1. Reactor Coolant Pressure Boundary PWR
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4
Number of Loops 2 or 4 3 or 4 4 4
Suitability of the components for non-destructive testing
Yes, with minor restrictions Yes
Components - Seamless forged rings for
vessels
Reactor pressure vessel, steam generators (primary side only)
Reactor pressure vessel, steam generators,
pressuriser - Seamless pipes Main coolant line
with minor restrictions
Main coolant line
Materials - Ageing-resistant ferritic
fine-grained structural steels with stabilised austenitic cladding
All components and pipes with nominal diameter > 400 mm
Like construction lines 1-3, but with optimised
qualities - Ageing-resistant stabilised
austenitic steels
All pipes with nominal diameter < 400 mm and component internals
- Corrosion-resistant steam generator tube material (Incoloy 800)
Yes (after exchange
of steam generators in
one plant)
Yes
Application of the rupture preclusion concept
Post-commissioning qualification Prior to commissioning
From the start of planning Reduction of embrittlement from
neutron radiation exposure
Use of dummy fuel elements
and special fuel element management
Optimised welding material and enlargement of water gap in the reactor pressure vessel to
reduce neutron fluence
1. Reactor Coolant Pressure Boundary BWR
Design Characteristics Construction Line 69 Construction Line 72 Re-circulation pumps integrated
in the reactor pressure vessel
8 to 10 8
Suitability of the components for non-destructive testing
Yes,
with minor restrictions
Yes
Components - Seamless forged rings for
reactor pressure vessels
No Yes, except: closure head, bottom
head
- Seamless pipes Yes,
after replacement of pipes
Yes
Materials - Ageing-resistant ferritic
fine-grained structural steels
Reactor pressure vessel, main-steam and feedwater pipes - Ageing-resistant stabilised
austenitic steels
Pipes *), partly backfitted by replacements, in addition reactor pressure vessel internals and cladding Application of the break
preclusion concept
Post-qualification partly through pipe replacement
Prior to planning;
under review **) Reduction of embrittlement from
neutron radiation exposure
Special fuel element management (low leakage loading)
*) for KRB II: only stabilised austenitic pipes are used
**) for KRB II: the break preclusion concept was approved by the competent authority with a modification licence
2. Emergency Core Cooling PWR
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4 Number of emergency core
cooling trains/capacity
4 trains of at least 50 % each
Pump head of high-pressure pumps
Approximately 110 bar
Secondary circuit shutdown in case of small leaks
Manually or fully automatic
Automatic partial shutdown
or fully automatic
fully automatic
Number of borated water flooding tanks
3 or 5 4,
in some cases twin tanks or 4 flooding pools Pump head of
low-pressure injection pumps
1 plant 8 bar 1 plant18 bar
Approximately 10 bar
Accumulators (injection pressure)
1 per loop (26 bar);
1 plant without accumulators
1 or 2 per loop (25 bar)
2 per loop (25 bar)
Sump pipe before outer penetration isolation valve
Single pipe (1 plant without
sump suction pipe)
Guard pipe construction,
some with leakage monitoring
Guard pipe construction with leakage monitoring
Place of installation of the active emergency core cooling systems
Separate building, reactor building
or annulus
Annulus
2. Emergency Core Cooling BWR
Design Characteristics Construction Line 69 Construction Line 72 Number of trains of the
high-pressure
safety injection system (capacity)
1 train
(steam turbine, up to 50 bar main steam pressure,
approx. 300 kg/s)
3 trains
(electric pumps, 3 x 70 kg/s)
Diversified
high-pressure safety injection system
1 train (electric pump approx. 40 kg/s)
No
Pressure relief 7 to 11 safety and pressure relief valves,
additionally 3 to 6 motorised pressure relief valves
11 safety and pressure relief valves,
additionally 3 motorised pressure relief valves
Intermediate-pressure injection system
No 1 train
(additional independent RHR system; electric pump, 40 bar) Number of low-pressure
emergency core cooling trains/capacity
4 trains of 50 % each 3 trains of 100 % each
Low-pressure safety system with diversified injection
1 x 100 % core flooding system
No
Backfeed from containment sump
Yes, via active systems
Yes,
via passive systems with 4 overflow pipes Place of installation of the
emergency core cooling systems
In separate rooms of the reactor building
In separate rooms of the reactor building, intermediate-pressure system
in a bunkered building
3. Containment Vessel PWR
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4 Type Spherical steel vessel with surrounding concrete enclosure, annular gap and constant internal subatmospheric pressure Design pressure
(overpressure)
1 plant 2.99 bar 1 plant 3.78 bar
4.71 bar 5.3 bar 5.3 bar
Design temperature 1 plant 125°C 1 plant 135°C
135°C 145°C 145°C
Material of steel vessel (main structure)
BH36KA;
HSB50S
FB70WS;
FG47WS;
BHW33
FG51WS;
15 MnNi 63;
Aldur 50/65D
15 MnNi 63
Wall thickness of steel vessel in the spherical region remote from discontinuities
Up to 25 mm Up to 29 mm Up to 38 mm 38 mm
Airlocks - Equipment airlock Single or
double seals without evacuation
Double seals with evacuation
- Personnel airlock Single or double seals
without evacuation
Double seals with evacuation
- Emergency airlock One with single seal
One with double
seals and evacuation
Two
with double seals and evacuation
Penetrations
- Main steam line One isolation valve outside of containment
- Feedwater line One isolation valve each inside and outside of containment - Emergency core cooling
and auxiliary systems
With a few exceptions, one isolation valve each inside and outside of containment
One isolation valve each
inside and outside of containment - Ventilation systems One isolation valve each inside and outside of containment
3. Containment Vessel BWR
Design Characteristics Construction Line 69 Construction Line 72
Type Spherical steel vessel
with pressure suppression pool located in the thorus
Cylindrical pre-stressed concrete shell with annular pressure suppression pool Design pressure
(overpressure)
Up to 3.5 bar 3.3 bar
Design temperature Approximately 150°C
Material of steel vessel (main structure)
WB25; Aldur50D, BHW25 TTSTE29
Wall thickness of steel vessel outside the concrete support
Depending on geometry and design:
18 mm to 50 mm, 18 mm to 65 mm, 20 mm to 70 mm, 25 mm to 70 mm
8 mm steel liner
Number of pipes in the pressure suppression pool
Depending on the plant:
58, 62, 76 or 90
63
Immersion depth of pipes in the pressure suppression pool
2.0 or 2.8 m 4.0 m
Inertisation of the air in the pressure suppression pool
Yes Yes
Inertisation of the drywell Yes No
Airlocks In all cases double seals with evacuation
- Equipment airlock None
- Personnel airlock Leading to control rod drive chamber, for personnel and for equipment transports
- Emergency airlock One from control rod drive chamber One from control rod drive chamber and one above pressure suppression pool
Penetrations - Main steam line/
Feedwater line
One isolation valve each inside and outside of containment
- Emergency core cooling and auxiliary systems
Emergency core cooling system in the area of the pressure suppression pool and several small pipes with two isolation valves
outside of containment, otherwise one isolation valve each inside and outside of containment - Ventilation system Two isolation valves outside of containment
4. Limitations and Safety Actuation Systems PWR
4.1 Limitations
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4 Reactor power limitation 1 plant yes,
1 plant no
Yes
Control rod movement limitation
Yes
(monitoring of shut-down reactivity ) Limitations of coolant
pressure, coolant mass and temperature gradient
Coolant pressure
Partially Yes
4.2 Safety Actuation Systems
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4 Actuation criteria derived
from accident analysis
Largely, yes Yes
Different physical actuation criteria for reactor protection system
Yes, or higher-grade
redundancy
Yes, or
diverse actuation channels
Failure combinations Random failure, systematic failure,
consequential failures, non-availability due to maintenance Testing of reactor protection
system is possible during power operation
Yes, largely by automatic self-monitoring (of functional readiness)
Actuation of protection systems
Apart from a few exceptions, all actions are performed automatically, and manual actions are not required within the first 30 min
after the onset of an accident.
4. Limitations and Safety Actuation Systems BWR
4.1 Limitations
Design Characteristics Construction Line 69 Construction Line 72
Fixed reactor power limitation Yes,
speed reduction of forced-circulation pumps Variable reactor power limitation Yes,
control rod withdrawal interlock start-up interlock of forced-circulation pumps
Local power limitation Yes,
control rod withdrawal interlock
Yes,
control rod withdrawal interlock and speed reduction of forced-circulation pumps
4.2 Safety Actuation Systems
Design Characteristics Construction Line 69 Construction Line 72 Actuation criteria derived
from accident analysis
Largely, yes Yes
Different physical actuation criteria for reactor protection system
Yes, or
higher level of redundancy
Yes, or
diversified actuation channels
Failure combinations Random failure, systematic failure,
consequential failures, non-availability due to maintenance Testing of reactor protection
system is possible during power operation
Yes, largely by automatic self-monitoring (of functional readiness)
Actuation of protection systems Apart from a few exceptions, all actions are performed automatically, and manual actions are not required within the first
30 min
after the onset of an accident.
5. Electric Power Supply PWR
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4 Number of independent
off-site power supplies
At least 3
Generator circuit breaker Yes
Auxiliary station supply in the case of off-site power loss
Yes, load rejection to auxiliary station supply
Emergency power supply 2 trains with 3 diesels altogether, or
4 trains with 1 diesel each
4 trains with 1 diesel each
Additional emergency power supply for the control of external impacts
2 trains 1 - 2 trains, unit support system at
one double-unit plant
4 trains with 1 diesel each
Uninterruptible DC power supply
2 x 2 trains 4 trains (except for 1 plant with 2 x 4 trains)
3 x 4 trains
Protected DC power supply 2 hours
Separation of trains Intermeshed emergency power supply,
physical separation of the
emergency power supply
grids
Partially intermeshed emergency power
supply, physical separation of the emergency power
supply grids
Largely non-intermeshed emergency power supply, physical separation of the emergency power supply grids
5. Electric Power Supply BWR
Design Characteristics Construction Line 69 Construction Line 72 Number of independent
off-site power supplies
At least 3
Generator circuit breaker Yes
Auxiliary station supply in the case of off-site power loss
Yes, load rejection to auxiliary station supply
Emergency power supply 3 or 4 trains with 1 diesel each
5 trains with 1 diesel each Additional emergency power
supply for the control of external impacts
2 or 3 trains with 1 diesel each
1 - 3 trains with 1 diesel each
Uninterruptible DC power supply 2 x 2 trains 2 x 3 trains
Protected DC power supply 2 hours
Separation of trains Partially intermeshed emergency power supply,
physical separation of the emergency power supply grids
Largely non-intermeshed emergency power supply, physical separation of the emergency power supply grids
6. Protection against External Impacts PWR
Design Characteristics Construction
Line 1 Construction
Line 2 Construction
Line 3 Construction Line 4 Earthquake Design of components important to safety
in accordance with site-specific load assumptions Aircraft crash and pressure
waves from explosions
Not considered in the design,
later risk assessment,
separate emergency
systems
Different designs, separate emergency
systems
Design in accordance with the nuclear safety regulations
(→ Article 17 (i)),
emergency systems integrated in the safety system
6. Protection against External Impacts BWR
Design Characteristics Construction Line 69 Construction Line 72 Earthquake Design of components important to safety
in accordance with site-specific load assumptions Aircraft crash and pressure
waves from explosions
Different designs, up to status of construction line 72, emergency systems separate, or
integrated in the safety system
Design in accordance with the nuclear safety regulations
(→ Article 17 (i)), emergency systems integrated in the safety system