Technology and Operation

Transcription

Technology and Operation
Gösgen nuclear power plant
Our aim is the safe, reliable and long-term
operation of our power plant.
High availability and cost efficiency are
achieved through mature technology, a prudent
mode of operation and the collective expertise
of our staff.
While output and efficiency have undergone a
steady increase over the years, safety has always
been our topmost priority and will continue to
remain so.
We intend to show that electricity generation
from nuclear power continues to be safe,
environmentally friendly and cost efficient
even under harsher competitive conditions.
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Plant layout
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Reactor building
Emergency feed building
Switchgear building
Reactor auxiliary building
Emergency diesel buildings
External system transformers 220 kV
Vent stack
Store for low and intermediate-level waste
Special emergency building
Cooling water intake structure
Cooling tower make-up water treatment building
Sludge depot
Setting pond for calcium precipitates
Sludge thickener
Service water pump house
Cooling tower and sound-absorbing wall
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Circulating water pump house
Turbine building
Block transformers 380 kV
Storage building
Garages and fire-brigade building
Workshop and spare parts store
Demineralising system building
Auxiliary boiler and heating station
Visitor centre
Training and simulator building
Staff restaurant
Entrance area
Administration building
Spent fuel storage building
Dry cooling towers
Entrance to underground carpark
Operating results of Gösgen nuclear power plant (KKG)
Year
Full power
Capacity
Electricity generated
Annual costs
Generating cost
hours
factor %
in bn kWh
in CHF million
in centimes/kWh
1980
6535.7
74.4
5.950
377.4
6.3
1985
7376.9
84.2
6.746
415.0
6.2
1990
7796.5
89.0
7.131
402.0
5.6
1995
8152.1
93.1
7.821
407.0
5.2
2000
8105.5
92.3
7.804
320.0
4.1
2005
7840.7
89.5
7.583
329.1
4.34
2006
8370.5
95.6
8.099
333.6
4.12
2007
8434.2
96.3
8.159
297.3
3.64
2008
8235.7
93.8
7.964
316.6
3.98
2009
8349.1
95.3
8.072
374.8
4.64
3 April 2007: The KKG attains the milestone of 200 billion kilowatt hours of electricity generated, as the first electricity power
plant in Switzerland to do so. This it achieved during a total of 217,000 hours of operation over a period of some 28 years.
KKG_techn_Broschuere_e_KKG_techn_Broschuere_e 09.06.11 11:19 Seite 1
Ta ble of Conte nts
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Contribution to Switzerland’s electricity supply
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Plant design and special technical features
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Reactor coolant system
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Auxiliary and secondary systems
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Safety precautions
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Steam and power conversion system
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Cooling water systems
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Station service power supply
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Operation and maintenance
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This brochure provides an overview of the
key technical features of the Gösgen nuclear
power plant (KKG). Nuclear heat generation is
treated as part of the overall system here.
Readers do not require any detailed expert
knowledge. The brochure is intended for
those with an interest in technical matters.
Environmental aspects
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2
Nuclear fuel cycle
52
2
Upgrading, retrofitting, modernisation
Kernkraftwerk Gösgen-Däniken AG (KKG)
4658 Däniken, www.kkg.ch
© KKG, 2011
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Contribution to Switzerland’s
electricity supply
The reactor building of the Gösgen nuclear power plant.
The move into nuclear power
Switzerland initially generated its electricity
solely on the basis of hydropower, since the
country has no fossil energy resources to tap.
With the economic boom that followed on
from the Second World War, demand for electricity rose rapidly in the 1950s, and the further expansion of hydropower soon came up
against its limits on both landscape and economic grounds.
While the electricity supply companies were
planning fossil-fired electricity generating
plants, the Swiss Federal Council opted for
the introduction of nuclear power at the start
of the 1960s. The decisive arguments in
favour of nuclear power were its low generation costs, the dependable supply and environmental protection. Clean nuclear energy
was to complement clean hydroelectric
power.
Planning work on the first nuclear power
plants commenced without delay, and the
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Annual output
1030
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1985
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Net production (billion kWh)
2000
940
2005
Nominal power (in MW)
Annual electricity generation has increased by 2 billion kWh since the
start of operation.
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Contribution to Sw itzerland’s electricity supply
first 350 MW nuclear power plant, Beznau 1,
was brought on stream in 1969 already. Several nuclear power plants were initially
planned in Switzerland, and five of these
were finally constructed at four sites. The five
reactors were connected to the grid between
1969 and 1984. With a combined net power
output of 3253 MW, these plants meet approximately 40 % of Switzerland’s electricity
requirements.
Operating results
Since commercial operation commenced in
November 1979, the Gösgen nuclear power
plant (KKG) has achieved higher than average levels of availability and operating
safety. In 1980, the KKG generated 5.9 billion
kWh electricity and, today, the figure is some
8 billion kWh annually, covering around
13 % of the country’s demand. Up to 31 December 2009, the KKG had generated a total of 222 billion kWh and achieved a high
Gösgen nuclear power plant at the southern
foothills of the Jura mountains.
Load diagram
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Jan.
Feb.
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Dec.
Planned outages for refuelling and annual maintenance are scheduled for midway through the year.
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Contribution to Switzerland’s electricity supply
average capacity factor of 90 %. Over the
years, the cost of generating electricity fell
from 6.3 centimes per kWh in 1980 to 4.64
centimes per kWh in 2009.
Both minor and major modifications have
been approved by shareholders, with the aim
of constantly improving on the operational
and safety parameters of the plant. These
have included advanced fuel management,
improvements to turbine efficiency and the
retrofitting of a pressure relief system for the
reactor coolant system. Together with reduced outage periods, this has contributed
significantly to the 15 % increase in net electricity generation that has been achieved
since the KKG was first brought into operation, corresponding to an extra 2 billion kWh
or so per year. During this time, the radiological releases to the environment and staff
radiation doses have been way below the
limits set by the authorities.
High safety standards, reliable operation,
low emission values, cost-efficiency and
also a permanent dialogue with the public
have all helped to ensure that the KKG is
readily accepted by the local community.
The population of the canton in which the
KKG is located and, more particularly, the
surrounding communities, made this very
Water vapour evaporating from the cooling
tower.
Emission of radioactive substances (annual dose in mSv)
1000
100
Average dose for the Swiss population due to natural occurrences, with fluctuation range
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Maximum allowable dose in the vicinity of the nuclear power plant due to its emissions
Threshold of importance according to the Swiss Radiation Protection Ordinance
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Actual dose in the vicinity due to the nuclear power plant
0.0001
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The radioactive releases are well below the authorised limits.
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Contribution to Sw itzerland’s electricity supply
clear in the four public referendums held
on nuclear power in 1979, 1984, 1990 and
2003. The KKG has approximately 500 employees, most of whom live in the direct
vicinity of the power plant. Additional temporary staff are taken on primarily during the
annual refuelling and maintenance outage.
how well a plant is operated and maintained.
A high availability means that only a few incidents have occurred and hence also constitutes a measure of reactor safety. The availability, taken together with the capacity
factor, is the most comprehensive characteristic value employed for assessing a power
plant. The KKG has continuously increased its
capacity factor over the years, maintaining it
at a high level. The value of 95.5 % for 2009
is much higher than the average value for
pressurised light water reactors (84.2 %).
Availability and capacity factor
Availability is the term used to describe the
ability of a plant to convert thermal power
into electricity independently of the actual
quantity generated. External events which
restrict power generation and are beyond
the operational management’s control do
not reduce the availability of the plant. The
capacity factor, by contrast, is a measure of
the actual use made of the plant. The availability is taken as an indicator of the performance capability and the reliability of a
plant – from both a technical and an economic point of view. It is also an indication of
%
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Capacity factor and availability of the overall plant
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Capacity factor
2003
2004
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Availability
A high capacity factor and availability are indicators of efficient plant operation and a good technical
state of the plant.
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Plant design and special
technical features
Reactor building and containment during the construction phase in 1976.
1973, the communes of Däniken and Gretzenbach had granted their approval of the
land-use zoning plan, and the Cantonal Government of Solothurn had issued the permits
and approvals required under the terms of
the water legislation. In February 1973, the
Kernkraftwerk Gösgen-Däniken AG operating
company was founded, and the decision to
start on the construction work was taken.
The KKG commissioned Kraftwerk Union AG
of Mülheim (now Areva NP) to construct the
turnkey power plant with a pressurised water
reactor. The site development work, construction supervision and other project work
was assigned to the former Motor-Columbus
Ingenieurunternehmungen AG. The initial site
development work was completed by summer 1973 already and was followed by soil removal, levelling and the lowering of the
groundwater level. In mid-December of that
same year the first concrete was poured for
the foundations of the reactor building.
Planning, construction
and commissioning
The fundamental investigations into the suitability of the site started in 1966 and, in May
1969, a consortium was set up to conduct the
initial project planning. Comprehensive geological, seismic, ecological and meteorological studies were conducted prior to selection
of the site. In 1970, the consortium filed an
application for the construction of a nuclear
power plant with river-water cooling. In order
to keep the thermal pollution of the rivers
Aare and Rhine to a minimum, the Swiss Federal Council took the decision in March 1971
to allow only closed-circuit cooling for all future nuclear power plants. This made it necessary to reconfigure the project for coolingtower operation rather than river-water
cooling.
In 1972, the Swiss Federal Department of
Transport and Energy (now UVEK) issued its
approval for the site. By the beginning of
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Plant d esign and specia l technical features
The first self-sustained chain reaction was
initiated on 19 January 1979. Then, on 6 February 1979, the first Swiss nuclear power plant
in the 1000 MW category fed electricity into
the grid for the very first time. The commissioning trials were delayed, however, due to
an incident at the US Three Mile Island nuclear power plant near Harrisburg, since the
Swiss Federal Council demanded a check on
the safety systems and operating regulations.
Following the successful completion of the
commissioning trials, the KKG commenced
commercial operation in November 1979 with
a gross electrical power output of 970 MW.
On 20 December, a start was made on supplying process steam to a cardboard factory
in Niedergösgen. This steam supply was the
largest of its kind from a European nuclear
power plant.
crease in the nominal thermal power from
2808 MW to 3002 MW was performed in several stages. It was achieved, in particular, by
extending the active fuel length inside the
fuel rods and improving the corrosion resistance of the cladding. With these modifications, the plant was able to operate with the
maximum authorised thermal power of 3002
MW from July 1992 onwards, resulting in a
gross electrical power of 990 MW.
Further electrical power increases were
achieved in two stages in 1994 and 1995
solely by improving the efficiency of the turbine system. The modernisation of the lowpressure turbines resulted in more efficient
use of the thermal energy in the reactor and,
as of 1 January 1996, the gross electrical
power was stepped up to 1020 MW. This
marked the most extensive retrofitting programme since the plant’s start-up and led to
an extra 300 million kWh electricity being
generated per year, corresponding to the production of a medium-sized Swiss run-of-river
hydropower plant.
This scheduled increase in the thermal and
electrical power of the KKG was in line with
the targets of Switzerland’s «Energy 2000»
programme, which provided for a 10 % increase in power from the country’s existing
Power increase
The experience acquired during the first few
years of operation showed that the plant still
had considerable power reserves and hence,
in May 1985, an application was submitted
for the gross thermal power to be increased
by 7 %. The Federal Council granted the necessary approvals in December 1985. The in-
Schematic diagram of a pressurised water reactor
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Reactor
Steam generator
Reactor coolant pump
Pressuriser
High-pressure turbine
Water separator
Superheater
Low-pressure turbine
Condenser
Condensate pump
Low-pressure preheater
Feedwater tank
Feedwater pump
High-pressure preheater
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nuclear power plants. Since the turn of the
millennium, a large number of modernisation and retrofitting measures have considerably increased the efficiency of the plant
and hence directly influenced its generating
power. These measures have included the
optimisation of the turbines and the reheaters, the fitting of additional water separators and the replacement of the evaporators in the cooling tower. As of 1 January 2010,
the gross electrical power was increased to
1035 MW.
Sections through the reactor building
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Plant location and site layout
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The KKG is located at the edge of the southern foothills of the Swiss Jura, about halfway
between the towns of Olten and Aarau and
close to major consumers in the northern
Swiss lowlands. The plant is in a loop of the
river Aare and covers an area of 14 hectares.
The area belongs to the municipality of
Däniken in the canton of Solothurn. Approximately 300 metres to the east of the site is
the 380 kV high-voltage switchyard, one of
the most important junction points in the
Swiss high-voltage grid.
The site area was filled and raised to protect
the plant from flooding. It is now 382 m
above sea level and hence at least one metre above the highest water level that can be
expected in the river Aare. The ground under
the plant consists of a 20 to 30-m thick layer
of gravel, on a solid limestone formation,
which provides a stable basis for the plant.
The KKG is located in an area of low seismic
activity. When the site was selected, it was
not only the load-bearing capability of the
ground that was crucial, but also its closeness to the grid, its proximity to the river Aare
for the cooling water supply and the ease of
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Ground plan
+ 18.40 m
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Reactor
Steam generator
Pressuriser
Pressuriser relief tank
Accumulator
Borated water storage tank
Personnel lock
Fuel storage pool
Fuel assembly transfer
equipment
Cask loading pool
Refuelling machine
Delay bed
Access shaft
Store for new fuel
Emergency lock
Reactor service floor
Storage space for reactor
closure head
Ventilation system
Main steam and feedwater
valve compartment
Main steam and feedwater
valve
Exhaust silencer
Polar crane
Steel containment
Annulus
Surge tank for component
cooling system
Residual heat removal pump
Safety injection pump
Equipment hatch
Access door
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+ 12.00 m
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Cross-section
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access for heavy goods vehicles. A direct connection to the railway system facilitates the
transport of heavy loads to and from the site.
When positioning the various buildings and
plant facilities, care was taken to ensure a
functional and space-saving arrangement.
There is a clear spatial divide between the nuclear and the conventional parts of the plant,
confining the radioactive systems to a welldefined, specially controlled area. The easy
access to buildings, systems and components is also an advantage for maintenance
work.
The compact arrangement of the buildings
ensures short pipe and cable connections
between the individual sections of the plant.
The cable ducts and piping for redundant
safety-relevant systems are always fed into
the buildings separately. The arrangement of
the turbine hall and the reactor building ensures a short energy path from the reactor to
the transformers, which are located on the
eastern side of the turbine building. Electricity is transferred from the transformers to the
380 kV switchyard via an overhead power
line.
environment against radiological impacts
from postulated severe incidents. It prevents
the uncontrolled release of radioactive material to the outside.
The containment is in an off-centre position
inside the reactor building, which has an
outer shell in reinforced concrete. The containment, together with the reactor building,
forms a twofold safety casing. The reactor
building protects the radioactive plant components from external impacts; it is designed
to withstand earthquakes, shockwaves from
explosions and aircraft crashes.
The pressure-resistant containment with gastight welds is embedded in a shell-shaped
foundation ring at the base but is otherwise
designed as a self-supporting structure.
When the containment was designed as a
fully pressurised structure, it was assumed
that a reactor coolant pipe could burst, with
the full water content of the reactor coolant
system and also one of the steam generators
evaporating. The steel shell is thus designed
to withstand an overpressure of 4.89 bar at
a temperature of 135°C for such a case.
Access to the containment is through a pressure-resistant and gas-tight lock.
The reactor auxiliary building houses the processing facilities for waste water, concentrates and waste gases, the central air supply
and extraction system for the controlled area,
workshop facilities, laboratories for the analysis of radioactive materials, decontamination facilities and also storage for low and intermediate radioactive waste. In June 2007,
after building work lasting 20 months, a
three-storey extension was completed, providing an additional 8000 cubic metres for
workshops and storage. This extra space has
allowed the storage of materials to be optimised and fire protection to be improved.
Controlled area
The nuclear section of the plant comprises
the reactor building, the reactor auxiliary
building and the external spent fuel storage
building, completed in 2008, which together
form a closed controlled area. Access to this
controlled area is via a central guarded entrance.
The reactor’s spent fuel storage pool, together
with the plant components containing radioactivity that are at reactor operating pressure, are enclosed by a spherical steel shell.
This safety barrier (containment) protects the
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The extension was designed as an autonomous building with a dilatation gap (air
gap) separating it from the reactor auxiliary
building. This ensures that the dynamic behaviour of the reactor auxiliary building remains unchanged in the event of an earthquake. To achieve this separation, the narrow
extension had to be anchored in the ground
with 54 tension and compression piles so as
to secure it against tilting during an earthquake. These piles are 13 m long and 1.3 m in
diameter. In order to eliminate the seismic
forces, the 2 m thick foundation plate contains a massive 280 kg of reinforcement per
cubic meter concrete, which is roughly five
times the amount of steel used in a conventional building. This explains why 700 tons of
steel were required to construct the extension
building.
On 8 April 2008, the supervisory authorities
issued the operating permit for an external
spent fuel storage building. Since there was
no space in the reactor building to extend the
internal spent fuel storage pool, a new storage building, serving this same purpose, was
built outside the existing building, to the
north-west of the vent stack, in the direct
proximity of the reactor auxiliary building.
This new building comprises a tract for all the
control systems with a skywalk to the reactor
auxiliary building and two dry cooling towers.
The internal structures of the building are
separated from the exterior walls, and the
spent fuel storage pool is protected against
tremors by springs and damping elements.
The building in reinforced concrete is 37 m
long, 17 m wide and 25 m high. The outer
structures of the spent fuel storage building
are at least 1.5 m thick. This ensures that the
building is protected against exceptional
events, such as earthquakes, flooding and an
aircraft crash. The spent fuel is brought into
the building in spent-fuel transport casks via
the onsite railway system. In its final configuration, the storage pool in the building will
hold up to 1008 spent fuel assemblies. This
External spent fuel storage building.
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pool then increases the storage capacity of
the storage pool inside the reactor building,
which holds approximately 600 spent fuel
assemblies.
The pool cooling system comprises four symmetrically arranged independent trains with
two trains being connected to a cooling tower
in each case. The heat from the spent fuel assemblies is eliminated into the outside air via
an intermediate cooling circuit with natural
circulation. This involves the intermediate
coolant flowing through heat exchangers that
are hung inside the storage pool. The heat is
then dissipated to the outside air by natural
circulation via water/air heat exchangers.
With very high outside temperatures and a
large spent fuel inventory in the storage pool,
fans are available to boost the air circulation
in the cooling towers. The first spent fuel assemblies were loaded into the new storage
pool in mid-May 2008.
Fuel assemblies being replaced during the
annual maintenance outage.
building. The decay heat is eliminated via a
dedicated cooling system connected up to
the compact storage pool. The spent fuel assemblies can remain in interim storage in
the compact storage pool for a period of several years.
The sickle-shaped annular space between
the outer reactor building shell and the containment serves to house and protect the
loading and transfer pool, the access shaft,
the emergency and regular cooling system,
the fresh fuel store and the waste gas delay
bed. The spent fuel assemblies are loaded
into the transport casks in the loading and
transfer pool. They are moved from the compact storage pool to the loading and transfer
pool by a remote-controlled transfer facility to
this end. The transport casks are moved into
and out of the annular space via the access
shaft.
Refuelling
Once a year the power plant is shut down for
refuelling. It takes some two to three weeks to
discharge the spent fuel assemblies, reposition the assemblies remaining in the reactor
core, load the fresh fuel assemblies and to
carry out inspection and maintenance work.
The fuel assemblies discharged from the reactor core are first placed in high-density
racks in the spent-fuel reactor storage pool.
There are more than 600 storage positions in
this pool, which can take not only spent fuel
assemblies but also instrumentation thimble tubes, control elements and tools. In the
compact storage pool, the radiation and decay heat are allowed to subside before the
fuel assemblies are conveyed in special
transport casks to the spent fuel storage
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Reactor well with reactor core opened, set-down platform with upper internals, and spent fuel storage pool.
The Gösgen pressurised water reactor is licensed to operate at a nominal thermal
power of 3002 MW. It has an operating pressure of 154 bar and an average operating
temperature of 308°C. The reactor coolant
system comprises the reactor, the pressuriser
system and three parallel circulation loops.
Each of these three identical loops consists of
a steam generator, a reactor coolant pump
and the connecting pipework.
into an upper and a lower part secures the reactor core inside the pressure vessel. The
lower internal structure, along with the core
grid and a core barrel, positions the core in
such a way as to ensure an even flow of
coolant through the core as a whole. The
shroud on the lower part of the core support
structure which is hung in the reactor pressure vessel also serves as a shield to protect
the reactor pressure vessel against neutron irradiation.
The coolant enters the reactor through three
inlet nozzles at a temperature of 292°C and
flows down through the annular gap between the core barrel and the reactor pressure vessel. At the semi-spherical base of the
reactor vessel, the coolant flow is deflected
through 180°. As it flows up through the reactor core, the coolant heats up to 325°C.
The heat is then transferred to the three
steam generators through the three outlet
nozzles. The coolant flows through the core
at an overall rate of 53,000 tons per hour and
Reactor pressure vessel
The reactor pressure vessel that houses the
reactor core is made of a fine-grained low-alloy steel, which combines a high weld quality with ductility, plus a low susceptibility to
embrittlement under neutron irradiation. The
removable reactor vessel head is fastened
on by 52 pre-tensioned bolts. The nozzles
for the control rod drive systems and core instrumentation are located at the top, on the
vessel head. A core support structure split
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Reac tor coola nt syste m
the rods is 3550 mm. The fuel rods are fixed
in position by spacers. The design of the fuel
assemblies, with open sides, promotes the
transverse mixing of the coolant and thus
ensures more uniform heating. There are
more than 36,000 fuel rods in the reactor
core, equivalent to a fuel pellet column length
of around 130 kilometres.
Control rod drive mechanism
Control rod guide assembly
Control assemblies
Upper core support
Coolant outlet
The reactor power is controlled by neutron
absorbers. Short-term control is performed by
means of control rods which govern the neutron flux and hence the reactor power. Fortyeight of the 177 fuel assemblies inside the reactor core are equipped with control
assemblies, each of which comprises 20 control rods, in addition to the 205 fuel rods.
Each fuel assembly has 20 free fuel rod positions which are used for guide thimbles.
For those fuel assemblies in positions without
control rods, some of the guide thimbles are
used for the core instrumentation probes.
These monitor the power density distribution within the core.
The control rods are activated by electromagnetic ratchet jack drive units which are located on the pressure vessel closure head. To
adjust the reactor power, the control rods
can be moved into the reactor core to a
greater or lesser depth. To achieve a fast reactor shutdown, all the rods are fully inserted
into the reactor core. This is done by removing the current from the electromagnetic restraining coils.
Support column
Grid plate
Fuel assembly
Pressure vessel
Core shroud
Core barrel
Lower core support
Flow skirt
is equally distributed over the three circulation loops.
Fuel assemblies
The reactor core comprises 177 tightlypacked, identical fuel assemblies. Each fuel
assembly has an array of 15 by 15 (i.e. 225)
possible fuel rod positions, 205 of which are
occupied. Inside each fuel rod, a column of
fuel pellets is enclosed in a gas-tight and
pressure-resistant-welded Zircaloy cladding
tube. The fuel pellets are made of either sintered uranium dioxide (UO2) containing enriched fissile uranium-235 or a mixture of uranium dioxide (UO2) and plutonium dioxide
(PuO2). The height of the fuel pellet stacks in
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Steam generators
Steam generator
The three steam generators transfer the heat
from the reactor coolant to the water/steam
circuit. Designed as standing U-tube heat exchangers, they convert feed water into live
steam to drive the turbo-generator system.
The collection chamber is connected to the
reactor coolant circuit pipes via inlet and outlet nozzles. The reactor coolant flows out of
the collection chamber and through the Utubes to the outlet chamber, giving off heat as
it flows. From the outlet chamber, the coolant
is directed to the primary coolant pump. The
bundle of U-tubes, made of an exceptionally
corrosion-resistant material, is supported at
a large number of points and is rolled into
and welded onto the steam generator’s tube
plate.
The incoming feed water flows downwards by
natural circulation between the vessel wall
and a shroud surrounding the tube bundle
before moving upwards again, giving off
steam, once it has absorbed the heat. In the
steam dome above the tube plate, the residual steam moisture is separated off before
the dried steam is eliminated through the
outlet nozzle.
Steam dryers
Manhole
Steam separators
Feedwater inlet nozzle
Feedwater ring line
Heating tubes
Shroud
Vessel
Tube support grids
Hand hole
Support and guide brackets
Tube sheet
Reactor coolant inlet
Reactor coolant outlet
suriser surge line. Pressure control is performed by means of an electric heater in the
water section of the pressuriser and a facility for spraying water into the steam section.
Using the spray system, the steam can be
condensed and hence the pressure reduced.
By generating heat with the electric heater
rods, water can be evaporated and hence
the pressure raised.
Pressuriser
The purpose of the pressuriser is to keep
the operating pressure in the reactor coolant
system constant. A change in the reactor
power produces variations in the temperature and volume and, without a pressuriser,
these would lead to pressure fluctuations.
The pressuriser is an upright container with
a capacity of 42 cubic metres, which is partly
filled with water. It is connected up to one of
the three reactor coolant loops via the pres-
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Steam outlet
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Reac tor coola nt syste m
Reactor coolant pumps and
pipework
Reactor coolant
system
The heated coolant flows from the reactor
pressure vessel through the coolant pipes to
the three steam generators. The reactor
coolant pumps transport the cooled-down
coolant back to the reactor pressure vessel.
The reactor coolant pumps are vertical, single-stage, centrifugal pumps with an overhung-mounted impeller. The key components
Pressure vessel
Steam generators
Coolant pumps
Pressuriser
of the pumps are a spherical pump casing, an
impeller mounted on the drive shaft and a
two-part diffuser, which is screwed onto the
pump casing. The pump casing is welded to
the reactor coolant piping. The drive motor is
a high-voltage, asynchronous motor of conventional design.
The seals on the reactor coolant pumps are
made up of a three-stage hydrodynamic end
face seal and a non-return seal. This latter
seal takes over the sealing function if the upstream seals fail. In the hydrodynamic sealing
system, which was fitted in 2008, pressure relief takes place via the three seals. 40 % of
the pressure is eliminated at each of the first
two stages and the remaining 20 % at the
third seal. Each stage is designed to withstand the full pressure differential.
Motor flange
Motor lantern
Axial-radial bearing
Shaft coupling
Seal housing
Radial bearing
Diffuser
Impeller
Pump casing
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! ! !
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! KKG_techn_Broschuere_e_KKG_techn_Broschuere_e 09.06.11 11:20 Seite 17
Auxiliar y and secondar y systems
Coolant treatment systems
stances in the reactor coolant system as low
as possible, corrosion and fission products
are removed. Coolant purification is performed by mixed bed filters filled with two different ion exchange resins. After purification,
the coolant can be degassed. Coolant degassing constitutes an additional purification step.
In the coolant storage system the coolant is
separated by evaporator systems into boric
acid and demineralised water (fully desalinated and degassed water). Boric acid and
demineralised water are used for controlling
reactivity.
All the chemical substances needed to treat
the coolant are prepared in the chemical control and feed system. This same system is
The volume control system links the highpressure reactor coolant system to the lowpressure auxiliary and secondary systems.
The reactor coolant system is filled and
drained by means of the volume control system. The volume control system offsets the
temperature-related volume fluctuations in
the reactor coolant which occur during reactor start-up and shutdown and with reactor
load changes. It also supplies sealant water
to the high-pressure shaft seals on the reactor coolant pumps.
Some 30 tons of coolant per hour are taken
out of the reactor’s primary circuit for purification. To keep the level of radioactive sub-
Chemical and volume control and waste processing systems
Fuel pool
purification system
Vent stack
Coolant
purification system
Coolant
degassing system
Volume control
system
Seal water
supply system
Gaseous waste
processing system
Coolant
treatment system
Coolant
storage system
Boric acid and demineralised water control system
Chemical
control system
Demineralised water
Chemicals
Reactor building drains
Water from laundry and showers
Concentrate
processing system
Drum store
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River
Liquid waste processing system
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also used to feed the chemicals into the
coolant, via the boric and demineralised water control unit. The appropriate amounts of
coolant are removed and conveyed to the
coolant storage tanks for interim storage. If
the boron content of the coolant needs to be
increased, then boric acid will be fed in. In the
reverse case, the boron content will be reduced by adding fully demineralised water. A
total of six tanks are available for coolant
storage, each with a capacity of 100 cubic
metres.
Systems for residual heat
removal, emergency cooling
and pool cooling
The systems for residual heat removal have
both operational and safety-related functions. Following a routine shutdown of the reactor, they take over the cooling of the reactor core, while, in a loss-of-coolant incident,
they ensure the emergency cooling of the
core. Additionally, these same systems are
used to cool the spent fuel storage pool.
During reactor shutdown the decay heat is
initially dissipated by the steam generators.
Later on, the residual heat removal system
takes charge of reducing the temperature still
further.
In each coolant loop, the heat absorbed is released into the headwater channel of the
river Aare via a cooling train which contains
an intermediate cooling circuit. This latter circuit forms the barrier between the reactor
coolant and the river water.
Two pool cooling lines are available for cooling the spent fuel storage pool, which are
connected up to the residual heat removal
system. There is also a further cooling line
which is independent of the residual heat removal system.
The efficiency of the residual heat removal
system means that the reactor can be cooled
down within just a few hours. The residual
heat removal pumps suck coolant out of the
coolant pipes leading away from the reactor
and feed the coolant, via the residual heat exchangers, into the pipes leading back to the
reactor coolant system.
In a loss-of-coolant incident, the residual heat
removal system has to ensure that the reactor core remains flooded, irrespective of the
Accumulators (pressurised storage tanks) for
emergency cooling water.
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leakage rate, and must also take charge of
the long-term elimination of heat from the reactor pressure vessel. The system is designed
in such a way that, even in the event of a reactor coolant line suffering a complete fracture, the reactor core will remain covered
with borated water, and sufficient cooling
will be provided.
Borated emergency cooling water is stored in
six accumulators, which are connected up to
the three reactor coolant loops via pipes with
check valves. If a major leak occurs, and the
reactor coolant pressure falls below the pressure in the accumulators, the accumulators
will empty their contents into the reactor
pressure vessel via the reactor coolant lines.
Once the pressure in the reactor coolant system drops below 10 bar, the low pressure
feed system starts up, and the residual heat
removal pumps deliver borated water from
the four storage tanks to the coolant loops via
separate feed lines.
In the event of a small or medium-sized leak
with a gradual reduction in pressure, the safety
injection pumps in the high-pressure safety injection system will start up first of all. These
feed in borated water from the storage tanks
until such time as the pressure has dropped far
enough for the system to automatically switch
over to the low-pressure feed-in.
The water fed into the reactor core first fills up
the reactor pressure vessel and then flows
through the fracture into the lowest point of
the containment, the so-called sump. Once
all the borated water from the storage tanks
and accumulators has been fed in, the water
in the containment sump is sucked out by the
heat removal pumps and conveyed back into
the reactor pressure vessel via the residual
heat exchangers.
Both the low-pressure and the high-pressure
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safety injection systems consist of three completely independent feed lines, with each line
being allocated to a coolant loop. In addition,
there is also a back-up line, which can be
brought in for each of the other three lines. A
single feed line is sufficient to control a lossof-coolant incident. As all the instrumentation
is connected to the emergency power supply,
the operational availability and functionality
of the emergency and residual heat removal
systems is maintained even under the most
extreme conditions.
Ventilation systems
Inside the controlled zone, supply-air, wasteair and air-conditioning systems take care of
ventilation, heating and cooling. Ventilation
in the plant and operational areas is performed primarily in circulation mode. In normal operation, only about 1000 cubic metres
of air per hour is fed into and extracted from
the containment. The small quantities of supply and waste air mean that the air ducts
into the containment require only a small
cross-section. In the plant areas that house
the reactor coolant system, any impurities in
the room air can be retained by the bypass
flow filters in the ventilation system.
The ventilation systems keep the pressure
constantly below that of the operational areas and the outer atmosphere thus always
ensuring a flow of air from areas with a low
level of radioactivity to those with a potentially higher level. This tiered low-pressure
system prevents any transfer of contaminated
air from the plant areas to the service compartment areas.
The air that is sucked out of the containment
in order to maintain the low pressure is
cleaned in the exhaust air unit before being
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released through the vent stack. The aerosol
and iodine filters used have a separation efficiency of more than 99 %. The specific activity of the waste air is monitored at the vent
stack. Instruments measure the aerosol,
iodine, and noble gas as well as the carbon14 activity. In conjunction with the airflow
measurements, this then makes it possible to
monitor the overall activity released.
form of diluted and suspended substances.
The activity of waste water from the laboratories, laundry and showers, by contrast, is
much lower.
The waste water streams are collected in separate groups of storage containers as a function of their origin. The purification is performed in evaporators. The distillate is stored
in monitoring tanks and checked by sampling. Once it is sufficiently purified, it can be
released into the headwater channel of the
river Aare, with a record being kept of the activity and quantity.
The evaporator concentrates that retain the
radioactivity of the waste water are collected
in concentrate tanks and stored there until
they are solidified. The retention factor for the
radioactivity is up to 99.9999 %. Each year,
about 7000 cubic metres of waste water are
generated, resulting in only 15 cubic metres
of concentrates requiring further processing.
Nuclear off-gas system
A further contribution to the air released
through the vent stack comes from the offgas system. Not all the fission products present in the form of noble gases dissipating
from the coolant can be retained by the
mixed bed filters in the coolant purification
system. This mainly involves the noble gases
of xenon and krypton. Effective removal of
these gases can be achieved through the
coolant degasification system. By evaporating and subsequently condensing the
coolant, these gases can be eliminated and
conveyed to the nuclear off-gas system.
The off-gas-system compressor circulates a
permanent flow of purge gas. Part of the
purge gas is directed over a bed of active
charcoal, where the noble gases are retained
until such time as their activity has largely decayed.
Waste processing and storage
All the radioactive waste generated during
operation of the power plant is processed in
such a way that it can be handled and stored.
This waste includes ion exchange resins, filters and filter residues, concentrates from
the waste-water evaporators, cleaning materials and items of clothing. With the exception
of the ion exchange resins from the reactor
coolant cleaning unit, the operational waste
normally only has a low level of activity.
The ion exchange resins and evaporator concentrates are dried and then embedded in bitumen in standard 200-litre drums before
being placed in the on-site interim store.
Combustible waste and small pieces of metal
can be processed in the plasma furnace at
the Central Interim Storage Facility for Ra-
Waste water processing facility
The purpose of the facility for processing radioactive waste water is to collect and purify
all the waste water that results within the
controlled area.
Waste water from the reactor coolant system
and the nuclear auxiliary and secondary systems can have a high specific activity in the
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Treatment of liquid radioactive waste
Tanks for chemicals
Waste water
Liquid waste tanks
Sulphuric acid
Antifoaming agent
Complexing agent
Sodium hydroxide
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1
7
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2
Evaporator
1 Recirculating pump
2 Evaporator feed pump
3 Recirculating and
demineraliser feed pump
4 Discharge pump
5 Sludge pump
6 Concentrate pump
7 Chemical feed pump
5
Evaporator
Sludge
Distillate
Monitoring tanks
Mixed-bed
filter
3
Discharge to
headwater channel
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6
Concentrate tanks
Bituminisation plant
dioactive Waste (ZZL) in Würenlingen, with
the resultant slag being immobilised in glass.
Filter cartridges and metals with intermediate
activity are embedded in concrete or placed
in massive shielded casks. No further processing is required prior to their final disposal at a later date.
On average, about 50 litres of operational
waste is generated per day at the KKG which
is in a form suitable for final disposal. Intermediate-level waste amounts to about 20
drums per year, and low-activity waste to
about 60 drums. Contaminated plant components and tools that are reusable are decontaminated.
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The KKG has two separate underground storage facilities for waste. The storage facility for
low-level waste has a capacity of 4300 drums
while the one for intermediate-level waste
holds 600 drums. If necessary, waste drums
can also be stored in the ZZL. Up to the end
of 2008 approximately 1000 drums with low
and intermediate-level waste had been transferred to the ZZL, all safely conditioned for
storage and final disposal.
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Safety precautions
Special emergency building.
Inherent safety
The highest priority in reactor safety technology is the safe enclosure of all the radioactive fission products that are generated
during nuclear fission. The safety measures
must be designed to ensure that, during both
normal operation and incidents, no radioactivity is released from the plant in an uncontrolled manner which could present a danger
to people or the environment.
Preventing the occurrence of incidents is similarly a priority. Administrative and structural
measures must be in place to detect malfunctions at an early stage and to eliminate
these, or at least restrict their impact and ensure that they do not escalate into an incident
which could affect the environment. Effective safety precautions allow for the possibility of faults and dysfunctions in both people and materials. Systematic precautions
thus require a fault-tolerant technical plant
design with sufficiently large contingency reserves to cope with any incidents too.
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In a light water reactor like the KKG, light water, i.e. normal, purified and demineralised
water, is used as the moderator and coolant.
The coolant water moderates the neutrons
generated by nuclear fission; it decelerates
high energy neutrons emitted from the fuel to
the «thermal velocity» at which they can trigger nuclear fission again.
The so-called inherent safety is based on the
properties of the moderator and the fuel. If
the coolant temperature increases and steam
bubbles form, then the density of the water is
reduced and fewer neutrons are decelerated.
At the same time, when the fuel temperature rises, more neutrons are absorbed by the
fuel-carrier material, uranium-238, and hence
fewer neutrons are available to trigger nuclear fission once again. Assuming a loss-ofcoolant incident caused by a major leak, the
chain reaction would immediately come to a
standstill, both through the increased neu-
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Safety precautions
쐍 safety level 4 (measures to limit the consequences of extremely improbable postulated accident scenarios).
Special requirements apply to the design and
operation of the plant (safety level 1), including:
쐍 sufficient safety margins in the design of
the systems and components
쐍 careful selection of the materials and comprehensive material testing
쐍 comprehensive quality assurance during
manufacture, installation and commissioning
쐍 system and component designs geared to
easy maintenance
쐍 a high level of redundancy in the safety-related components
쐍 a high degree of automation to reduce the
possibility of human error
쐍 a prudent mode of operation
쐍 regular repeat tests and inspections
쐍 permanent monitoring of key process parameters
쐍 automatic triggering of counter-measures
once predefined limits are attained
쐍 systematic recording, evaluation and
safety-related exploitation of our own and
external operating experience
쐍 comprehensive and continuous training of
operating staff
tron absorption with a higher fuel temperature and through the lack of a moderating effect due to steam bubble formation inside
the reactor core.
Safety principles
The nuclear safety of modern light water reactors like the KKG is based on a concept of
graded safety precautions, the «defence in
depth» concept. A distinction is drawn between
쐍 safety level 1 (measures for avoiding operational disturbances and incidents)
쐍 safety level 2 (measures to limit the impact of disturbances and incidents (anomalous operating conditions) that may occur
despite the precautions and prevent the
occurrence of accidents)
쐍 safety level 3 (measures to limit the consequences of accidents within specified
limits preventing unacceptable radiological
releases)
Safety barriers
To master anomalous operating conditions
(safety level 2), the systems have been designed on the basis of special safety principles. Special limiting devices and protection
systems for the plant equipment ensure that
disturbances of commercial operation have
only limited consequences. This is achieved
by either lowering the reactor power or, in the
case of a defective component, by switching
over to a standby sub-assembly. The use of
Nuclear fuel matrix
Cladding tube
Reactor vessel
and reactor coolant loop
Concrete shield
Reactor containment (steel shell)
Reactor building
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Sa fety precautions
limiting devices ensures that reactor scrams
can be avoided. Each reactor scram that is
avoided saves wear and tear on the plant.
accidents that could occur when highly inflammable or explosive substances are handled, as well as for the outbreak of fire on site.
Passive and active safety systems are available for controlling incidents. The passive
systems take effect by simply being there –
such as the many different barriers in concrete and steel which ensure reliable containment of the radioactivity and provide
shielding against direct radiation from the
reactor core. Also in this category are the accumulators for the emergency core cooling
system, which do not first have to be activated if they need to be brought into operation. The active safety systems implement
the actions triggered by the reactor protection system by means of actuators and a
range of different units. They require a trigger
signal and a power supply. The active safety
measures are the emergency and residual
heat removal system, the emergency feed
system, the emergency diesel generators and
the special emergency system.
Mastering accidents
Steam
Steam
Steam
Special safety systems are available for controlling accidents (safety level 3). These ensure
that the reactor can be shut down at any time
if necessary and that the decay heat still generated after shutdown will be eliminated. The
incidents that the plant must be able to master are so-called design-basis incidents. These
include the fracture of one of the reactor
coolant pipes, a live steam or feedwater pipe,
or the rupture of a steam-generator steel tube.
Accidents caused by external impacts are
also taken into account for plant design, and
the power station is protected against natural events such as earthquakes, storms, lightning and flooding, as well as against manmade occurrences including sabotage and
aircraft crashes. Allowance is also made for
Emergency cooling and residual
heat removal systems
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3
Reactor
Steam generator
Containment
Reactor building
Accumulator
Borated-water storage tank
High-pressure injection pump
Low-pressure injection pump
Heat exchanger
Containment sump
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7
10
8
9
8
Redundancy 1
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7
10
9
8
Redundancy 2
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7
10
9
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Redundancy 3
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9
Redundancy 4
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Safety precautions
The single-failure criterion applies for all the
safety systems in the event of an accident. A
malfunction in an individual component, subsystem or system may not impair the functional integrity of the system as a whole.
To ensure a basic level of technical safety, all
systems of particular importance are installed
in duplicate or more. This redundancy principle applies for all safety-relevant systems.
These include the emergency and residual
heat removal system, the nuclear component
cooling circuits, the emergency feedwater
system, the service cooling water system, the
cold water system and the containment isolation. The emergency and residual heat removal systems, for example, essentially consist of three identical injection lines, each of
which has two accumulators, one high-pressure safety injection pump, one low pressure
safety injection pump, one residual heat exchanger and one borated water storage tank.
Each of the three lines meets all the required
safety functions by itself. In addition, there is
a reserve line which is connected to each of
the other three lines. This multiple arrangement ensures adequate availability of the
overall system both when repairs or maintenance work are carried out and in the event
of a malfunction in a sub-system. The KKG
also has a special, duplicate emergency system which guarantees that the plant can be
shut down safely in the event of extreme external incidents and even in the case of a
postulated terrorist attack, such as a deliberate aircraft strike. A reliable supply to the
steam generators is particularly important
for residual heat removal. This function is fulfilled by the feedwater system. In addition to
the three feedwater pumps, this system incorporates two start-up and shutdown
pumps which are connected to the emer-
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Emergency feedwater supply
4
3
Turbine
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1
2
2
2
1
2
3
4
5
6
7
8
Feedwater
Steam generator
Containment
Reactor building
Annulus
Emergency feed building
Emergency feedwater pump
Emergency feedwater tank
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7
210 m3
210 m3
210 m3
210 m3
8
8
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gency power supply system and power up
automatically if all the feedwater pumps fail.
The emergency feedwater system is separate
from the water/steam system. Its purpose is to
ensure that the reactor cools down, by feeding
demineralised water into the steam generators
if this can no longer be fed in via the feedwater system or the start-up and shutdown system. The emergency feedwater system is triggered by the reactor protection system if there
is an insufficient water level in the steam generators. Each steam generator is assigned a
pump and an emergency demineralised water
storage tank with a capacity of 210 cubic metres. A further pump and demineralised water
storage tank can be connected up to any of the
three steam generators. All in all, there are
thus 840 cubic metres of demineralised water
available as an emergency supply.
If heat can no longer be removed via the water/steam loops and the emergency feedwater system, the special emergency system will
take over. This could happen as a result of an
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Sa fety precautions
extreme external impact involving the failure
of the switchgear building, the turbine building, the reactor auxiliary building, the water
supply and the third-party supply. The feedwater system, the start-up and shutdown system, the emergency feed system and the special emergency system have a total of eleven
pumps at their disposal for supplying the
steam generators. A single pump is sufficient
to ensure the removal of the residual heat.
The special emergency building is split into
two separate sections. Each of these sections
houses a train of the special emergency system. The building is designed in such a way
that the special emergency system is protected against external impacts, including aircraft crashes, sabotage, fires and earthquakes. Each train of the special emergency
system comprises a feedwater system, a
residual heat removal system, an additional
boron-injection system and well pumps, an
emergency power system, 48V batteries,
rectifiers, a reactor protection system, a demineralised water tank with a capacity of 500
cubic metres, and an emergency diesel generator. From each emergency feed pump,
there is a feed line to a steam generator. To remove the residual decay heat, demineralised
water is fed into at least one steam generator.
The water evaporates, and the steam is released into the atmosphere through the main
steam safety valves. The residual decay heat
can be removed over a period of ten hours,
without need for intervention by the operating
staff. The structural enclosure and physical
separation of the redundant sub-systems provide protection against extensive impacts,
such as fire, flooding or even an aircraft crash.
The electrical cabling and cooling water pipes
are laid in separate locations, for example,
and the instrumentation and control system
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lines are positioned in different sections of the
switchgear building.
In some cases, application of the fail-safe
concept provides additional protection. Wherever possible, the safety systems are designed in such a way that malfunctions or the
loss of the power supply will trigger appropriate safety-related actions. The fail-safe concept has been implemented, inter alia, in the
reactor emergency shutdown system, which
remains effective even if power is lost. The
control rods are attached to the drive mechanism by means of electromagnets. If the
power is cut off, this retaining function is lost
and the force of gravity causes the rods to fall
into the reactor core and shut it down.
As result of the analysis of accidents abroad
(Three Mile Island 2 and Chernobyl), special
emergency measures (safety level 4) were
introduced which ensure that, even for very
rare accident scenarios (simultaneous multiple failure of components and equipment),
the consequences for the neighbourhood of
the power plant will remain limited. To protect
the containment in the highly unlikely event
of a beyond-design-basis incident, a filtered
containment venting system was installed in
1993. Through controlled and filtered venting,
this system prevents containment failure due
to excess pressure. The system is activated by
opening the isolation valves and assures an
effective retainment of aerosols and iodine in
the scrubbing fluid. The separation efficiency
for coarse and fine aerosols is in excess of
99.9 % and for elemental iodine, in excess of
99.5 % percent.
Reactor protection system
The reactor protection system (safety level 3)
monitors the state of the reactor by measur-
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Safety precautions
Containment venting system
Containment isolation valves
Rupture disk
Scrubber unit
Venturi
Scrubbing water
Metal-fibre filter
Throttle
Piping penetrations of
the containment
7
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6
5
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8
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Inside
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1
Nuclear auxiliary building
1
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3
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compare the measured values with the specified limit values and convert them into the
binary signals of «allowed» or «prohibited»
before transmitting them to the logic component. There, they are linked together in
such a way that the necessary commands
are generated for each protective function
from a predefined set of signals.
Vent stack
Safety review
Periodic, comprehensive safety reviews are
carried out in order to assess the safety of
Switzerland’s nuclear power plants. These
periodic reviews permit an overall assessment of the plant’s current safety status. They
take into account all the available results and
the experience acquired from routine inspections, tests, recurring checks, safety
analyses and operational experience.
The safety concept developed for nuclear
technology is based on hypothetical incidents and engineering experience. It is laid
down in laws, decrees, rules, guidelines and
recommendations, including those governing
the design of components and fire protection.
Probabilistic safety and risk analyses (PSAs)
have also been developed for verifying the
design. These have now become established
in practice for assessing nuclear power
plants. A PSA permits reliability assessments
for safety-relevant systems to be conducted
on the basis of the ascertained probabilities
of failure. In addition to this, complex accident sequences involving the failure of safety
sub-systems can be analysed with the aid of
probability considerations. Risk analyses include the assessment of possible damage
outside the plant. A comprehensive PSA was
carried out for the KKG in 1993. The study
M
Annulus
Steel containment
ing characteristic process parameters, including pressure, temperature, neutron flux
and activity. If safety-related threshold values
are exceeded or not attained, the reactor protection system shuts down the reactor before
the design limits are reached. It registers malfunctions and, if necessary, sends signals to
the emergency systems to trigger their active
intervention, such as the closure of the building isolation valves or the start-up of the
emergency cooling systems.
The reactor protection system includes the
full range of devices and installations necessary for triggering protection measures, from
the instrumentation, via the logic component, right through to the control level.
At least two physically different process parameters are taken into account for triggering
protective measures. These are conveyed via
instrumentation lines from the measuring
points to the transmitters, where they are
converted into analogue signals which are
then taken up by the limit-value units. These
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Sa fety precautions
identified, described and quantified accident
sequences and their causes which could lead
to severe reactor core damage. The core damage frequency for the KKG, as established in
the PSA, is in the same range as the frequency for future advanced reactors.
In 1999, work was completed on retrofitting
an independent third cooling line for the
spent fuel storage pool. This additional storage-pool cooling system supplemented the
two existing cooling systems which, as part of
the overall emergency and residual heat removal system, ensure the removal of heat
from the spent fuel assemblies. This project
paid consideration to the results of the PSA
and illustrates the fact that new findings from
safety research are implemented in the plant.
Over the period from 2000 to 2008, the KKG
invested more than CHF 100 million in enhancing plant safety. Amongst the key improvements were the conversion of the pressuriser safety valves and seismic-response
retrofitting measures.
One such measure was the structural reinforcement of the emergency feed building.
The building with the emergency feed system, the chillers and the emergency feedwater tanks is located in the space between the
reactor building in the north, the switchgear
building in the south and the reactor auxiliary
building in the west. The emergency feed
building did not have continuous wall plates
around the containment isolation in these
three main directions to deflect seismic
forces into the foundation. It would not have
been possible to provide proof of seismic resistance, as is required for the periodic safety
review, for such an irregular building structure. The existing building thus had to be reinforced with additional supporting elements.
Safety valves being retrofitted in a reactor
coolant loop.
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Safety precautions
An extensive retrofit of the pressuriser valve
station was implemented in 2005, after three
years’ preparatory work. In addition to two
safety valves, more than 60 fittings, including
all the inlet and outlet lines had to be replaced. The most comprehensive retrofitting
project since the start-up of the reactor focused on areas in the highest safety category.
The benefit achieved in safety engineering
terms was the creation of an additional option for controlled heat removal in the event
of an accident, which is independent of the
existing safety facilities.
The effectiveness of these technical improvements was confirmed in the second
comprehensive periodic safety review completed in 2008. On an international scale,
the KKG has a higher-than-average safety
performance. According to the results of the
updated PSA (status 2008), the risk profile of
the KKG is comparable with that of a new,
third generation nuclear power plant, even
through the technical facilities employed for
the safety precautions differ on points of detail. The KKG operates a comprehensive ageing-surveillance programme, making it possible to identify the need for replacement
investments in good time.
Special emergency system (schematic diagram for a single redundancy)
Special emergency building
Pipe duct
Annulus
Steel containment
7
1 Special emergency feedwater tank
2 Special emergency diesel
and pump
3 Special emergency feedwater pump
4 Control valve
5 Steam generator
6 Well water pump
7 Residual heat exchanger
6
M
M
M
1
4
M
G
~
3
2
5
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29
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Steam and
power conversion system
Low-pressure turbines and generator.
Live steam system
The conventional steam and power conversion system does not differ essentially from
that of a fossil thermal power plant. It essentially comprises the turbine, generator, condenser, condensate and feedwater pumps,
preheaters and the feedwater storage tank,
all of which are located in the turbine building.
The function of the steam and power conversion system is to use the energy released
by the live steam coming from the steam
generators to drive the turbine and the generator coupled to it. After the steam has
passed through the low-pressure turbines, it
is condensed in the condenser.
The condensate is preheated in several
stages and fed back into the steam generators by the feedwater pumps, via the feedwater storage tank. As in all other thermal
power plants, demineralised water is used in
the water/steam circuit and this is prepared
in an on-site facility.
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From the three steam generators in the reactor building, the live steam travels at 280°C
and a pressure of some 62 bar through three
separate pipelines to the live-steam valve station, which houses the spatially separate
safety valves, blow-off valves and isolating
valves. The three live-steam pipes enter the
turbine building via a pipe route, where the total live-steam flow is subdivided into four
lines.
The steam is then conveyed into the doubleflow high-pressure section of the turbine system via four quick-acting stop valves and control valves that are arranged in series. If
necessary, the quick-acting stop valves can interrupt the steam supply to the turbine as a
protective measure.
As the steam leaves the high-pressure turbine, it is still at a pressure of 11 bar with a
moisture content of some 13 % and a temperature of 187°C. To prevent the low-pressure
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Steam and p ower conversion syste m
turbine from being damaged by erosion, the
steam is conveyed via a combined moisture
separator and reheater. This dries the steam
and heats it to around 250°C before the steam
enters the three double-flow, low-pressure
turbines via the inlet nozzles on both sides
and releases its residual useable energy. The
steam is reheated between the high and lowpressure turbines using live steam.
If the turbine system is switched off, livesteam bypass stations divert the steam that
has been produced but not taken up by the
turbines directly to the condensers. The steam
is then eliminated through three quick-acting
electro-hydraulic diverter valves. The livesteam bypass station is designed for a turbine
trip with the reactor power automatically being reduced to 40 %.
If the live-steam bypass station fails, the reactor is scrammed and steam is blown off
through the live-steam safety valves in order
to limit the pressure. A specific and controlled
Reactor coolant system and steam, condensate and feedwater cycle
Live steam
62 bar
6
2
7
2
3
4
3
G
5
~
8
1
2
22
22 bar
3
21
20
19
1
2
3
4
5
6
7
8
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9
10
11
12
13
14
15
16
17
18
19
20
21
22
Reactor
Steam generator
High-pressure turbine
Low-pressure turbine
Water separator
Superheater
Water separator
condensate pump
Condenser
Live steam bypass station
Main condensate pump
Low-pressure condensate cooler
Low-pressure condensate cooler
Low-pressure condensate pump
Feedwater tank
Feedwater pump
High-pressure condensate cooler
High-pressure preheater
Reheater condensate cooler
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10 bar
18
16
4,7 bar
15
1,5 bar
14
0,3 bar
17
9
13
Circulating water system
10
12
11
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0,085 bar
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Stea m and power conversion system
pressure reduction can be initiated via the
blow-off valves in this case.
Turbine-generator unit
The single-shaft turbine-generator unit comprises the high and low-pressure turbine sections as well as the generator, the exciter and
the pilot exciter. The unit is 55 metres long
and rotates at 3000 revolutions per minute.
A box-type condenser is located under each
of the low-pressure turbine casings, which is
rigidly welded to the outer low-pressure
housing. The turbine foundation consists of a
base plate, which is joined to the building
structure by a spring/damper system.
The waste steam from the low-pressure turbine condenses in the adjacent condenser by
releasing the condensation heat into the
main cooling water circuit, which, in turn,
eliminates the heat to the atmosphere via
the cooling tower. The remaining condensate, at a temperature of about 45°C, is
pumped by the main condensate pumps
through three parallel lines of the low-pressure preheater system to the horizontally
arranged cylindrical feedwater tank. Steam
for the low-pressure preheaters is extracted
at the low-pressure turbine.
The double-pole, three-phase synchronous
generator is designed for a nominal power of
1190 megavolt amperes. It consists of the
housing with the bearings, the spring-suspended laminated core with the stator winding, the shaft seal and the current bushing, as
well as the rotor with its brush-free, directcurrent excitation.
In a large-scale generator like this, the stator
winding, together with the circuit ring and
the high-voltage current bushing, are cooled
directly with water, while the rotor winding is
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Live steam pipes running to the turbine building.
cooled with hydrogen. Since the generator is
filled with hydrogen, it is equipped with a
pressure-resistant, gas-tight housing. The hydrogen coolers are positioned vertically on
the turbine side of the generator’s end section. The generator, along with its oil supply,
gas supply, primary water supply and the exciter system are monitored by extensive protection devices which detect inadmissible
operating conditions and leakages, etc.
The power generated at a voltage of 27 kilovolts is fed into the grid via the generator
circuit-breaker, the three block transformers
and the 380 kilovolt switchgear.
Feedwater system
The feedwater tank with a capacity of 500 cubic metres can correct short-term mass-flow
fluctuations in the water/steam circuit. The
feedwater is thermally degassed inside the
feedwater tank; in other words, the non-condensable gases inside the water are expelled.
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Steam and p ower conversion syste m
The feedwater pumps pump the feedwater,
which is at approximately 180°C, out of the
feedwater tank and back into the steam generators via the high-pressure condensate
cooler, high-pressure preheater and preheater drain cooler. Before the preheated
feedwater enters the steam generators, it is at
a temperature of 218°C.
The condensate and feedwater pump systems both consist of three pumps, two of
which are required for full-power operation in
each case. The third one is on standby and
automatically switches on if one of the running pumps fails.
The heating steam for the feedwater tank is
extracted at the outlet from the high-pressure
turbine. The high-pressure preheaters obtain
their heating steam from a tapping point on
the high-pressure turbine.
hour. The pressure is approximately 12 bar,
and the temperature in excess of 200°C. The
quantity of heat transferred is equivalent to
approximately 45 megawatts of thermal
power. The delivery of process heat commenced in December 1979, and, during the
first year, the cardboard factory was already
able to save 11,500 tons of heavy oil in this
way. In 1996, the system was extended by a
small district heating network in the municipalities of Niedergösgen and Schönenwerd.
In 2009, a separate water/steam circuit was
built for Cartaseta Friedrich & Co., a paper factory located in Däniken. This facility is designed for a maximum throughput of 10 tons
of steam per hour, at a pressure of 15 bar.
Process steam extraction
A special evaporator plant at the KKG generates process steam for nearby heat consumers. The customers for process steam extraction include the Aarepapier cardboard
factory in Niedergösgen, which produces corrugated cardboard and cardboard packaging
material.
A heat exchanger in the turbine building removes approximately 1 % of the steam from
the live-steam system in order to heat a water/steam circuit that runs to the cardboard
factory. The steam generated in the heat exchanger flows through a 1.8-kilometre-long
steam line to the cardboard factory, where
the heat is distributed to the various consumers before the condensate is returned, via
feed pumps, to the KKG’s evaporator system.
The steam line to the cardboard factory has
a maximum capacity of 70 tons of steam per
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Process steam for the Cartaseta paper factory.
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33
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Cooling water systems
Lime being precipitated in the settling basin.
through troughs and pipes, and sprayed out
through nozzles. The water then runs as a film
over the plastic elements positioned beneath
the troughs and pipes. The heat exchange
with the air that is rising as a result of the natural-draught chimney effect then takes place
over this large wetted area. This results in
the evaporation of 400 to 700 litres per second, depending on the weather.
The evaporated water is replaced by additional treated water from the headwater
channel of the Gösgen hydropower plant.
The main cooling water system is a dedicated system for removing the heat from the
condensers.
Main cooling water system
The main cooling water system removes condensation heat that has developed in the
turbine condensers and can no longer be
used and releases it to the atmosphere via
the cooling tower circuit. The cooling tower is
150 metres high and has a hyperbolic shell in
reinforced concrete resting on 50 pillars with
their own individual foundations. It is a natural-draught, wet-type cooling tower.
From the cooling tower basin, which is directly beneath the cooling tower, water is
supplied to the two main cooling-water circulation pumps through two separate, parallel, underground intake channels. These
pump the water through the turbine condensers and, from there, back to the cooling
tower.
The water is heated up by 14°C in the condensers and then, at a height of 14 metres inside the cooling tower, it is distributed over
the full cross-section of the cooling tower,
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Auxiliary cooling water systems
During normal operation, about 2.2 cubic
metres of water per second are taken from
the headwater channel. This water is conducted through a culvert under the river Aare
into the auxiliary cooling water pump build-
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Cool ing wa te r syste m s
ing, where it is divided over the nuclear and
the conventional auxiliary cooling water systems.
The auxiliary cooling water system for the
conventional plant area is responsible for
cooling all the cooling units in the turbine and
machine hall, the main cooling water pumps
and two chillers. The replacement water
needed for operating the cooling tower is
also taken from this system and decarbonised. This process softens the water. Calcium hydrogen carbonate is transformed into
insoluble calcium carbonate and precipitated. The calcium carbonate is recycled as a
raw material for the cement industry and as
an agricultural fertiliser.
The remaining water, which has been
warmed up by a maximum of 6.5°C, is fed
back into the headwater channel. This
amounts to around 1.5 cubic metres per second on average. The return pipe passes under the river Aare in a pipe that runs parallel
to the intake pipe.
The nuclear auxiliary cooling water system
ensures the elimination of heat from the
emergency diesel generators, the chillers with
their emergency power supply and the nuclear-component coolant circuit under all
conceivable conditions, with the exception of
an aircraft crash or extreme third-party actions. A draining system to the river Aare ensures the reliable drainage of the auxiliary
cooling water even when the drainage system to the headwater channel is not available.
A second water intake is located at the lower
water channel of the Gösgen hydropower
plant. The mechanically purified water is
transported to the auxiliary cooling water
building through a buried pipeline by two
diesel-driven pumps. This redundant cooling water supply is only required in emergency situations, if the cooling water supply
from the headwater channel fails.
Hydroelectric power station Gösgen
Basic cooling water system
Lower-water channel
Headwater channel
1
8
1 Cooling water inlet
2 Nuclear service cooling water
pumps
3 Conventional service
cooling water pumps
4 Nuclear cooling heat exchangers
5 Emergency diesel coolers
6 Chiller units heat exchangers
(secured supply)
7 Chiller units heat exchangers
(non-secured)
8 Second cooling water intake
9 Overflow and outlet
10 Setting pond for calcium
precipitates
11 Cooling tower
12 Main cooling water pumps
13 Main condensers
14 Conventional plant coolers
(closed circuit)
15 Transformer intercoolers
(closed circuit)
River Aare
9
11
10
12
3
2
4
13
14
5
15
7
6
5
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35
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Station service
power supply
A generator transformer pole being delivered.
The power plant uses about 5 % of the electricity generated to cover its own requirements – mainly for driving the big coolant
pumps. During normal operation, this power
will be taken off between the generator circuit breaker and the block transformer and
conducted to the four separate 10 kV unit
bus bar distributors via two three-winding
transformers. This arrangement means that
the power plant can be supplied with electricity from the grid via the block transformer
even if the generator is at standstill, such as
during maintenance outage. Conversely, in
the event of disruptions to the grid, such as
if the 380 kV high-voltage switchgear has to
be opened during normal operation, the
generator can continue to supply enough
power for all the onsite needs. The KKG then
operates in stand-alone mode and can be
called upon to restabilise the 380 kV power
grid. As an additional reserve, power can
also be supplied from the 220 kV grid, ensuring a full supply to all the distribution
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lines, or at least to individual ones, within a
matter of seconds.
Splitting the bus bar distribution into four
trains corresponds to the redundancies and
the fourfold emergency and residual heat removal systems. Two of these four trains are
sufficient to shut down and cool the reactor
safely even if additional incidents arise.
If the power supply to one train fails, there will
first be an automatic switchover to one of
the reserve supplies. If the power supply is
not restored within a few seconds, the reactor power will be reduced. If a second train
fails simultaneously, an automatic reactor
scram will be initiated. In addition, each of
the four trains is split into normal, emergency
and direct-current networks. The normal networks, with 10-kilovolt and 380-volt distribution, supply big motors, from 500 kilowatt
upwards, as well as the consumers for regular power operation.
The separate emergency networks supply the
key safety-related units, such as the emer-
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Station service p ower supply
gency and residual heat removal systems,
from the 6-kilovolt and 380-volt emergency
distribution systems. A 2940-kilowatt diesel
generator unit is assigned to each train. This
starts up automatically if the main bus voltage in the associated train falls below 80 %
for 2 seconds. Once the diesel start-up time
of a maximum of 15 seconds has elapsed, the
reactor protection system then sequentially
reconnects the key safety-related consumer
groups. Each of the four diesel generator sets
can independently cover the electricity requirements of the associated safety system
line for several hours.
Instrumentation and control systems, such as
the reactor protection system, which need
to operate without interruption during the
diesel start-up phase too, are supplied in duplicate via diode-decoupled, battery-based
48-volt or 220-volt direct-current bus bars.
For particularly important components, which
must be able to implement rapid isolation
and shutdown measures at any time, there
are four non-interrupted, secure 380-volt networks fed by battery-based rotary converters.
For the extremely unlikely case of more than
two safety system lines failing at once, there
are an additional two special emergency systems ready to come into operation. Their
diesel generators start up automatically and
are self-supporting with electricity and
coolant for 10 hours.
Great importance is attached to the electrical
and spatial separation of the four trains. It
must be ensured that no interaction occurs in
the event of electrical malfunctions or a fire.
The cable routing for the different lines is
also kept strictly separate, with the trains insulated against each other from the emergency diesel generator units, via the
switchgear, right through to the electrical
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consumers. The redundancy concept with 6
trains, 4 of which are emergency power networks and 2 emergency standby networks
with a total of 6 diesel generators, has been
consistently implemented in the building
structure, with the switchgear building split
into four parts. This strict spatial separation
is clearly evident when looking at the emergency diesel generator buildings and the special emergency building. These are separate
from the switchgear building and more than
60 metres away from each other. This then
also makes allowance for the consequences
of a hypothetical aircraft crash.
The service power network described above,
which has 35 transformers, supplies the energy for approximately 1400 motors and 950
electric valves.
Electricity feed into the switchyard.
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The shift team in the control room monitoring plant operation.
Plant operation
tration will be highest at this point in time too.
It is then constantly reduced to compensate
for the burn-up of the fuel during the cycle. In
order to achieve reactor criticality, demineralised water is fed into the coolant and the
same amount of borated coolant removed.
This then reduces the boron concentration. To
ensure sufficient shutdown reactivity during
boric acid depletion, the control rods are retracted beforehand.
In steady-state operation, the control rods
are inserted just a short way into the core to
make short-term adjustments to the reactor
power. They are only inserted all the way into
the core in the case of a reactor scram. This
ensures that the maximum possible shutdown reactivity is available, and that the
power distribution within the core is subject
to a minimum of perturbation.
If the removal of reactor heat is impaired due
to the failure of components in the coolant or
water/steam circuits, the reactor power will
be reduced through the insertion of control
The design of the power plant is such that it
can be operated at constant full power or reduced power, as well as in load-following
mode. Operation at constant full power is
preferred for economic and technical reasons; frequent load changes put a strain on
the systems and components, which could
affect the service life of the plant.
In light water reactors, fuel assemblies cannot
be replaced while the reactor is in operation.
This is why the fuel assemblies have a fuel reserve, i.e. surplus reactivity, at the beginning
of each operating cycle. This is reduced in the
course of the operating cycle through fuel
burn-up and the increase in the concentration of fission products. The surplus reactivity is mainly offset by neutron-absorbing boric
acid in the coolant. Since the surplus reactivity is highest at the beginning of an operating cycle, due to the insertion of fresh fuel
assemblies, the requisite boric acid concen-
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38
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rods to such an extent that the balance between heat generation and heat removal is
restored again. The plant can then continue
to operate at reduced power.
signals are captured in separate walk-in
measuring transducer compartments, and
transferred in feedback-free mode, interference-suppressed and protected against external voltages and, in important cases, verified by comparison measurements. Signal
processing is also performed in separate
electronic compartments in the switchgear
and special emergency buildings. The instrumentation and control system, together
with the reactor protection system, are the
principal consumers supplied by the 48V direct-current power distribution. Most of the
measured process parameters are recorded
and displayed in the control room.
Instrumentation and control
The instrumentation and control systems include all the electric and electronic systems
for the monitoring, control and adjustment of
process parameters. This takes in the measurement, data transfer, processing and display of operating parameters such as neutron
flux, pressure, temperature and mass flow.
The KKG uses primarily the Iskamatik B,
Teleperm C/XS and Simatic systems for instrumentation and control. The measured
Information system
The most important sub-systems within the
information systems relating to plant operation are the process data information system,
the training simulator and the security computer system. The process data information
system is an auxiliary tool for plant operation
and monitors the operational state of the
plant. It supplements the conventional plant
instrumentation. The shift personnel and system engineers are supplied with current and
historical information in the form of 7000
alert messages and 1700 process parameters
from the power plant process as a whole. The
shift personnel are trained on a full-scope
simulator, which is a 1:1 copy of the control
room. The simulator training covers normal
plant operation as well as plant incidents. In
addition, a soft-panel simulator is used to familiarise employees with plant sub-systems;
the control consoles of the control room are
displayed on computer screens here.
The security computer system supports the
work of the security guards in terms of access
Water from the reactor coolant loops being analysed.
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ance. Additional tasks include the monitoring
and documentation of modifications to the
plant.
Maintenance means the constant upkeep of
the power plant. Since maintenance and the
quality of the maintenance have a significant
impact on the safety, availability and lifetime
of the plant, the entire power plant with all its
systems, components, devices, equipment
and replacement components is subject to
regular, systematic maintenance. Periodic
tests in the form of inspections and functional checks constitute part of the maintenance schedule and serve the purpose of
certifying the safety of the plant, together
with its systems and components.
With the introduction and further development of suitable diagnostic procedures for
monitoring the status of the plant, preventative maintenance (based on pre-defined test
intervals), is increasingly being replaced by
condition-based maintenance. This requires
detailed knowledge of the components and
their potential weak points and pays particular consideration to design, material, manufacture, assembly, the calculation basis, operational demands, historical test results and
the operating behaviour of the components.
Components that are subject to pressure and
convey radioactivity in the reactor coolant
system are inspected, tested and maintained
over the entire lifetime of the power plant.
Special attention is paid to the reactor pressure vessel, whose welds are inspected from
the inside with ultrasonic testing equipment.
The ultrasonic test also makes it possible to
detect surface flaws and defects inside the
wall. This method is likewise suitable for detecting defects produced during manufacture or generated during operation. These
remote-controlled, periodic tests with ultra-
Process parameters displayed on the simulator.
control, video camera surveillance and alarm
management. Together with biometric systems and non-contact identification systems,
it helps to process and monitor up to 1000
employee entries to the site each day, as well
as more than 20,000 site entries by visitors
each year.
Maintenance and quality
assurance
To ensure that malfunctions of components
are highly unlikely right from the start, all the
structures and plant components that have
an impact on plant safety are inspected on a
regular basis. The authorities and independent experts are also consulted to this end.
These activities form part of the quality assurance, which also includes recurring periodic tests performed during both operation
and maintenance outages. The procurement
and installation of replacement components
is also checked by the KKG’s quality assur-
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sound are performed in different inspection
areas of the reactor pressure vessel and on
the pressure vessel lid.
The fuel assemblies, core support structure,
reactor coolant pumps and steam generators are subjected to special checks. Fuel assemblies can be inspected during outages
and repaired if necessary. The outer surface
of the fuel assemblies can be examined with
underwater cameras. In addition, the socalled sipping test makes it possible to check
the tightness of the fuel assemblies. Underwater cameras are also used to visually inspect the core support structure. Furthermore, areas which are vulnerable to
operation-induced cracking, are examined
by ultrasonic testing. A similar method is ap-
plied to the reactor coolant pumps, which
are also subject to periodic visual inspections. Due to their easier accessibility, however, the defect-free status of most of the areas subject to elevated stressing can be
proven through additional surface-crack inspections.
A remotely-controlled eddy current probe
passes through the steam generator heating tubes, proceeding from the coolant chambers. The probe responds to both material
cracking and differences in wall thickness,
like those brought about by corrosion or mechanical erosion.
Coordinating working plans for the annual revision outage.
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Plant life management
The aim of plant life management is to maintain the safety levels necessary for troublefree operation in accordance with the regulations and to create a sound basis for
planning the service life of the plant. The systematic monitoring of ageing phenomena
and degradation effects covers not only mechanical and electrical systems but also constructional aspects.
Working on the basis of the latest knowledge,
testing methods and operational experience
gained both in Switzerland and elsewhere,
all the ageing mechanisms and effects that
can be captured are investigated and evaluated and subsequently used to identify the
necessary countermeasures. The limited lifetime of the block transformer, for example,
due to the operating loads and the ageing of
the insulation system, prompted the replacement of the three transformer poles and the
spare pole. The 10-kV medium-voltage cables
were replaced on account of the ageing of
the plastic insulation. In addition, after many
years in operation, the control-rod cladding
had been subject to mechanical wear and
was thus replaced. One example of the ageing of mechanical engineering components is
Ultrasonic testing of the pressure vessel.
the wall thinning of heat exchanger tubes in
the low-pressure preheaters, due to droplet
erosion. When this was observed, all three
low-pressure preheaters were replaced by
new ones during the 2008 and 2009 maintenance outages, with the heat-exchanger
tubes, shroud and tube support fittings made
of a new erosion-resistant material. One example of maintenance work on structural
plant components which became necessary
due to ageing was the work carried out on the
reactor dome and the vent stack. In 1997 the
reactor dome was completely cleaned and
underwent preventive sealing in order to protect the structure from environmental impact.
Similar work was carried out on the outside of
the vent stack in 2009.
Replacing a low-pressure pre-heater.
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The familiar effects of ageing include thermomechanical fatigue, vibration damage, radiation embrittlement, and corrosion. The investigation of ageing and making allowance for
it in all its different forms is a prerequisite for
achieving as accurate as possible an estimate
of the residual life of the plant and for investing in measures to extend this service life.
It is basically possible for all the plant components that could potentially curtail the
service life to be repaired or replaced. The
residual technical service life of the plant essentially depends on how the components
and systems are treated. With strict observation of the required safety levels, the residual
service life is determined more by economic
aspects than by purely technical ones. Today, it is assumed that the KKG will be able to
operate a good twenty years longer than the
forty years for which it was originally planned.
Vent stack following refurbishment.
Replacing the high-pressure turbine.
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43
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The KKG and its surrounding area.
Impact of cooling tower
operation
measuring accuracy of such a survey. Between 1980 and 1984, further comprehensive investigations were carried out in order to
monitor the movements of the shadow cast
by the steam plume emerging from the cooling tower. MeteoSwiss evaluated more than
2.5 million photos of the shadow movements
to this end. Between 1976 and 1983, sunshine
recorders were operated at eight locations in
the surrounding area. The measurements revealed that the reduction in sunshine duration
varies as a function of the weather and is essentially confined to a small area to the north
of the cooling tower. Where there is any reduction in the duration of sunshine at all, this
is considerably less than one hour per day.
The steam plume, which consists of pure water vapour is generally less than 200 m high
in summer, but can rise to more than 800 m,
depending on the humidity. In the vicinity of
the KKG, there are no areas subject to the
shadow for a length of time which would lead
to compensation entitlements. The cooling
The cooling tower, with a height of 150 m, is
a conspicuous feature of the landscape between Olten and Aarau and can be seen from
a long way off. Before granting the construction licence, the Swiss Federal Office of Meteorology and Climatology (MeteoSwiss) investigated the potential impact of cooling
tower operation on the environment. The
comprehensive investigations were completed in 1984 and revealed no appreciable
adverse effects on the environment.
The investigations provided no evidence of
significant variations in precipitation in the
area around the cooling tower, and there was
no proof of any increase in the formation of
fog or black ice. A change of less than 0.2°C in
the average annual temperature above
ground level was determined, together with
an increase of at most 3 % in the annual humidity; these slight fluctuations are within the
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tower has also the desirable side-effect of
partly washing out the air.
gases are detected in discharge measurements, this will automatically trigger the isolation of potential release paths.
The liquid radioactive substances released
are mainly tritium, which originates essentially from the boron burn-up. The liquid discharges also contain traces of activation
products, especially cobalt-60. Occasionally, antimony-124 and iodine-131 are also
found. Effluent water is only discharged if
the authorised limits are met. The authorised limits for the release of radioactive
substances and the programmes for monitoring these emissions are specified in the
operating licence and in the discharge regulations issued by the licensing authorities.
The KKG measures the discharges into the
environment and reports these to the authorities on a monthly basis. The data is verified by independent control measurements
conducted by the authorities themselves.
The release of radioactive substances to the
environment is documented so that evidence of the type and quantity of discharges
can be provided at any time.
As part of the immission monitoring, samples of water are taken from the river Aare
each week. Sediments from the Aare are
similarly examined. The stationary airborne
immission monitoring system involves the
measurement of the local dose at 24 locations within a radius of 5 to 7 kilometres of
the plant. The dosimeters are read and evaluated four times a year. Further dosimeter
readings are taken at a total of 32 points on
the plant site, at the cooling tower and
around the perimeter fence, which are similarly evaluated on a quarterly basis. To
record the environmental radioactivity, air
filters are evaluated once a week and the
rainfall examined.
Release of radioactive
substances
During normal operation, the plant releases
slight quantities of radioactive substances
into the environment with its waste water
and exhaust air. The airborne releases include radioactive noble gases and radioactive iodine, which result from nuclear fission, radiocarbon (carbon-14) which comes
from the activation of oxygen, and also radioactive aerosols, which originate primarily
from the activation of construction materials. If elevated concentrations of noble
Aerosol collector.
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Once a month, dust particles collected on
four Vaseline-coated panels in the vicinity of
the plant are examined. In addition to this,
gamma-spectroscopic measurements are
conducted once a year to determine the ac-
tivity concentration of selected radionuclides. The carbon-14 content of tree leaves
is measured. Samples of the soil, grass, milk
and crops and also fish from the river Aare
are taken once a year and examined in order
to check for possible deposits in the ground,
in foods and in animal feedstuffs.
Since 1993, the Swiss Federal Nuclear Safety
Inspectorate (ENSI) has been operating an
automatic dose-rate monitoring network
(Maduk) in the vicinity of the nuclear power
plants. At 16 locations close to the KKG there
are sensors (Geiger-Müller counters), which
transmit the readings to the central ENSI
computer every ten minutes, where they are
automatically checked and compared with
the natural background radiation. The current measurements are posted on
www.ensi.ch. The Maduk network supplements the National Emergency Operations
Centre (NEOC) network for the automatic
monitoring of radioactivity. The NEOC network consists of 60 stations distributed over
the entire country, which similarly measure
the local dose rate. These measurements
are publicly available at www.naz.ch. To
record the natural and man-made sources of
radiation over a wider area, the NEOC carries
out aerial radiometry measurements from a
helicopter every one to two years, covering
an area of 70 square kilometres around the
KKG. Alongside the ENSI, the Department
of Environmental Radioactivity of the Federal Office of Public Health is responsible for
monitoring immissions from the nuclear
power plants. The results of the emission
and immission monitoring are published
annually in the Federal Office of Public
Health FOPH report on «Environmental radioactivity and radiation dosages in Switzerland».
Sensor for automatic dose-rate measurement.
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The radiation dose received by the general
public from the immissions is calculated on
the basis of the emission values. ENSI specifies the maximum values for radioactivity
emissions to the environment in such a way
that no one in the local population is exposed to more than the source-linked reference dose level of 0.3 millisievert per year.
The radiation doses that result for the local
population from the radioactivity emissions
are several orders of magnitude lower than
those from natural radiation sources. By way
of comparison: in Switzerland the average
dose from natural radiation sources is 3 millisievert per year, with extreme values of 1 to
25 millisievert per year.
For the point in the vicinity of the KKG that
is subject to the hypothetical maximum impact, a maximum annual whole-body exposure of less than 0.01 millisievert has been
calculated in the period since the plant was
brought into operation, taking into account
all possible exposure pathways. At no point
in the vicinity of the KKG have harmful effects due to radioactivity from the KKG been
observed since the power station first came
on stream.
Sampling water from the river Aare.
for the acquisition of meteorological data for
emergency response planning to MeteoSwiss. The technically optimised MeteoSwiss stations feed site-specific data into
the dense MeteoSwiss measuring network.
Meteorological data capture
Since 2007, the acquisition of meteorological
data that is required at all nuclear plant sites
for purposes of incident analysis has been
performed with new, standardised MeteoSwiss meteorological stations. The former
meteorological station on the site of Aarepapier AG, which was brought into operation by
the KKG in 1982, was then no longer required
and was dismantled in 2009. The Swiss Federal Law of 4 October 2002 on civil protection
and support services assigned responsibility
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Nuclear fuel cycle
Spent fuel assemblies and radioactive waste from reprocessing are stored at the ZZL facility.
Uranium mining
The nuclear fuel cycle is a term used to describe the full range of activities and services
related to the fabrication, use and final disposal of nuclear fuel. This includes uranium
mining, conversion and enrichment, the fabrication of fuel assemblies, and interim storage, as well as the reprocessing of spent fuel
and the final disposal of waste from reprocessing operations and spent fuel assemblies. The nuclear fuel cycle also takes in the
recycling of uranium and plutonium obtained
during the reprocessing of spent fuel assemblies. Since the corresponding services are
provided at different locations, suitable transport casks have to be available.
The primary energy source in today’s nuclear
power plants is uranium. Uranium is used in
the fuel assemblies inside the reactors of nuclear power stations. The term «fuel element
supply» denotes the chain of services from
uranium mining through to the final loading
of the fabricated fuel assemblies into the reactor.
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Uranium is a heavy metal, with low radioactivity, which is found in a large number of
minerals and is some 500 times more widespread than gold. Uranium ore, which is
used as the raw material for nuclear fuel, is
obtained, inter alia, by mining. The most
productive uranium mines are in Canada,
Australia, Kazakhstan, Niger, Namibia and
Uzbekistan. The largest deposits of uranium
have been found in Australia, Kazakhstan,
Canada, Russia and South Africa.
The uranium ore is crushed and ground in an
ore-processing plant. A uranium concentrate
(U3O8 – commonly known as «yellow cake»)
is obtained from the host mineral in a multistage chemical leaching and extraction
process. This concentrate is turned into uranium hexafluoride (UF6) in a further conversion process, which has the characteristic
properties required for the subsequent enrichment process.
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Nucl ea r fuel c ycl e
From the 1970s up to the 1990s, the KKG
bought natural uranium on the international
market and also obtained it through a partnership with a US mining company. Instead of
natural uranium, it is also possible to use the
uranium and plutonium recovered during
spent fuel reprocessing as an energy source
for fuel assemblies.
contributes to reducing stocks of military uranium. By using reprocessed uranium, the KKG
is able to make savings of some 180 tons of
natural uranium each year.
Fuel assembly fabrication
Following enrichment, the uranium hexaflouride (UF6) is converted into uranium dioxide (UO2), which is the starting material for
fuel pellets. These ceramic pellets are inserted into Zircaloy cladding tubes, which
are welded so that they are gas-tight. Two
hundred and five such fuel rods are made up
into a fuel assembly. The enrichment level of
the KKG fuel assemblies is between 4.5 and
around 5 % uranium-235. Fuel assemblies of
this type can achieve average burn-ups of 55
to 65 megawatt days per kilogram.
Uranium can be replaced by plutonium as a
primary energy source. Mixed oxide (MOX)
fuel assemblies contain a mixture of uranium
dioxide (UO2) and plutonium dioxide (PuO2).
The uranium carrier material is depleted, i.e.
it contains virtually no uranium-235. The
added plutonium is obtained during the reprocessing of spent fuel assemblies and is itself a mixture of several plutonium isotopes.
The external appearance of a MOX fuel assembly does not differ from that of a uranium fuel assembly.
Plutonium is bred in a light water reactor
through the conversion of uranium-238. In a
conventional uranium fuel assembly, plutonium thus makes a contribution of some
40 % to the power generated. In a reactor
core with one third MOX fuel assemblies, the
contribution of the plutonium to the reactor
power can even be as high as some 60 %.
The reprocessing of around 400 tons of KKG
fuel assemblies gave rise to some four tons
Uranium enrichment
Natural uranium is a mixture of uranium-238
(99.28 %), fissionable uranium-235 (0.71 %)
and a very small amount of uranium-234. Today, light water reactors use uranium fuel
containing about 4 to 5 % uranium-235. The
process involved in increasing the uranium235 concentration of natural uranium to the
concentration required in reactor operation is
called enrichment. Various isotope separation techniques have been developed for the
enrichment of natural uranium. Only the gas
diffusion technique and gas centrifuge technology are used on a commercial scale, both
of which require uranium in a gaseous form
(UF6).
The enrichment of uranium can also be
achieved by mixing it with other higher enriched uranium. This blending process, which
gives the typical enrichment levels required
for light water reactors, is employed in Elektrostal’s fuel fabrication plants in Russia. To
manufacture fuel pellets, uranium from spentfuel reprocessing with a residual enrichment
of less than 1 % uranium-235 is blended with
uranium from Russian stocks which has an enrichment of 20 to 30 %. Since 2000, the KKG
has been using fuel assemblies made from reprocessed uranium which are fabricated in
Russia under licence from the fuel supplier
Areva NP. This is sparing on resources and
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Nucl ear fuel c ycle
of plutonium which was used in the KKG reactor in the form of MOX fuel assemblies over
the period 1997 to 2007.
recommendations issued by the International
Atomic Energy Agency (IAEA). The regulations
are designed to protect people and the environment from harmful radiation and to protect the material being transported from external impacts. In the case of irradiated fuel
assemblies that are being transported to a reprocessing site or to an interim storage facility, the requisite protection is provided by
transport casks which constitute a radiation
shield. Prior to certification, proof must be
submitted of the fact that the casks can withstand the most severe of accident scenarios
and remain completely tight.
Reprocessing
Irradiated fuel assemblies contain approximately 95 % uranium, 1 % plutonium and
4 % fission products. The precise composition depends on the discharge burn-up of
the fuel assemblies. During the reprocessing
operation, the structural materials are separated from the fuel. The fuel is split into uranium, plutonium and fission products by
chemical means. The extracted energy carriers of uranium and plutonium are reused in
fuel fabrication and recycled in the reactor.
The fission products are embedded in a glass
matrix, which is then welded into a steel container. These fission products form the highlevel waste. The structural materials from the
irradiated fuel assemblies are processed into
intermediate-level waste. Each year, 3.7 cubic
metres of high-level waste and 3 cubic metres
of intermediate-level waste are produced
through the operation of the KKG.
Irradiated fuel assemblies can be disposed of
either with or without reprocessing. Whether
the uranium and plutonium are recycled or
not is subject to political influences. In
Switzerland, for example, a ten-year moratorium came into effect in 2006 prohibiting the
transport of irradiated fuel assemblies to reprocessing sites.
Interim storage
The Central Interim Storage Facility in Würenlingen (ZZL) can take high-level, intermediate-level and low-level radioactive waste. This
also includes the vitrified high-level and intermediate-level waste from reprocessing
and from irradiated fuel assemblies from the
nuclear power plants. Prior to transfer into a
final repository, all high-level waste needs to
be placed in interim storage (i.e. cooled) for
30 to 40 years on purely physical grounds in
order to remove decay heat. The ZZL has sufficient capacity for an even longer period of
storage.
Geological repository
After more than 30 years of investigations
and research, comprehensive knowledge and
a basis for decision-making have now been
acquired for the establishment of repositories
in deep geological formations that are required for radioactive waste. At the end of
June 2006, the Federal Council approved the
«Demonstration of Disposal Feasibility» for
Transport of irradiated fuel
assemblies
The transport of irradiated fuel assemblies
and other radioactive materials is subject to
statutory regulations, which are based on
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Nucl ea r fuel c ycl e
high-level radioactive waste issued by Nagra (National Co-operative for the Disposal of
Radioactive Waste). This then provided conclusive evidence of the basic feasibility of
permanent, safe disposal of all the nuclear
waste in Switzerland. In April 2008, the Federal Council approved the «Deep Geological
Repository Plan», a land-use-planning instrument which specifies the site selection
procedure for deep geological repositories.
Different potential sites were identified, marking the first step towards site selection. Site
selection will be conducted in a transparent
and democratically supported process.
Nuclear fuel cycle
Fuel assemblies
Fuel assemblies
Fabrication of mixed
oxide (MOX) fuel
assemblies
Interim storage facility
for fuel assemblies
and radioactive waste
Reprocessing
of fuel assemblies
Enrichment
Interim storage
facility (ZZL)
Würenlingen
Fuel assemblies
La Hague
Radioactive waste
Depleted uranium
Fabrication
of uranium
fuel assemblies
Reprocessing plant
Fuel assemblies
Gösgen NPP
Radioactive waste
Plutonium
Uranium
Conversion
Conditioning
Extraction and
Purification
Fuel element
Rock laboratory
assembly, Lingen
Mont Terri
(Photo: Areva)
(Photo: BGR)
Uranium ore
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Repository
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Upgrading, retrofitting,
modernisation
1981
In order to enhance plant safety and operational reliability, some CHF 700 million has
been invested in major projects since the
plant was commissioned, in addition to the
regular maintenance work. A selection of key
milestones in the operation of the plant and
in terms of the technical improvements made
is included in what follows:
쐍 15 May: official opening of the nuclear
power plant. End of the two year warranty
period. Plant taken over from the general
contractor, Kraftwerk Union AG.
쐍 Further improvements: replacement of the
feedwater tank, modifications to the steam
generators, overhauls of two low-pressure
turbines
1979
1982
Process heat pipeline to the cardboard factory.
Spent fuel transport cask.
쐍 19 January: initiation of the first self-sustained chain reaction
쐍 6 February: the first electricity is fed into the
Swiss national grid
쐍 30 October: full-power operation commences
쐍 20 December: start of the process steam
supply to the cardboard factory in Niedergösgen
쐍 Modernisation of the turbine system to improve efficiency
쐍 A new wing added to the administration
building
쐍 First shipment of spent fuel assemblies to
the reprocessing plant in La Hague, France
1983
쐍 Complete renewal of the insulation for the
three steam generators
1980
쐍 Comprehensive improvements, especially
to the conventional part of the plant
쐍 Capacity increase for the spent fuel storage
pool
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1984
쐍 Completion of the tube replacement in the
three condensers
쐍 Chemistry in the water/steam circuit
switched to pure hydrazine conditioning
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Upgrading , retrofitting , m odernisation
1985
쐍 Reworking of the seating surfaces for the
core structures in the reactor pressure vessel
쐍 The Federal Council approves the increase
in nominal thermal power
1990
쐍 Extension of the active fuel length inside
the fuel assemblies
1986
쐍 Conversion work on the overpressure-protection devices for the reactor coolant system
쐍 Completion of the improvement work to
the low-pressure turbines, which had commenced in 1981
1991
쐍 Completion of a programme spanning several years involving improvements to reactor components
쐍 Fuel assemblies with exceptionally corrosion-resistant Duplex cladding used for the
first time
쐍 Renovation of the power distribution for
peripheral facilities
1987
쐍 Extension and alterations to the switchgear
building
쐍 Multi-storey extension to the storage and
workshop building and reconstruction of
the big-component store
1992
쐍 From July onwards: plant operated with the
licensed maximum thermal power of 3002
MW
1988
1993
Exchanging the bolts on the core shroud.
쐍 Replacement of the bolts on the core
shroud of the reactor pressure vessel completed
쐍 Spare generator rotor procured
Installing a gas scrubber for the filtered
venting system.
쐍 Filtered venting system for the containment retrofitted
쐍 Introduction of an electronical information and documentation system
1989
쐍 Alterations to the pilot valves for the live
steam isolation valve system
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!( ,&0 * !11 * .** ,&
' * * .'' * &
"* * *& &* & *
"," "&'
*& & & * & ' .* &' , ''' & * &'* *
# ."**& $
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Upgrading , retrofitting , m odernisation
terim Storage Facility for Radioactive Waste
(ZZL) in Würenlingen
tor coolant system, expanding the pool
storage capacity for spent fuel assemblies
in a new storage building, extending the reactor auxiliary building with an annex,
procuring a spare excitation set, replacing
internal fittings for the cooling tower, efficiency-improving work on turbines and water separator/reheaters and procurement
of a new generator circuit-breaker.
2000
2003
Transporting the generator stator to the KKG.
쐍 Generator stator replaced
쐍 Fuel assemblies made from reprocessed
uranium used for the first time
쐍 Delivery taken of the training simulator
from STN Atlas, Bremen, now Rheinmetall
Defense Electronics
Fitting the new generator circuit breaker.
쐍 Start of renovation work on the internal fittings of the cooling tower
쐍 Replacement of the hydraulic-mechanical
speed monitoring device in the turbine system
쐍 Replacement of the generator circuitbreaker
2001
쐍 Process computer replaced by a process
data information system
쐍 ZZL in Würenlingen brought into commercial operation
쐍 First return shipment of vitrified high-level
waste to the ZZL from reprocessing in La
Hague
쐍 Start of several years’ upgrading work on a
number of buildings to improve seismic
resistance and intrusion protection
2004
쐍 Certification of the process-oriented KKG
management system, which was introduced in 2003, by the Swiss Association for
Quality and Management Systems (ISO
9001:2000 for quality management, ISO
14001:1996 for environmental management and OHSAS 18001:1999 for occupational health and safety management)
2002
쐍 Modernisation projects involving the investment of more than CHF 200 million at
the planning stage. This includes retrofitting a pressure relief system to the reac-
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Upgrading , retrofitting , modernisation
쐍 Integrated emergency management system implemented: revised operation and
emergency handbook introduced
2007
Renovating the internal fittings of the cooling
tower.
쐍 Licence granted for the construction and
operation of the new spent fuel storage
building in accordance with the nuclear
energy legislation
쐍 Modernisation of the security computer
system
Building work on the spent fuel pool storage
building.
쐍 Controlled pressure relief system retrofitted
to the reactor coolant system
쐍 Structural modification of the turbine area
to improve plant efficiency
쐍 Replacement of the reheater bundle
쐍 Zinc added to the coolant for the first time
쐍 Analogue turbine control system partially
replaced by a digital one
쐍 Replacement of a 220 kV external grid
transformer
쐍 Extension to the reactor auxiliary building
and the new wing of the administration
building brought into service
쐍 Nuclear waste accumulated over 28 years
of reactor operation is conditioned
2006
2008
2005
쐍 Replacement of the generator excitation
equipment
쐍 Review of the probabilistic safety analysis
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쐍 New spent fuel pool storage building
brought into service
쐍 Seal systems on all reactor coolant pumps
replaced
쐍 The three 380 kV generator transformer
poles and the spare pole replaced
쐍 Two low-pressure preheaters replaced
쐍 Regular ten-year safety assessment completed
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Characteristics
Characteristics
1 (3-times)
2 (2-times)
8 (3-times)
3 (4-times)
4 (2-times)
5 (3-times)
7 (4-times) 6 (6-times)
1
2
3
4
5
6
7
8
Feedwater pumps
Auxiliary/start-up feedwater pumps
Emergency feedwater pumps
Special emergency feedwater pumps
Reactor coolant pumps
Low-pressure injection pumps
High-pressure injection pumps
Accumulators
380 kV
220 kV
G
~
G
~
HP
6 kV
LP
DG
DG
DG
DG
DG
DG
380 V
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HP
LP
G
DG
= High-pressure turbine
= Low-pressure turbine
= Generator
= Diesel generator
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380-kV switch yard
220-kV switch yard
Electrical systems
Switch gear
Main transformer
380/27 kV
27 kV AC powerline
Startup transformer
60/32/32 MVA
Startup transformer
60/32/32 MVA
Auxiliary transformer
60/32/32 MVA
Auxiliary transformer
60/32/32 MVA
Main generator
1190 MVA
10-kV AC non-essential bus train 4
Various motor supplies
380-V AC non-essential
bus train 4
bus train 3
380-V AC non-essential bus train
bus train 2
bus train 1
380-V AC bus for
pressurizer heaters
380-V AC bus for
pressurizer heaters
220-V DC bus for
control rods
Emergency diesel 3550 kVA
220-V DC bus for
control rods
6-kV AC essential
bus train 4
6-kV AC essential
bus train 3
6-kV AC essential
bus train 2
380-V AC essential bus train 4
6-kV AC essential
bus train 1
24/48-V DC bus train 4
220-V DC bus
train 4
220-V DC bus
train 3
220-V DC bus
train 2
220-V DC bus
train 1
Motor
generator set
175 kVA
Motor
generator set
175 kVA
Motor
generator set
175 kVA
380-V AC regulated
bus train 3
380-V AC regulated bus train 4
Motor
generator set
175 kVA
Special emergency diesel 750 kVA
Special emergency diesel 750 kVA
380-V AD special
emergency bus train 6
220-V DC bus
380-V AD special
emergency bus train 5
Motor generator
standby set
175 kVA
380-V AD special emergency bus train 7
380-V AC
regulatet bus
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380-V AC uninterruptable distribution for process computer
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24/48-V DC bus
train 6
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Inte rnet addresses
Internet addresses
쐍 Swiss Federal Office of Energy (SFOE)
www.bfe.admin.ch
쐍 National Cooperative for the Disposal of
Radioactive Waste (Nagra), a technical
and scientific organisation set up by
those responsible for the disposal of nuclear waste (the Swiss Confederation and
the Swiss nuclear power plant operators)
www.nagra.ch
쐍 Federal Office of Public Health (FOPH)
www.bag.admin.ch
쐍 Swiss Federal Nuclear Safety Inspectorate
ENSI, the federal licensing authority
responsible for the nuclear safety and
security of Swiss nuclear facilities
www.ensi.ch
쐍 The Nuclear Forum, a scientific-technical
organisation
www.nuklearforum.ch
쐍 Decommissioning and waste disposal
funds
www.entsorgungsfonds.ch
쐍 Paul Scherrer Institute (PSI), a multidisciplinary research institute for the natural
and engineering sciences
www.psi.ch
쐍 Demonstration of the feasibility of radioactive waste disposal
www.entsorgungsnachweis.ch
쐍 Radioactive waste
www.radioaktiveabfaelle.ch
쐍 Grimsel test site (GTS), an underground
scientific laboratory in a crystalline rock
formation run by Nagra, located at the
Grimsel Pass, Haslital, Canton Bern
www.grimsel.com
쐍 Decommissioning fund
www.stilllegungsfonds.ch
쐍 Swissnuclear, the nuclear power subsection of Swisselectric (the organisation of
Swiss electricity grid companies)
www.swissnuclear.ch
쐍 Mont Terri rock laboratory (FMT) in an
opalinus clay formation located near
St. Ursanne, Canton Jura
www.mont-terri.ch
쐍 Association of Swiss Electricity
Companies
www.strom.ch
쐍 Nuclear-power internet portal
www.kernenergie.ch
쐍 ZWILAG Zwischenlager Würenlingen AG,
the central interim storage facility for all
types of waste for the operators of
Switzerland’s nuclear power plants
www.zwilag.ch
쐍 National Emergency Operations Centre
(NEOC), the federal centre of expertise for
exceptional events
www.naz.ch
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Periodicals
Regular publications
쐍 Annual Report of Kernkraftwerk Gösgen-Däniken AG (www.kkg.ch)
쐍 Bulletin of the Swiss Nuclear Forum, Bern (www.nuklearforum.ch, covers general nuclear power
related topics in short summaries and operating figures for Switzerland’s nuclear power plants,
published monthly)
쐍 Nagra Annual Report, National Cooperative for the Disposal of Radioactive Waste (Nagra), Wettingen (www.nagra.ch)
쐍 Supervision Report, Radiation Protection Report, Swiss Federal Nuclear Safety Inspectorate
(ENSI), Brugg (www.ensi.ch, reports compiled by the public authorities on the operation of
Swiss nuclear power plants, the confederation’s supervisory activities and radiation protection)
쐍 Environmental radioactivity and radiation doses in Switzerland, Federal Office of Public Health
(FOPH), Department of Radiation Protection, Bern (www.bag.admin.ch, compilation of the results of radioactivity monitoring, published annually)
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Key technical data
Power
Gross electrical output
Net electrical output
Thermal reactor output
1035 MW
985 MW
3002 MW
Reactor building
Outside diameter
Height above base plate
Wall thickness in the cylindrical part
Wall thickness in the dome
Thickness of base plate
63.6 m
56.8 m
1.6 m
1.2 m
2.8 m
Steel containment structure
Inside diameter
Wall thickness
Design overpressure/temperature
52 m
32 mm
4.89 bar/135°C
Inside diameter
Wall thickness of cylindrical shell (without cladding)
Material
Cladding thickness
Total height including closure head
Design pressure/temperature
Weight without internal structures
Weight of core internal structures
4360 mm
221 mm
22NiMoCr3-7
6 mm
10,827 mm
175 bar/350°C
360 t
135 t
Reactor
Coolant and moderator
Fuel
Number of fuel assemblies
Overall weight per assembly
Fuel rods per fuel assembly
Arrangement
Overall length of fuel rods
Active fuel length of a rod
Outer diameter of fuel rods
Cladding tube material
Cladding tube wall thickness
Total uranium weight in core
Enrichment reload fuel assemblies
Average burnup at discharge
Average heat flux density
Average linear power
Number of control assemblies
Absorber rods per control assembly
Absorber material
H2O
uranium (UO2) and Mox (UO2 and PuO2)
177
666 kg
205 (Mox: 204)
square lattice configuration
3860 mm
3550 mm
10.75 mm
Zry-4/DX ELS 0.8
0.725 mm
76 t
4.6–4.95 % U-235 equivalent
55–65 MWd/kg HM
67.5 W/cm2
228 W/cm
48
20
AgInCd
Plant overview
47
26
30
47
25
24
27
3
containment
sump
1
3
28
3
69
14
68
70
6
14
66
12
35
76
13
16
19
17
78
64
15
8
9
18
57
37
21
58
38
11
43
44
Residual heat removal system
28 Residual heat removal pump
29 Residual heat exchanger
30 Accumulator
31 Borated-water storage tank
32 Safety injection pump
Off-gas system
33 Recombiner
34 Waste gas compressor
35 Delay line
36 Vent stack
Nuclear component cooling system
23 Component cooling pump
24 Component cooling heat exchanger
Coolant purification system
12 Mixed-bed filter
13 Coolant degassing system
Spent fuel pool cooling and
purification system
25 Spent fuel pool
26 Spent fuel pool purification pump
27 Mixed-bed filter
Nuclear service water system
22 Nuclear auxiliary service water
pump
Chemical control system
10 Boric acid tank
11 Boric acid pump
45
41
42
Coolant storage and
treatment system
14 Coolant storage tank
15 Demineralised-water recirculation
pump
17 Preheater
18 Evaporator
19 Condensate pump
20 Degasifier
21 Degasifier extraction pump
Volume control system
5 Recuperative heat exchanger
6 High-pressure cooler
7 High-pressure reducing station
8 Volume control surge tank
9 High-pressure charging pump
40
39
46
1 Reactor
2 Steam generator
3 Reactor coolant pump
4 Pressuriser
64
65
20
10
67
77
7
34
49
5
33
48
31
36
49
2
2
32
48
47
4
22
29
49
2
23
headwater
channel
48
60
61
59
50
51
53
54
54
G
~
54
Generator
Demineralisation
system
73
52
52
52
62
62
62
74
55
75
71
56
72
63
Component drain system
37 Drain tank and drain pump
53 High-pressure turbine
54 Low-pressure turbine
Liquid waste processing system
38 Liquid waste tank
40 Evaporator
41 Monitoring tank
42 Discharge pump
43 Concentrate tank
44 Concentrate pump
45 Condenser
46 Waste solidification facility
Moisture separator drains system
55 Moisture separator drain tank
56 Moisture separator drain pump
Main feedwater system
66 Feedwater tank
67 Main feedwater pump
68 High-pressure preheater
69 Reheater drain cooler
70 Start-up/shutdown pumps
Auxiliary steam system
57 Auxiliary steam manifold
58 Auxiliary boiler
Main cooling water system
71 Cooling tower
72 Circulating water pump
Process steam system
59 Process steam generator
60 Process steam superheater
61 Process steam to cardboard factory
Emergency feedwater systems
73 Emergency feedwater tank
74 Demineralised-water refilling pump
75 Emergency feedwater pump
76 Special emergency feedwater tank
77 Special emergency feedwater pump
78 Well water pump
Live steam system
47 Live steam safety valve
48 Live steam relief station
49 Live steam isolation valve
50 Moisture separator
51 Superheater
52 Live steam bypass station
Main condensate system
62 Condenser
63 Main condensate pump
64 Low-pressure preheater
65 Auxiliary drain pump
Key te chnica l da ta
Drive system
Number of coolant loops
Operating gauge pressure
Coolant inlet temperature
Coolant outlet temperature
Coolant flow rate
magnetic jack
3
154 bar
292°C
325°C
15,984 kg/s
Steam generators
Number
Height
Diameter
Shell material
Tube sheet material
Tube material
Tube dimensions
Design pressure/temperature
Total weight
3
21,200 mm
3570/4860 mm
fine-grained steel
fine-grained steel
Incoloy 800
Ø 22 x 1.2 mm
175/87.3 bar/350°C
380 t
Reactor coolant pumps
Number/type
Discharge head
Design flow rate per pump
Speed
Motor power (design)
3 single-stage mixed-flow
centrifugal pumps
84.4 m
5328 kg/s
1490 rev/min
9200 kW
Pressuriser
Height
Diameter
Volume
Operating pressure/temperature
Heating power of the heating rods
13,400 mm
2400 mm
42 m3
154 bar /344°C
1400 kW
Steam and power conversion system
Live-steam flow rate
Live-steam conditions at steam generator outlet
Steam moisture at steam generator outlet
Exhaust wetness
Condenser pressure
Cooling-water temperature
Condenser circulating water flow rate
Feedwater heating temperature
Number of feed-heating stages
5890 t/h
64.5 bar/280.3°C
max. 0.25 %
10 %
80 mbar
22°C
120,500 m3/h
218°C
5
Turbine
Fourfold-casing single-shaft condensing turbine with a double-flow high-pressure turbine and
three double-flow low-pressure turbines. Steam drying and reheating between the high-pressure turbine and low-pressure turbines.
Key te chnica l da ta
Speed
Turbine gross effective power
Length of turbine-generator system
3000 rev/min
1035 MW
55 m
Generator
Apparent power
Power factor (cos )
Terminal voltage
Frequency
Cooling rotor winding
Cooling stator winding
1190 MVA
0.9
27 kV
50 Hz
hydrogen (6 bar), 7 bar abs.
water (27 kg/s)
Generator transformer
Number/type
High voltage side
Low voltage side
Power capacity
3 single-phase units and 1 stand-by unit
409 kV
27 kV
1200 MVA
Main feedwater pumps
Number/type
3 double-flow double-stage
radial centrifugal pumps
812 m
844 kg/s
8600 kW
Discharge head (backing and main pump)
Design flow rate per pump
Motor power
Cooling tower
Number/type
Height
Diameter at base
Diameter at top opening
Throat diameter
Bottom shell thickness
Minimum shell thickness
Water flow rate
Warm water temperature
Cold water temperature
Dry bulb temperature
Wet bulb temperature
Air-flow rate
Water evaporation rate
natural circulation wet-type
150 m
117 m
74 m
70 m
750 mm
160 mm
33.8 m3/s
36°C
22°C
7.8°C
6.2°C
25,400 m3/s
0.4–0.7 m3/s
Circulating water pumps
Number/type
2 single-stage, mixed flow
centrifugal pumps
20.5 m
16.9 m3/s
248 rev/min
4100 kW
Discharge head
Nominal flow rate per pump
Speed
Motor power

Technology and Operation

Transcription

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