back t o No. 5 c o i l on fuel flow Rotameter.
Regulating rod withdrawn 0.1 in.
1445
For the past
P
0400
0635
I
0750
0825
0915
0925
1020
1100
1 I30
a
s'
d
..
11, 1954
k
!
k
hr have noticed corresponding fluctuations in the
following instruments: fuel flow Rotameter, fuel motor current
recorder, period meter, log N recorder, safety levels 1 and 2, reactor
AT, the s i x individual tube AT'S, and the micromicroammeter. A l l
these fluctuations a t present are small, especially on the micromicroammeter. The largest fluctuations appear t o be on the pump
motor current, between 9 and 10 amp; the p i l e period meter, M t o f 400
sec; and the s i x AT recorders, k 2 t o 3OF. No extraneous audio noise
has been detected. Noticed increase in fuel pump pressure since 0800
o f 0.2 psi. A t 0800 pressure was 0.2 psi, a t present it i s 0.4 psi.
1
1
229
d
d
E X P E R I M E N T H-11 ( c o n t i n u e d )
Run
Time
(2)
1935
2032
November
11, 1954
4
I
End of xenon run, Regulating rod has been withdrawn -0.3 in., which
corresponds t o 0.01% Ak/k as compared t o -0.30% Ak/k if a l l xenon
c
had stayed in system,
Had demonstration for Air Force personnel, mostly 1-Mw nominal power.
Set reactor back t o original condition l e f t a t 1935.
Objective:
Stop Sodium System B l o w e r and O b s e r v e E f f e c t on P o w e r
1
2105
Took data.
2
21 13
21 19
Sodium system blower motors off,
Took data, Blowers back on, Maximum deviation of micromicroammeter was
from 53 t o 45, 15%.
Reduced power by cutting off fuel system blower.
Reduced speed on sodium system blower.
Regulating rod inserted; reactor power, * 10 kw.
Started fuel system blower motor,
Started fuel system helium blower.
Approached low heat exchanger temperature reverse (115OoF) which shut off
fuel system blower motor. Withdrew regulating rod.
Reactor power back up t o 1 Mw.
Cut o f f fuel system biower. Cooling reactor w i t h sodium system blowers
only.
100 kw.
Reactor power,
2150
2155
2158
2200
2201
2204
2210
2230
2237
-
L.
L.
t
k
i
Objective:
2237
0605
0735
0835
Measure Xenon Buildup at
Consider experiment t o have started a t
-
t oF u l l
2237 a t
Power
'/lo f u l l power,
i.e.,
-100 kw.
November
12, 1954
Regulating rod withdrawn from 5.05 t o 5.15 in. Reactor mean temperature
dropped from 1304 t o 1303OF.
Sodium AT up due t o addition of helium i n rod cooling system. Reactor
mean temperature down t o 1302OF.
No appreciable xenon buildup. End of experiment.
Obiective:
0837
0855
0907
0910
0917
230
D e t e r m i n e M a x i m u m R e a c t o r P o w e r and C h a r a c t e r i s t i c s o f P o w e r O p e r a t i o n
Demonstration of reactor operation.
Reactor power up t o 1 Mw. Maximum fuel system helium flow. Reactor
mean temperature,
1340OF.
No. 2 rod cooling helium blower on slow speed. Helium pressure,
25 in. H,O.
No. 2 rod cooling blower off. Helium pressure, 9 in. H,O
.
L
No. 2 rod cooling blower on,
-
-
'
?
8
I
t
k
E X P E RlMENT H-14 ( c o n t i n u e d )
Run
Time
(2)
0919
1
1
1
d
2
0922
0935
1015
1055
1102
1109
1110
1 1 1 1 :30
1112
1113
1115
1117
1125
i
12, 1954
No. 2 rod cooling blower off. Small wobble in AT probably due t o
helium blowing on thermocouple, No systematic effect on micro-
-
microammeter can be observed.
1 Mw.
Started demonstration. Average reactor power,
Demonstration ended. Reactor power back t o 1 Mw.
Reactor mean temperature raised t o 135OOF.
Fuel system blower off.
Power leveled off a t -350 kw. Sodium system blowers off.
Power leveled off a t -210 kw. Rod cooling blowers off.
Power leveled o f f a t 200 kw.
Started bringing reactor up t o full power.
14-sec period observed,
Heat exchanger outlet temperature reduced from 1380 t o 1175OF.
L o g N from 4 t o 30.
Back t o power of >1 Mw.
Moved shim and regulating rods t o obtain new heat balance.
1150
1230
1244
1306
1306
1308
1325
1326
1330
1331
Sodium system blower speed up t o 2000 rpm. Extracted power: from
fuel, 1920 kw; from rod cooling, 20 kw; from sodium, 653 kw; total,
2.6 Mw.
No. 2 rod cooling blower started.
Front sodium system blower off; reactor power reduced t o 100 kw.
1.5 Mw.
Reactor mean temperature and power leveled off; power,
Reduced fuel system helium blower speed t o minimum.
Fuel AT nearly constant.
Raised fuel system helium blower speed t o maximum.
Fuel AT essentially restored.
F u e l system blower turned on; speed reached 1500 rpm.
Blower speed reduced t o 1000 rpm.
Reactor AT about 200OF.
Regulating rod withdrawn t o raise mean temperature from 1320 t o
1334
Regulating rod inserted t o lower mean temperature from 1325 t o
1336
1350
1400
1615
Regulating rod withdrawn t o raise mean temperature,
After demonstration, power leveled o f f t o 1.5 Mw.
Demonstration of reactor operation.
End of demonstration of reactor operation. Reactor run more or less
steadily a t 2 Mw until time for scram,
1930
2004
Reactor power up t o
Reactor scrammed.
1140
i
November
i
t
i
--
1325OF.
i
1320OF.
3
J
-2.5
i
Mw.
i
h
b
1
1
#
23 1
i
h
Appendix U
THE A R E BUILD"
The
ARE building, shown in Fig. U.1, i s a m i l l
type of structure that was designed t o house the
ARE and the necessary f a c i l i t i e s for i t s operation.
The building has a f u l l basement 80 by 105 ft,
a crane bay 42 by 105 ft, and a one-story service
wing 38 by 105 ft. The reactor and the necessary
heat disposal systems were located in shielded
p i t s i n the part of t h e basement serviced by the
crane. One half the main floor area was open t o
the reactor and heat exchanger p i t s i n the basement
below; the other one half housed the control room,
o f f i c e space, shops, and change rooms.
In one half the basement were the shielded
reactor and heat exchanger pits; the other one
half of the basement was service area and miscellaneous heater and control panels.
The control
room, office space, and some shops were located
on the f i r s t floor over the service area. T h e f i r s t
floor does not extend over the one half o f t h e
basement that contains the pits.
The crane i s
a floor-operated, 10-ton, bridge crane having a
maximum lift of 25 ft above the main floor level.
Plan and elevation drawings of the building are
shown i n Figs. U.2 and U.3.
The entire reactor system was contained i n three
interconnected pits: one for the reactor, another
for the heat exchangers and pumps, and a t h i r d
for the fuel dump tanks. These pits, which were
sealed at the top by shielding blocks, were located
in the large crane bay of the building. T h e crane
bay was separated from the control room and
offices,
and the heating and air conditioning
systems maintained the control room at a slightly
higher pressure than that of the crane bay.
8
d
t
t
k
t
i
e
k
1
e
i
c
Fig.
U.l.
The
ARE Building.
I
i
?
i
1
t
i
I
1
B
i
T
___--.__.-
""I
. . . . . . . . . . IY
LLL IC
T
T
/
e
z
a
-I
FZ
4
m
2
w
0
f
4
..
$
-
'i
I
1:
,
234
"
1-
e
W
U
i
a
J
m
w
u)
0
Q
3
1-u19 1102
--
-
=r
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UJ
u
m
m
z
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)
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i
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b
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?
BIBLIOGRAPHY
Aircraft Nuclear Propulsion Project Quarterly Progress Reports for the Periods Ending:
March 10, 1951
ANP-60
June 10, 1951
ANP-65
September 10, 1951
ORNL-1154
December 10, 1951
ORNL-1170
March 10, 1952
ORNL- 1227
June 10, 1952
March 10, 1953
0RN 11294
OR NL-1375
ORNL- 1439
ORNL-1515
June 10, 1953
ORNL-1556
September 10, 1953
ORNL-1609
December 10, 1953
ORNL- 1649
March 10, 1954
ORNL- 1692
June 10, 1954
ORNL-1729
September 10, 1954
0RNL-1771
ORNL- 1816
September 10, 1952
December 10, 1952
4
1
December 10, 1954
-
i
235
DATE
TITLE
AUTHOR(s)
REPORT NO.
1948
Metals Handbook (Nickel and Nickel A l l o y s ,
p 1025-1062)
American Society for Metals
12-1 5-49
The Properties of Beryllium Oxide
M. C. Udy, F. W. Boulger
BMI-1-78
5-21-51
A c t i v a t i o n of Impurities in B e 0
W. K. Ergen
Y F 20-14
4-1 7-5 1
Perturbation Equation for the K i n e t i c Response of a L i q u i d - F u e l Reactor
N. Smith et a l .
ANP-62
6 -5-5 1
The Contribution of the (n,2n) Reaction t o
the Beryllium Moderated Reactor
C. B. M i l l s
N. M. Smith, Jr.
7-25-51
Radiation Damage and the ANP Reactor
L. P. Smith
AN P -67
*
8-13-51
A l k a l i Metals Area Safety Guide
Y-12 A l k a l i and L i q u i d Metals
Safety Committee
y-811
i
8-29-51
P h y s i c s Calculations on the A R E Control
J. W. Webster
Y-FIO-71
-
e
Rods
R. J. Beeley
Results of C r i t i c a l i t y Calculations on
B e 0 and B e Moderated Reactors
y - F10-55
J. W. Webster
0. A. Schulze
A N P -66
1-5-52
P h y s i c s o f the Aircraft Reactor Experiment
C, B. M i l l s
Y -F 10-77
1-8-52
Statics
C. 8. M i l l s
Y -F 10-81
1 -1 1-52
The ARE with C i r c u l a t i n g Fuel-Coolant
C. B. M i l l s
Y-F 10-82
1-14-52
A Flux Transient Due t o
t i v it y Coef f i ci ent
C. B. M i l l s
y-F10-63
1-2 2- 52
Heat Transfer
R. N. L y o n
CF-52-1-76
1-29-52
Safety Rods for the ARE
W. B. Manly
CF-52-1- 192
in Nuclear
Preliminary
P o s i t i v e Reac-
a
Reoctors
G
c
P;
li
Some
-A
i
c
10- 15-51
of the A N P Reactor
Report
‘
E. S. Bomar
1-29-52
E f f e c t o f Structure on C r i t i c a l i t y o f the A R E
of January 22, 1952
C. B. M i l l s
Y - F I 0-89
3 -10-52
A Simple C r i t i c a l i t y Relation for B e Moderoted Intermediate Reactors
C. B. M i l l s
Y-F 10-93
3-20-52
Health P h y s i c s Instruments Recommended
T. H. J. Burnett
c F-52-3-147
T. H. J. Burnett
C F -52-3-1 72
Y-F 10-96
for ARE B u i l d i n g
- ARE
3 -2 4-5 2
Induced A c t i v i t y in Cooling
3-28-52
Optimization of Core Size for the C i r c u lating-Fuel ARE Reactor
C. B. M i l l s
4-16-52
P h y s i c s Considerations o f C i r c u l a t i n g Fuel
Reactors
W.
4-22-52
Note on the
5-8-52
Statics
6-2-52
Reactor Program of the Aircraft Nuclear
Propulsion Project
8-8-52
The ARE C r i t i c a l Experiment
Water
Linear K i n e t i c s of the ANP
C i r c u l a t i n g - F u e l Reactor
o f the ARE Reactor
K. Ergen
Y-F 10-98
F. G. Prohammer
Y-F 10-99
C. 8. M i l l s
Y-F 10-103
Wm. B. C o t t r e l l (ed.)
0 RN L-1234
C. B. M i l l s
D. Scott
Y-F 10-108
c
236
TITLE
DATE
7 -3 0-52
AUTHOR( s)
Loading of ARE Critical Experiment Fuel
REPORT NO.
D. Scott
Y-B23-9
Tubes
a
9-52
Welding Procedure Specifications
P. Patriarca
p5-1
9-52
Welders Qualification Test Specifications
P. Patriarco
qt5-1
9-25-52
Radiation Through the Control Rod Penetrations of the ARE Shield
H. L. F. Enlund
y- F30-8
10-22-52
Xenon Problem
W. A. Brooksbank
CF-52-10-187
11-1-52
Structure of Norton’s Hot-Pressed Beryllium
Oxide Blocks
L. M. Doney
CF-52-11-12
11-24-52
Aircraft
J. H. Buck
ORNL-1407
Summary
- ANP
Reactor
Report
Experiment
Hazards
W. B. Cottrell
11-13-52
Effects of Irradiation on Be0
11-17-52
Structure
11-25-52
An On-Off Servo for the ARE
of Be0 Block with the 1’/8
Central Hole
in.
G. W. Keilholtz
CF-52-11-85
L. M. Doney
CF-52-11-146
S. H. Hanauer
CF-52-11-228
E. R. Monn
J. J . Stone
1-7-53
Delayed Neutron Damping of Non-Linear
Reactor Oscillations
W. K. Ergen
CF-53- 1-64
1-9-53
ARE Regulating Rod
E. R. Mann
S. H. Hanauer
CF-53-1-84
1-12-53
A Guide for the Safe Handling of Molten
Fluorides and Hydroxides
Reactor Components Safety
Comrnittea
Y-B3l-403
1-27-53
Delayed Neutron Activity in
Fuel Reactor
H. L. F. Enlund
CF-53-1-267
Y
-
a
Circulating
B. Cottrell
1 27-53
Components of F Iuor ide Sy stem s
Wm.
1-27-53
Delayed Neutron Activity in the ARE Fuel
Ci rcuit
H. L. F. Enlund
CF-53-1-317
2-45
Some
B. B. Betty
W. A. Mudge
Mech. Eng. 73,
123 (1945)
2-1 1-53
Heating by Fast Neutrons in
Crete Shield
F. H. Abernathy
H. L. F. Enlund
CF-53-2-99
2-26-53
Methods o f Fabrication o f Control and
Safety Element Components for the Aircraft and Homogeneous Reactor Experi-
J . H. Coobs
ORNL-1463
i
Engineering Properties of Nickel and
High-Nickel Alloys
a
Barytes Con-
CF-53-1-276
E. S. Bomar
ments
3- 17-53
Corrosion by Molten Fluorides
L. S. Richardson
D. C. Vreeland
W. D. Manly
3-30-53
The Kinetics of
Nuclear Reactor
W.
4-7-53
Minutes of the Final Meeting of the ARE
Design Review Committee
the
Circulating-Fuel
K. Ergen
W. R. G a l l
ORNL-1491
CF-53-3-231
CF-53-4-43
237
DATE
TITLE
AUTHOR( s)
REPORT NC
5-18-53
ARE Control System Design Criteria
F. P. Green
CF-53-5-238
6-19-53
The Stability of Several Inconel-UF4 Fused
Salt F u e l Systems Under Proton Bombard-
W. J. Sturm
0 RNL -1530
ment
7-20-53
Current Status of the Theory of Reactor
Dynamics
R. J. Jones
M. J. Feldman
W. K. Ergen
c F- 53-7- 137
-
7-20-53
Interpretation
Joel Bengstan
c F-53-7 190
7-20-53
Composition of ARE for Criticality Calcu-
J. L. Meem
Secret
of Fission Distribution in
ARE Critical Experiments
rough drl
tp
i
*
t
I ation s
8-3-53
ARE Fuel Requirements
J. L. Meem
CF-53-8-2
8-10-53
Thermodynamic and Heat Transfer Analysis
of the Aircraft Reactor Experiment
B. Lubarsky
ORNL-1535
9-1-53
ARE Fuel System
G. A. Cri sty
CF-53-9-2
9-1 -53
Static Analysis of the ARE Criticality Ex-
C. B. M i l l s
CF-53-9-19
J. L. Meem
CF-53-9-15
C. B. Mil Is
ORNL-1493
E. S. Bettis
CF-53-9-53
B. L. Greenstreet
(A Preliminary Report Pending
Reception of the ARE Criticality Report)
periment
9-3-53
Experimental Procedures on the ARE (Pre-
I im indr y)
’
9-22-53
The General Methods of Reactor Analysis
Used by the ANP Physics Group
9-25-53
Supplement to
Aircraft Reactpr Experiment
Hazards. Summary Report (ORNL-1407)
W. B. Cottrell
10-18-53
Preliminary Critical Assembly for the Aircraft Reactor Experiment
D. Callihol
D. Scott
ORNL-1634
12-1-53
ARE Design Data
W. B. Cottrell
CF-53-12-9
12-22-53
The lnhcur Formula for a Circulating-Fuel
Nuclear Reactor with Slug Flow
W.
12-18-53
Analytical
3-2-54
K. Ergen
CF-53-12-108
G. J. Nessle
c F-53-12-112
ARE Design Data Supplement
W. B. Cottrell
c F-54-3-65
4-7-54
ARE Instrumentation L i s t
R. G. Affel
CF-54-4-218
5-7 -54
Analysis of Critical Experiments
C. B. Mills
c F-54-5 -5 1
7-20-54
ARE Operating Procedures,
Nuclear Operation
W. B. Cottrell
CF-54-7-143
7-27-54
ARE Operating Procedures, Part I I , Nuclear J. L. Meem
Operations
and Accountability Report
ARE Concentrate
Part I,
on
Pre-
CF-54-7-144
L
7-31-54
Fuel Activation Method for Power Determination of the ARE
E. B. Johnson
c F-54-7-1 1
8-28-54
Critical Mass of the ARE Reactor
C. B. Mills
CF-54-8 -171
To be issued
Stress Analysis
R. L. Maxwell
J. W. Walker
0 RNL-1650
of the ARE
.

ORNL-1845: 1955-09 (19.9 MB PDF)

Transcription

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