Proceedings of

Proceedings of ICONE23
rd
23 International Conference on Nuclear Engineering
May 17-21, 2015, Chiba, Japan
ICONE23-1561
USE OF PA-231 FOR AXIAL POWER DISTRIBUTION FLATTENING OF THORIUM
FUEL CANDLE HIGH TEMPERATURE GAS-COOLED REACTORS
Peng Hong Liem
Nippon Advanced Information Service
(NAIS Co., Inc.)
416 Muramatsu, Tokaimura, Ibaraki, Japan
Hoai Nam Tran
Institute of Research and Development,
Duy Tan University,
K7/25 Quang Trung, Da Nang, Vietnam
Hiroshi Sekimoto
Tokyo City University
1-28-1 Tamazutsumi, Setagaya-ku,
Tokyo, Japan
Keywords: Pa-231, thorium fuel cycle, CANDLE, HTGR,
axial power distribution flattening, small sized long life reactor
with CANDLE active core height of 800 cm and reactor
thermal power of 30 MWth.
ABSTRACT
The innovative CANDLE (Constant Axial shape of
Neutron flux, nuclide densities and power shape During Life of
Energy producing reactor) burnup strategy has been
successfully applied to both fast and thermal reactors. In
particular for thermal reactor applications, CANDLE blocktype high temperature gas-cooled reactors (HTGRs) with either
uranium or thorium fuel cycle had been proposed and
investigated for their simple and safe reactor operation, and the
ease of designing a long life reactor. Small sized long life
CANDLE HTGRs with thorium fuel shows superior burnup
performance than the ones with uranium fuel but their axial
power peaks are relatively higher which may not be
advantageous during a depressurization accident. In this work,
we proposed and investigated the use of Pa-231 mixed
homogeneously in the (Th-232/U-233)O2 fuel kernel of the
TRISO particles to obtain lower axial power peaks. Addition of
Pa-231 decreases the required amount of natural gadolinium
burnable poison in the fresh fuel for establishing a valid
CANDLE HTGR design since Pa-231 has a large thermal
absorption cross section. Besides the role as a burnable poison
nuclide, Pa-231 also serves as a fertile nuclide during the
CANDLE burning since Pa-231 is finally transmuted to a fissile
U-233 nuclide. A promising analysis result shows that for U233 enrichment of 15 w/o and Pa-231 addition of 7.50 w/o, the
axial power peak is decreased from 5.9 to 3.6 W/cm 3 while the
averaged burnup is increased from 138 to 149 GWd/t. This
extends the core life time about 16 %, i.e. from 35 to 41 years
NOMENCLATURE
kinf
: infinite medium neutron multiplication factor
keff
: effective neutron multiplication factor
r (cm) : radial direction in 2-D R-Z reactor geometry
z (cm) : axial direction in 2-D R-Z reactor geometry
1. INTRODUCTION
The innovative CANDLE (Constant Axial shape of
Neutron flux, nuclide densities and power shape During Life of
Energy producing reactor) burning scheme was proposed by
Sekimoto et al. [1] originally aimed for fast reactors. Under this
burning scheme, the burning region moves autonomously with
a constant velocity along the core axis from bottom to top (or
from top to bottom) as shown in Fig. 1. As shown in the figure,
the core can be roughly divided into three regions: (1) fresh fuel
region (kinf <1), (2) burning region (kinf >1) and (3) spent fuel
region. When the burning scheme is applied to a block-type
HTGR, burnable poison (for e.g. natural gadolinium) is used to
adjust the kinf of the fresh fuel to be sub-critical.
The CANDLE HTGR burn-up process is as follows (Fig.
2). Neutrons leaked from the burning region into the fresh fuel
region will be absorbed by the burnable poison and the burning
region will move slowly into the fresh fuel region with depleted
burnable poison. In the burning region, depletion of fissile
material for energy production is accompanied by conversion of
fertile material into fissile material. The spent fuel region is the
region left by the burning region which contains mainly fission
products and depleted fuel.
Copyright © 2014-2015 by JSME
Start
On load
Finish
Fresh Fuel Region
Burning Region
Spent Fuel Region
Fig. 1 CANDLE burnup concept.
BP
Fissile
Fuel
BP
Fissile
Fuel
Neutron Flux
Neutron Flux
Burning
Direction
FP
Equilibrium
Condition
FP
End of Life
BP：Burnable Poison
FP：Fission Products
Fig. 2 CANDLE burnup application to HTGR.
After the CANDLE HTGR is operated for a certain core
life time, the reactor can be shut down for refueling. If the core
active height is design properly the CANDLE concept may
feature a long life HTGR design as will be shown later.
For a unique combination of core geometry and fresh fuel
composition, one can find an equilibrium critical condition
where CANDLE burning scheme is realized. Under the
equilibrium condition, the moving (axial) velocity of the
burning region is constant. Analytical codes for obtaining either
the equilibrium condition or for simulating the reactor start-up
have been developed. The details of the computational
procedures are not given here and readers should consult other
references [1, 2, 3, 4].
In our previous works as well as in the present work we
consider small sized long life prismatic/block-type HTGRs
adopting CANDLE burning scheme which can take full
advantages of CANDLE properties:
1.
Constant reactor parameters (e.g. power peaking,
reactivity coefficients etc.) during reactor operation. This will
simplify not only the reactor design itself and its licensing
process but also simplify its reactor operation and maintenance.
2.
No requirement for burn-up reactivity control
mechanism. Besides simplifying the reactor design, a severe
control rod ejection accident during full power operation (under
nominal pressure) can be avoided.
3.
Proportionality of core height to reactor core life. A
long life core can be easily designed by adjusting the core
height.
4.
Sub-criticality of fresh fuel. No criticality accident will
occur during transportation and storage of fresh fuels.
In addition, application of CANDLE burning scheme to
small sized long life HTGRs can be realized by the present
coated fuel particle and HTGR reactor technologies.
As for the fuel cycle, CANDLE HTGRs can be applied
for uranium fuel cycle [2], thorium fuel cycle [4] as well as for
effective incineration of weapon grade plutonium [5]. In our
previous work [4] we showed that CANDLE HTGRs with
thorium fuel gave better burnup performance than the ones with
uranium fuel. This fact is also true for other thorium fueled
HTGRs of the pebble bed type with once-through-then-out
(OTTO), multipass as well as peu-a-peu (PAP) schema as
described, analyzed and reviewed comprehensively by Liem et
al. [6]. Disregarding the type of reactor and fueling scheme, the
thorium CANDLE HTGRs show relatively higher axial power
peaking factors which may not be advantageous during a
depressurization accident, especially if one try to increase the
reactor thermal power.
In this work we investigated the use of Pa-231 mixed
homogeneously in the fuel kernels for flattening the axial
power peaking factor of the thorium CANDLE HTGRs. The
motivation of the present investigation is as follows. Pa-231 has
a large thermal neutron absorption cross section (201.7 barns),
which is about one or two orders greater than that of Th-232
(7.4 barns) and U-238 (2.7 barns). After absorbing two thermal
neutrons, Pa-231 is transmuted to U-233 fissile isotope.
Therefore, Pa-231 has a potential to be used as burnable poison
(BP) in the early stage and fertile fuel in the later stage of
burnup. The U-233 bred from Pa-231 will contribute to increase
the reactivity at the later stage of burnup, i.e. broadening the
burning region which in turn will lower the axial power
Copyright © 2014-2015 by JSME
peaking factors. A feasibility study of U-Th-Pa fuel in pebble
bed reactors was investigated, in which additional Pa-231
content of 3-5% increases fuel burnup performance of about
30% [7]. A difficulty is that natural abundance of Pa-231 is very
little due to its short half-life (3.28×104 years) compared to
other rare earth elements. However, Pa-231 can be produced
from Th-231 via a (n, 2n) reaction followed by a beta decay
with a half-life of about 1 d [8]. This reaction has threshold
energy of about 7 MeV, and would be suitable with fast
neutrons from an accelerator driven system or a fusion reactor
[9].
2.2 CANDLE Burnup Equilibrium and Critical Search
Procedure
In order to obtain a CANDLE burnup equilibrium and
critical condition, in the first stage, one has to determine the
fresh fuel kernel composition, i.e. the fissile U-233
(enrichment), natural Gd burnable poison and the small
addition of Pa-231 weight fractions (the rest is Th-232), and
search the CANDLE burnup equilibrium condition.
2. CALCULATION MODELS AND CONDITIONS
An in-house analytical tool for obtaining the CANDLE
burnup equilibrium condition has long been developed in our
previous works for both fast and thermal reactors. The code
takes into account the burning region movement, nuclides
burnup and criticality equations simultaneously. The details of
the computational procedures are given in references [1, 2, 3,
4]. In this section the group-constants preparation, CANDLE
burnup equilibrium and critical search procedure as well as the
calculation conditions are discussed.
2.1 Group Constants Preparation
The analytical tool used for obtaining the CANDLE
burnup equilibrium condition, in principal, needs (1) effective
microscopic cross sections and (2) burnup related data such as
fission product yields, decay constants, branching ratios and
depletion chain. The effective microscopic cross sections were
prepared using the collision probability (PIJ) module of SRAC
code system [10] with a SRAC library based on the JENDL-3.3
evaluated nuclear data [11]. The burnup related data were taken
directly from the SRAC library. Figs. 3 and 4 show the SRAC
calculation model for the group-constants generation. Since we
adopted the JAEA High Temperature Engineering Test Reactor
(HTTR, 30 MWth) type HTGR fuel, in the 2-D hexagonal fuel
lattice, the cell is divided into annulus fuel compact, graphite
sleeve, annulus helium coolant channel and graphite block. Use
of TRISO coated fuel particles in the HTGR fuel compact
demands double heterogeneity calculation feature which is
provided by SRAC code system in its PIJ module. Furthermore,
in the resonance energy region, the ultra-fine energy group
capability of SRAC (PEACO module with its MCROSS
ultrafine group library) was utilized to obtain accurate effective
accurate cross sections on the energy region.
After obtaining the effective microscopic cross sections
in 107 energy group (the largest number for SRAC code and
library) then the cross sections were collapsed into 4 energy
group (Table 1) to be used for the CANDLE burnup
calculations. The depletion chain which consists of 29 heavy
metal, 66 important fission product (including one pseudo
fission product) nuclides and 16 burnable poison nuclides is
based on SRAC’s THCM66FP chain [10].
Fig. 3 TRISO coated fuel particle and HTTR type fuel compact
Fig. 4 Cell calculation model by SRAC2006 code system
Copyright © 2014-2015 by JSME
In this stage, the obtained equilibrium condition may give
an effective neutron multiplication factor (keff) which is not
critical. In the second stage, only the natural gadolinium
burnable poison concentration is adjusted to obtain a critical
(keff=1.0) and equilibrium condition of CANDLE burnup. This
two-stage iterative procedure is terminated if the keff is near 1.0
within 1 % convergence criterion. There is a possibility that a
certain composition of U-233 enrichment and Pa-231 weight
fraction will not give a critical and equilibrium condition of
CANDLE burnup.
Table 1. Neutron energy group structure for CANDLE burnup
calculations (4-group)
Group
1
2
3
4
Fission
Slowdown
Resonance
Thermal
Energy (eV)
Upper
Lower
1.0000E+07 1.1109E+05
1.1109E+05 2.9023E+01
2.9023E+01 2.3824E+00
2.3824E+00 1.0000E-05
Lethargy
Upper
Lower
0
4.5000
4.5000
12.7500
12.7500
15.2500
15.2500
27.6310
Table 2. Design parameters of small sized long life thorium
CANDLE high temperature gas-cooled reactor
Thermal power (MWth)
30
Core
Diameter (cm)
230
Active height (cm)
800
Radial reflector (graphite)
Thickness (cm)
100
Coated fuel particle
Fuel
(U-233/Th-232)O2
U-233 enrichment (w/o)
10 and 15
Burnable poison material
Natural gadolinium
Type
TRISO
Kernel diameter (mm)
0.608
Particle diameter (mm)
0.940
Coating material
PyC / PyC / SiC / PyC
Thickness (mm)
0.060 / 0.030 / 0.030 / 0.046
Density (g/cm3)
1.143 / 1.878 / 3.201 / 1.869
Packing fraction (v/o)
30.0
Fuel compact
JAEA HTTR type
Inner diameter (cm)
1.00
Outer diameter (cm)
2.60
Graphite sleeve
Inner diameter (cm)
2.60
Outer diameter (cm)
3.40
Coolant annulus channel
Inner diameter (cm)
3.40
Outer diameter (cm)
4.10
Fuel pitch
Flat to flat distance (cm)
6.60
2.3 Calculation Conditions
In order to investigate the effect of Pa-231 addition on
the axial power peaking of a thorium CANDLE HTGR, we first
set the Pa-231 concentration equal to zero to obtain a
comparison design case for each U-233 enrichment (10 and 15
w/o), followed by several cases of Pa-231 concentration values.
It should be noted that the Pa-231 is mixed homogeneously in
the TRISO fuel kernels. The reactor design parameters shown
in Table 2 are identical with our previous work for thorium
CANDLE HTGR by Ismail et al. [4]. These design parameters
represent a small sized long life thorium CANDLE HTGR with
thermal power of 30 MWth. The active core height is set to be
800 cm to obtain continuous operation for more than 25 years
before refueling operation is needed. However as mentioned
before, the core life is easily extended since it is proportional to
its active core height.
3. RESULTS AND DISCUSSIONS
3.1 Thorium CANDLE-HTGR without Pa-231
First we consider the critical and equilibrium condition of
thorium CANDLE HTGR without Pa-231 addition for U-233
enrichment of 15 and 10 w/o. The natural Gd burnable poison
concentrations needed for the two enrichments were found to
be 8.75 and 5.40 w/o, respectively. For U-233 enrichment of 15
w/o, the burning region moving velocity is around 22.74
cm/year so that with an active core height of 800 cm the core
life time can reach approximately 35 years. A considerably high
discharge burnup of 138 GWd/t can be attained. The axial
power peak is around 5.93 W/cm3. For U-233 enrichment of 10
w/o, the core life time is found to be about 25 years while the
attained discharge burnup is around 96 GWd/t. However, the
axial power peak is higher than the one with U-233 enrichment
of 15 w/o, i.e. 6.10 W/cm3. The axial power peaks for the two
cases are also the maximum power peaks of the whole core and
their locations are in the radial core center (r=0 cm). At the
core and graphite reflector radial boundaries (r=115 cm) we
also observed increased values of power density but still lower
than the peak in the radial core center. This can be understood
since we do not adopt different fresh fuel compositions in the
radial direction.
Several important nuclide distributions, thermal neutron
flux distribution, power density and kinf in the axial direction at
the radial core center (r=0 cm) are shown in Figs. 5 and 6. In
these figures, the left hand side of the axial position
corresponds to the fresh fuel region. It can be observed that the
kinf values are less than unity in the fresh fuel region while
larger than unity at the major part of the burning region. In the
vicinity of burning region, one can observe the thermal neutron
flux peaks which are responsible for the axial power density
peaks.
In the burning region where the neutron flux is high so
that the burnable poison concentration (represented in the
figures by Gd-157) depletes sharply as well as the fissile U-233
concentration. On the other hand, one can observe the U-234
and U-235 concentrations build up at the burning region and
finally saturate at the spent fuel region.
The buildup of Pa-233 concentrations is attributed to Th232 neutron absorptions but the concentrations rapidly decrease
because of its short half life (around 27 days).
Copyright © 2014-2015 by JSME
15 w/o U-233, 8.75 w/o Nat. Gd (No Pa-231)
10 w/o U-233, 5.40 Nat. Gad (No Pa-231)
8.0E+13
Pa-231
7.0E+13
Nuclide Density (atoms/cm3)
6.0E+13
U-232
1.0E+19
5.0E+13
U-233
4.0E+13
U-234
3.0E+13
U-235
1.0E+17
Gd-157
2.0E+13
1.0E+16
Thermal
Flux
1.0E+13
1.0E+15
100
200
300
400
500
600
700
1.0E+18
3.0E+13
U-235
2.0E+13
Gd-157
Thermal
Flux
100
200
300
400
500
600
700
800
Axial Position (cm)
(a)Nuclide densities (left axis) and thermal neutron flux
(right axis)
10 w/o U-233, 5.40 Nat. Gd (No Pa-231)
8.00
Power Density
Thermal Neutron Flux
7.00
1.30
Infinite Multiplication Factor
6.00
1.20
5.00
1.10
4.00
1.00
3.00
0.90
2.00
0.80
1.00
0.70
0.00
0.60
500
600
700
800
Axial Position (cm)
Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s)
Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s)
U-234
0.0E+00
0
1.40
400
4.0E+13
1.0E+13
15 w/o U-233, 8.75 w/o Nat. Gd (No Pa-231)
300
U-233
1.0E+16
800
8.00
200
U-232
5.0E+13
1.0E+17
(a)Nuclide densities (left axis) and thermal neutron flux
(right axis)
100
Pa-233
6.0E+13
1.0E+19
Axial Position (cm)
0
Pa-231
1.0E+20
1.0E+15
0.0E+00
0
8.0E+13
7.0E+13
Pa-233
1.0E+20
1.0E+18
1.0E+21
Nuclide Density (atoms/cm3)
1.0E+21
1.40
Power Density
Thermal Neutron Flux
7.00
1.30
Infinite Multiplication Factor
6.00
1.20
5.00
1.10
4.00
1.00
3.00
0.90
2.00
0.80
1.00
0.70
0.00
0.60
0
100
200
300
400
500
600
700
800
Axial Position (cm)
(b)Power density, thermal neutron flux (left axis) and kinf
(right axis)
(b)Power density, thermal neutron flux (left axis) and kinf
(right axis)
Fig. 5. Critical and equilibrium condition of thorium CANDLE
HTGR with U-233 enrichment of 15 w/o and burnable poison
concentration of 8.75 w/o (without Pa-231 addition).
Fig. 6. Critical and equilibrium condition of thorium CANDLE
HTGR with U-233 enrichment of 10 w/o and burnable poison
concentration of 5.40 w/o (without Pa-231 addition).
The decay of Pa-233 contributes to the production of the fissile
U-233. A very small amount of Pa-231 in the burning region
mostly from the (n,2n) reaction of Th-232 can be observed. Its
product after absorbing one neutron followed by a beta decay,
i.e. the U-232 can also be observed saturated in the spent fuel
region, however the amount is also small so that U-232
contribution for the fissile U-233 production via neutron
absorption is negligible.
The effects of Pa-231 addition on the axial power peak and
other burnup characteristics of the thorium CANDLE HTGR
are summarized in Table 3. For example, in the case of 15 w/o
enriched U-233 with 7.50 w/o Pa-231 concentration, the annual
requirement of Pa-231 is around 64 kg/year at nominal power.
Several important nuclide distributions, thermal neutron flux
distribution, power density and kinf in the axial direction at the
radial core center are shown in Fig. 7 for enrichment of 15 w/o
and Pa-231 weight fraction of 7.50 w/o, and in Fig. 8 for
enrichment of 10 w/o and Pa-231 weight fraction of 5.00 w/o.
From Table 3 one can observe that a significant improvement
on the maximum power density can be achieved by Pa-231
addition for both U-233 enrichments, i.e. a reduction of more
than 40%.
3.2 Effects of Pa-231 Addition for Flattening the Axial
Power Distribution
For U-233 enrichment of 15 w/o, three Pa-231
concentrations were evaluated, i.e. 2.50, 5.00 and 7.50 w/o,
while for enrichment of 10 w/o, one Pa-231 concentration was
evaluated (5.00 w/o).
Copyright © 2014-2015 by JSME
15 w/o U-233, 2.00 w/o Nat. Gd (7.5 w/o Pa-231 addition)
10 w/o U-233, 0.40 w/o Nat. Gd (5.00 w/o Pa-231 addition)
8.0E+13
Pa-231
7.0E+13
Nuclide Density (atoms/cm3)
6.0E+13
U-232
1.0E+19
5.0E+13
U-233
4.0E+13
U-234
3.0E+13
U-235
1.0E+17
Gd-157
2.0E+13
1.0E+16
Thermal
Flux
1.0E+13
1.0E+15
100
200
300
400
500
600
700
1.0E+18
U-235
2.0E+13
Gd-157
Thermal
Flux
100
200
300
400
500
600
700
800
(a)Nuclide densities (left axis) and thermal neutron flux
(right axis)
10 w/o U-233, 0.40 w/o Nat. Gd (5.00 w/o Pa-231 addition)
8.00
Thermal Neutron Flux
1.30
Infinite Multiplication Factor
6.00
1.20
5.00
1.10
4.00
1.00
3.00
0.90
2.00
0.80
1.00
0.70
0.00
0.60
500
600
700
800
Axial Position (cm)
Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s)
Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s)
3.0E+13
Axial Position (cm)
Power Density
400
U-234
0.0E+00
0
1.40
300
4.0E+13
1.0E+13
15 w/o U-233, 2.00 w/o Nat. Gd (7.5 w/o Pa-231 addition)
200
U-233
1.0E+16
800
8.00
100
U-232
5.0E+13
1.0E+17
(a)Nuclide densities (left axis) and thermal neutron flux
(right axis)
0
Pa-233
6.0E+13
1.0E+19
Axial Position (cm)
7.00
Pa-231
1.0E+20
1.0E+15
0.0E+00
0
8.0E+13
7.0E+13
Pa-233
1.0E+20
1.0E+18
1.0E+21
Nuclide Density (atoms/cm3)
1.0E+21
1.40
Power Density
Thermal Neutron Flux
7.00
1.30
Infinite Multiplication Factor
6.00
1.20
5.00
1.10
4.00
1.00
3.00
0.90
2.00
0.80
1.00
0.70
0.00
0.60
0
100
200
300
400
500
600
700
800
Axial Position (cm)
(b)Power density, thermal neutron flux (left axis) and kinf
(right axis)
(b)Power density, thermal neutron flux (left axis) and kinf
(right axis)
Fig. 7. Axial distribution of important nuclides and thermal
neutron flux for U-233 enrichment of 15 w/o and burnable
poison concentration of 2.00 w/o (with Pa-231 addition of 7.5
w/o).
Fig. 8. Axial distribution of important nuclides and thermal
neutron flux for U-233 enrichment of 10 w/o and burnable
poison concentration of 0.40 w/o (with Pa-231 addition of 5.0
w/o).
The addition of Pa-231 also contributes in enhancing the
thorium CANDLE burnup performance as can be observed
from the increased discharged burnup and slower velocity of
burning region movement. These promising effects of Pa-231
addition are discussed below. Pa-231, with a higher absorption
cross section relative to Th-232, is transmuted to Pa-232 by a
neutron absorption which is shortly followed by a decay
reaction (half-life of about 1.3 day) into U-232 (see Figs. 7 and
8). Since the fission cross section of U-232 is in the same order
with its absorption cross section in thermal energy region, its
fission rate and transmutation rate to the fissile U-233 are
comparable.
Table 3. Effects of Pa-231 addition on the maximum power
density and other burnup characteristics.
Burning
Max.
Nat.
Ave.
U-233 Pa-231
Region
Power
Gd
Burnup
(w/o)
(w/o)
Velocity
Density
(w/o)
(GWd/t)
(cm/year) (W/cm3)
15.0
0.00
8.75
138
22.7
5.93
2.50
5.50
149
20.4
5.86
5.00
4.00
142
21.0
4.79
7.50
2.00
149
19.6
3.57
10.0
0.00
5.40
95
31.8
6.10
5.00
0.40
97
29.7
3.26
Copyright © 2014-2015 by JSME
As discussed in the previous subsection, the fissile U-233 is
also bred from Th-232 via Pa-233 but presumably slower due to
the longer half-life of Pa-233 (26 days).
The Pa-231 addition relaxes the rapid depletion of U-233
especially in the burning region which in turn will widen the
full height maximum width of the moving burning wave. This
phenomenon is responsible for the decrease of the maximum
power density in the axial moving direction of the burning
wave as clearly shown in Figs. 7 and 8. From the figures, one
can also observe that the Pa-231 concentration depletion profile
resembles the burnable poison’s profile which indicates that Pa231 plays also the essential role as a burnable poison in the
CANDLE burnup phenomenon. However, although not shown
here, our parametric calculation results showed that it is
impossible to obtain an equilibrium yet critical thorium
CANDLE HTGR by eliminating the burnable poison (natural
Gd) completely (i.e. using only Pa-231). Hence, Pa-231 can
only reduce the amount of burnable poison needed for a
physically realizable thorium CANDLE HTGR. Within the U233 enrichment considered here (15 and 10 w/o), the Pa-231
addition is also contributing in slowing down the velocity of the
burning wave so that double fold benefits are achieved
simultaneously i.e. larger discharged burnup and longer core
life time for the same active core height.
From Figs. 7 and 8, one can also observe that the kinf of
the spent fuel are slightly greater than one. This indicates that a
criticality safety measure must be taken into account for
handling (storage and transportation) of the spent fuel.
4. CONCLUSIONS
The effects of Pa-231 addition on decreasing the
maximum power peak in the thorium CANDLE HTGRs have
been investigated for U-233 enrichment of 15 and 10 w/o. For
U-233 enrichment of 15 w/o, three Pa-231 concentrations were
evaluated, i.e. 2.50, 5.00 and 7.50 w/o, while for enrichment of
10 w/o, one Pa-231 concentration was evaluated (5.00 w/o).
Comparing with the original thorium CANDLE HTGR
(without Pa-231 addition), for all cases Pa-231 addition
decreases the maximum power density and at the same time
increases the discharge burnup and core life time. From the
numerical simulations in the present work, it is shown that Pa231 plays partly the role of a burnable poison in the early stage
and of a fertile fuel in the later stage of burnup. The U-233 bred
from Pa-231 contributes to the reactivity increase at the later
stage of burnup, i.e. broadening the burning region (wave)
which in turn lowers the axial power peaks.
[2] Ohoka Y. and Sekimoto H., “Application of CANDLE
Burnup to Block-Type High Temperature Gas Cooled
Reactor”, Nuclear Engineering Design, Vol. 229 No. 1, pp.
15-23 (2004).
[3] Ohoka Y., Watanabe T and Sekimoto H., “Simulation
Study on CANDLE Burnup Applied to Block-Type High
Temperature Gas Cooled Reactor”, Progress in Nuclear
Energy, Vol. 47, No. 1-4, pp. 292-299 (2005).
[4] Ismail, Ohoka Y., Liem P.H. and Sekimoto H., “Long Life
Small CANDLE-HTGRs with Thorium,” Annals of
Nuclear Energy 34, pp. 120-129 (2007).
[5] Ohoka Y. and Sekimoto H., “Application of CANDLE
Burnup to Block-Type High Temperature Gas Cooled
Reactor for Incinerating Weapon Grade Plutonium,”
Proceedings of GENES4/ANP2003, September 15-19,
2003, Kyoto, Japan.
[6] Liem P. H., Ismail and Sekimoto H., “Small High
Temperature Gas-Cooled Reactors with Innovative Nuclear
Burning,” Progress in Nuclear Energy Vol. 50, pp. 251-256
(2008).
[7] Tran H.N. and Liem P.H., “Neutronic feasibility study of
U-Th-Pa based high burnup fuel for pebble bed reactors”
Progress in Nuclear Energy Vol. 80, pp.17-23 (2015).
[8] Shmelev, A., Saito, M., Artisyuk, V., “Multi-component
self-consistent nuclear energy system: on proliferation
resistance aspect,” Proceeding of 2nd Annu. JNC Int.
Forum on the Peaceful Use of Nuclear Energy, Tokyo,
Japan, Feb. 21-22, (2000).
[9] Imamura, T., Saito, M., Yoshida, T., Artisyuk, V.,
“Potential of Pa for gas cooled long-life core,” J. Nucl. Sci.
Technol. Vol. 39, pp. 226-233, (2002) .
[10] Okumura K., Kugo T., Kaneko K. and Tsuchihashi K.,
“SRAC2006: A Comprehensive Neutronics Calculation
Code System,” JAEA-Data/Code 2007-004 (2007).
[11] Shibata K. Kawano T., Nakagawa T., Iwamoto O.,
Katakura J., Fukahori T. et al., “Japanese Evaluated
Nuclear Data Library Version 3 Revision-3: JENDL-3.3,”
Journal of Science and Technology 39, pp. 1125-1136
(2002).
ACKNOWLEDGMENTS
Some figures related to CANDLE burnup and cell
calculation model used in the manuscript were provided by Drs.
Ohoka Y. and Ismail.
REFERENCES
[1] Sekimoto H., Ryu K and Yoshimura Y., “CANDLE: The
New Burnup Strategy”, Nuclear Science and Engineering,
Vol. 139, No. 306 (2001).
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