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Nordiskkerne-sikkcrheds-forskning
Nordiskkarn-sSkerhets-forskning
Pohjoismainenydin-turvaliisuus-lutkimus
Nordicnuclearsafetyresearch
RAK-2
DK9800001
NKS/RAK-2(97)TR-C4ISBN 87-7893-016-2
Description of Sizewell BNuclear Power Plant
Geir MeyerEgil Stokke
Institutt for energiteknikk (IFE)OECD Halden Reactor Project
Halden, Norway
September 1997
n h
Abstract
The intention of this report is to present a condensed technical description of Sizewell Bin a language understandable to non-technical personnel. It is unavoidable that someparts will be less precise than the technically initiated would like to see, but hopefullythe content still give a realistic picture of Sizewell B.
The technical description is based on publicly available material, of which the SizewellB safety report has been particularly useful. Nearly all figures and drawings found inthis description are reproductions of corresponding material in the safety report.
To keep the description from becoming too voluminous it has been necessary tocondense some background material down to a small volume. Hopefully this has notintroduced any errors or inaccuracies, possible oversimplification at certain points mustbe weighed against the wish to cover most of the topics in the agreed table of contentsfor these NKS reports.
The authors would like to thank Nuclear Electric pic and Sizewell B staff for thegoodwill they showed us and the background material made available.
The description reflects the authors understanding of the issues and is entirely theirresponsibility. The report has a limited distribution in accord with the agreeddistribution list, the authors take no responsibility for any unauthorised use of the reportor its contents.
NKS/RAK-2(97)TR-C4ISBN 87-7893-016-2
Information Service Department, Ris0, 1997
The report can be obtained from:NKS Secretariat Phone: +45 4677 4045P.O.Box 49 Fax: +45 4677 4046DK-4000 Roskilde http://www.risoe.dk/nksDenmark e-mail: [email protected]
TABLE OF CONTENTS
1. INTRODUCTION 1
2. SUMMARY OF DESIGN DATA 2
2.1 General design features 22.2 Comparison with similar reactor types 3
3. SITE AND REGION 4
3.1 Selection of the site 43.1.1 Geographical location 43.2 Description of the site 43.2.1 Site use and topography 43.2.2 Geology 4
4. SAFETY CRITERIA 5
4.1 Safety related design criteria 54.2 Classification of structures, systems and components 54.3 Conditions of design 64.4 Missile protection 64.5 Fire protection criteria 6
5. TECHNICAL DESCRIPTION AND DESIGN EVALUATION OF SYSTEMS,STRUCTURES AND COMPONENTS 75.1 Plant arrangement 75.2 Building and structures 75.3 Reactor core and other reactor vessel internals 95.3.1 Mechanical design 115.3.2 Nuclear design 135.3.3 Thermal and hydraulic design 155.4 Reactivity control systems 165.5 Reactor main coolant system 165.5.1 System information 165.5.2 Reactor vessel 175.5.3 Reactor coolant piping 195.5.4 Reactor coolant pumps 215.5.5 Steam generators 245.5.6 Pressuriser 275.5.7 Pressuriser relief tank 305.5.8 Safety and relief valve systems 305.5.9 Valves 315.6 Residual heat removal systems 315.7 Emergency core cooling systems 335.8 Containment systems 355.8.1 Overall system information 35
5.8.2 Containment structure 355.8.3 Containment penetrations 355.8.4 Containment isolation system 355.8.5 Pressure reducing and heat removal systems 355.8.6 Containment gas control system 375.8.7 Secondary containment 375.9 Steam and power conversion systems 375.9.1 Overall system information 375.9.2 Turbine-generator 375.9.3 Main steam supply system 395.9.4 Main condensers and evacuation system 395.9.5 Turbine gland sealing system 415.9.6 Turbine bypass system 415.9.7 Condenser cooling system 415.9.8 Condensate and feed water system 425.9.9 Condensate cleanup system 435.9.10 Steam generator blowdown system 445.9.11 Safety and relief valves 455.9.12 Other turbine auxiliary systems 455.10 Fuel and component handling and storage systems 455.10.1 New fuel storage 455.10.2 Spent fuel storage 455.10.3 Handling and inspection systems 455.11 Radioactive waste systems 465.11.1 Liquid waste systems 465.11.2 Gaseous waste systems 505.11.3 Solid waste systems 515.12 Control and instrumentation systems 525.12.1 Overall system information 525.12.2 Protection system 535.12.3 Regulating system 545.12.4 Instrumentation system 555.13 Electrical power systems 555.13.1 Main transformer and connected equipment 555.13.2 Plant distribution system 555.13.3 Standby power supply 55
6. FIRE PROTECTION 57
6.1 Buildings, layout and materials 576.2 Fire-fighting equipment 57
7. PLANT PERFORMANCE DURING NORMAL OPERATION 58
7.1 Phases of normal operation 587.2 Plant statistics 58
II
8. ACCIDENT ANALYSES 598.1 Malfunctions in different systems 598.1.1 Internal and external hazards 60
9. RADIATION PROTECTION 61
9.1 Basic radiation protection criteria and derived working rules 619.1.1 Dose limits 619.2 Shielding 629.3 Radiation 629.3.1 Radiation levels 639.3.2 Radiation monitoring 649.4 Monitoring systems and action levels for radioactive releases to the environment 649.4.1 Monitoring of releases to the atmosphere 649.4.2 Monitoring of releases to water 64
10. OFFSITE DOSE ASSESSMENT 65
10.1 Releases under operating conditions 6510.1.1 Radiation doses 6510.2 Releases under hazardous conditions 6510.2.1 Contamination of the environment 6510.2.2 Radiation doses 66
11. PLANNING, ORGANISATION AND ADMINISTRATIVE CONTROL 67
11.1 Production department 6711.1.1 Management support department 6711.1.2 Technical support department 6811.1.3 Training programme 6811.1.4 Operating procedures, instructions and orders 68
12. ORGANISATION OF THE AUTHORITIES 69
12.1 Local organisation 69
13. PROBABILISTIC SAFETY ASSESSMENT 70
14. REFERENCES 71
III
LIST OF FIGURES
Figure 1, Diaphragm wall 2Figure 2, Site layout 7Figure 3, Main plant layout 8Figure 4, Core arrangement during cycle 1 10Figure 5, Reactor vessel 12Figure 6, Control rod drive mechanisms 13Figure 7, Fuel assembly 14Figure 8, Fuel rod 75Figure 9, Reactor coolant system 17Figure 10, Reactor vessel and internals 18Figure 11, Reactor coolant main pipework - plan 20Figure 12, Reactor coolant main pipework - elevations 21Figure 13, Reactor coolant pump 22Figure 14, Reactor coolant pump support 23Figure 15, Steam generator 24Figure 16, Steam generator support 25Figure 17, The pressuriser. 27Figure 18, Reactor pressure control system 28Figure 19, Pressuriser supports 29Figure 20, Safety injection, residual heat removal and spray system 32Figure 21, Engineered safety features 34Figure 22, Reactor building heat removal systems 36Figure 23, Turbine, steam and condensate system 38Figure 24, Main condenser. 39Figure 25, One leg of the main steam system 40Figure 26, Main circulating water system 41Figure 27, Main feed water system 43Figure 28, Steam generator blowdown system 44Figure 29, Fuel route and storage 46Figure 30, Liquidwaste channel A and B, simplified 47Figure 31, Liquid waste system "Active laundry and hot shower" plus "Chemical drains tank ", both simplified48Figure 32, The secondary liquid waste system (simplified) 49Figure 33, The gaseous radwaste system (simplified) 50Figure 34, The solid waste processing (simplified) 51Figure 35, Defence-in-depth 52Figure 36, Instrumentation and control system 55Figure 37, Primary Protection System (PPS) and the protection computer system 54Figure 38, Main and essential electrical systems 56
IV
LIST OF TABLES
Table 1, Reactor core 9Table 2, Fuel 9Table 3, Reactor vessel details 19Table 4, Reactor coolant pipes details 20Table 5, Reactor coolant pumps details 23Table 6, Steam generator details 26Table 7, Pressuriser details 30Table 8, Residual heat removal pumps/low head safety injection pumps 31Table 9, High head safety injection pumps detail 33Table 10, Accumulators details 33Table 11, Refuelling water storage tank details 34Table 12, Spray pumps details 37Table 13, Turbine conditions 37Table 14, Circulating water pumps 42Table 15, Main feedwater pumps 42Table 16, Auxiliary feedwater pumps 42Table 17. Plant statistics 55Table 18, Dose limits 61Table 19, Direct radiation exposure of station staff. 62Table 20, Annual collective dose 63Table 21, Highest individual task dose 64Table 22, Fault frequency targets 66
1. INTRODUCTION
The Sizewell B Nuclear Power Station is the first Pressurised Water Reactor (PWR) built in UK.Originally planned as the first in a series there is at present uncertain whether there will bemore PWRs built in UK in the near future, and Sizewell B will remain the only PWR in UK forsome years to come.
The design of Sizewell B started in 1980s, based on the Westinghouse/Bechtel SNUPPS concept(Standardised Nuclear Unit Power Plant System). It has thus been possible to draw on theconsiderable knowledge and experience accumulated from design and construction of nuclearpower plants utilising this technology. The reference plant for the original design is theCallaway plant in Missouri, USA, but experience derived from the French nuclear powerprogram has also been taken into account in the construction of Sizewell B.
The planning/design period stretched over a number of years, in no small part due to a prolongedpublic debate over the safety and economy of nuclear power. Changes and additions to theoriginal design were introduced to satisfy very stringent demands to reliability and to lower theprobability for failure of safety significant systems. Safety analyses of the as-built NPP showsthat even with very conservative assumptions the risk to public health from credible accidents isacceptably low.
Construction orders for Sizewell B were placed in 1987, construction started 1988. Firstcriticality was achieved in March 1994, commercial operation in February 1995.
2. SUMMARY OF DESIGN DATA
2.1 General design features
Sizewell B is a 4-loop PWR of Westinghouse type with two turbines. Thermal capacity is 3411MWt, rated net electrical output is 1188 MWe (gross 1258MWe). The nuclear steam supplysystem and turbines are enclosed within a prestressed concrete containment structure of doubleshell type.
The reference plant for the original design is Callaway NPP, Missouri, USA. To fulfil UKrequirements the original design had to be modified, particularly as concerns capacity andredundancy of safety systems.
A special design feature is the subterranean diaphragm wall enclosing the central parts of theplant (reactor building, turbine plant, radwaste building etc. The wall is 0.8 m thick and goesdown to the London clay stratum (40-50 m below surface). It greatly facilitated construction ofthe plant as it effectively restricted the influx of water without lowering the ground water leveloutside the diaphragm wall. Thus there were no problems with settling of foundations whichcould have created problems for the adjacent Sizewell A, neither have there been any adverseeffects on agriculture from sinking ground water level. Presumably the diaphragm wall will alsobe an efficient barrier against contamination of groundwater in case radioactivity is released tothe ground within the plant.
DIAPHRAGM WAIL LAYOUT
Figure 1, Diaphragm wall
2.2 Comparison with similar reactor types
The Sizewell B plant is by design quite similar to recently constructed SNUPPS plants andnewer French NPPs. The reactor vessel is made by Framatome Energy, instrumentation andcontrol systems are largely by Westinghouse, and turbine generators are delivered by GECAlsthom.
The SNUPPS concept is based on Westinghouse primary circuit and steam supply systems, thestandardised plant layout is by Bechtel Power Corporation (USA). The reference plant isCallaway (Mo., USA), but the reference design has been modified to meet British standards andrequirements. Especially the design of safety systems and support systems important for safetyhas been altered to enhance reliability and performance. The number of high head injectionpumps has been increased from two to four, the capacity of accumulators is higher, additionalpumps are installed in the auxiliary feedwater system, and there are altogether four dieselgenerators feeding into four separate groups of safeguards equipment.
Sizewell B is also equipped with two separate shutdown systems, the Primary Protection Systemand the Secondary Protection System. Both function as independent systems able to shut downthe reactor and initiate start-up of safeguards plant.
The containment is of the large dry, double shell type, the primary containment design pressureis 4.46 bar (overpressure 3.4 bar). Free volume is app. 90 000 m3, this is one of the largestcontainments in the industry.
EdFs N4 series of nuclear stations is in many respects close to Sizewell B design. There has beenclose contact between Nuclear Electric and other utility organisations to derive maximum benefitfrom experiences made by the nuclear industry world-wide. EdFs Chooz Bl is the first of the1400 MWe N4 stations in operation, its instrumentation and control system is in principle fullydigitalized though there is a hardwired safety auxiliary panel which can be used for safeshutdown and controlling shutdown conditions. The experiences collected from operation ofSizewell B and Chooz B in the coming years should provide valuable input to the design of NPPsand to assessment of the reliability of new I & C technology.
3. SITE AND REGION
3.1 Selection of the site
The criteria for site selection of nuclear power plants firstly take into account the safety of thepublic and environmental factors. An ample supply of cooling water is of particular importancefor a large PWR. Economy and closeness to major electricity consumer areas are further aspectsto be considered when safety principles related to siting are satisfied.
The siting of a nuclear power station at Sizewell in East Suffolk was proposed in 1958, and aMagnox NPP (Sizewell A) with two reactors was constructed and put into operation in 1966. In1980 CEGB (Central Electricity Generating Board) proposed that the first PWR in Britainshould be constructed at the site close to the Magnox station. A public inquiry (1983-85) delayedthe final decision, consent to build was given in 1987.
3.1.1 Geographical location
Sizewell B is located in the same area as Sizewell Al and A2, two Magnox reactors of 250 MWethat started operation in 1966. The location is on the south-east coast of England (East Suffolk),geographical co-ordinates 52°12'N, 01°37'E. The closest cities are Ipswich (-30 km) and Norwich(-50 km), the distance from London is approximately 150 km.
3.2 Description of the site
The coastal area is flat and the land surrounding the station is partly marshland, partlyfarmland with some conifer plantations. The site is on the shoreland, sand covering most ofground down towards the shore on the east side.
3.2.1 Site use and topography
Sizewell B is situated adjacent to the Sizewell A plant, this makes for efficient utilisation of landarea but may have some adverse effects such as resettling of Sizewell A foundations. Thisrequired protective measures, the most striking being the construction of a 0.8 m thickdiaphragm wall extending 35-56 m down from ground level. The site is quite flat, and thediaphragm wall encloses all the main structures of process plant and radwaste building, totallength of wall is 1260 m.
3.2.2 Geology
There is a top layer of sand 40 to 50 meters thick, underneath there is a layer of London claywhich is nearly impermeable to water. Further down there are new layers of sand, clay and thenchalk at about 80 m below ground level.
4. SAFETY CRITERIA
The safety criteria for nuclear power plants in Britain must comply with the requirements setdown in the Nuclear Installations Act (1965) and the Health and Safety at Work Act (1974). Thedesign guidelines applied include certain numerical targets quoted below.
4.1 Safety related design criteria
The overriding concern is the safety of the public, this is transformed into requirements to befulfilled concerning the probability of accidents which may lead to health risks from radiationexposure. To limit the risk to the public the following numerical targets are set regarding theprobability of accidents:
• The total frequency of all accidents leading to uncontrolled releases should be less than 10"6
per reactor year.
• The overall design shall ensure that the frequency of any single accident that could lead to alarge uncontrolled release of radioactivity shall be less than 10'7 per reactor year.
o The predicted frequency of accident that could result in doses of one Emergency ReferenceLevel (ERL) should not exceed 104 per reactor year.
An example of ERL dose is a 100 mSv whole body dose to a member of the public.
In order to satisfy these criteria the reliability of engineered safeguard features has to becorrespondingly high. The shutdown system, emergency systems and the protective barriersmust be shown to independently function such that the above targets are met for any credibleaccident.
UK design criteria include the 30 minute rule,- no operator action shall be required for at least30 minutes after a design basis event sequence is initiated. In essence this means thatautomatics and safeguard systems are able to adequately control the power plant and ensurethat there are no uncontrolled releases during that period.
4.2 Classification of structures, systems and components
Building structures are classified in 3 categories according to their safety function:
• Category 1.Structures whose function or integrity is required to allow the reactor to be safely shutdownand cooled after a safe shutdown earthquake and which are designed to remain operationalafter that event. (Safe shutdown earthquake: Peak horizontal acceleration of 0.25 g.)
• Category S.Structures whose integrity or function is not directly required for nuclear safety, but whichare designed to behave, under safe shutdown earthquake loading, in a manner which wouldnot impair the function of Category 1 structures, systems or components.
• Category N.Structures which are not seismically designed.
4.3 Conditions of design
The overall safety goals, some of which are quoted above (4.1 Safety related design criteria) mustbe reflected in the design principles and translated into specific requirements at plant andsystem level. To be able to assess the safety of the plant as such it is necessary to perform acomprehensive analysis of all part systems where the effects of systems interactions are fullytaken into account. A Probabilistic Safety Assessment (PSA) that covers all protective ormitigating measures in the event a fault occurs is an effective tool for identifying weaknesses indesign and elucidating the interplay of subsystems.
An extensive probabilistic safety analysis has been performed for Sizewell B and is included inthe Pre-Operational Safety Report submitted to the Nuclear Installations Inspectorate (Nil) inNovember 1992. The analysis covered internal and external initiators, beyond design basisinitiating faults and non-core sources of radioactivity. Analysis of shut-down conditions were alsoincluded. A brief description of the Sizewell B PSA is found in section 13 Probabilistic safetyassessment.
4.4 Missile protection
The main barrier in missile protection is the containment structure, internal missiles mayconstitute a hazard but here only external missiles are discussed.
A directed missile or an aeroplane would have to severely damage or penetrate the containmentto cause a major release of radioactivity. A breach of containment does not necessarily imply thatthe primary circuit or other radioactivity confining barriers will be damaged. For accidentsoutside the containment, say damage to the control room or the turbine plant the shutdown andprotection systems should ensure that the reactor is brought to a stable condition.
Sizewell B has a double containment, the inner principal containment is a 1.3 m thickprestressed concrete cylindrical shell surmounted by a 1 m thick hemispherical shell. Theinternal steel liner is an airtight membrane, it also aids in distributing the load over a largerarea in case of developing cracks or fissures. Design pressure is 4.46 bar, tests on scaled-downmodels has shown that the more than twice this pressure is needed for the containment to fail.In view of the rugged structure of the containment the station should be well protected againstexternal missiles.
4.5 Fire protection criteria
The fire protection criteria aim firstly at prevention, secondly at fire control and mitigationshould a fire occur.
The design provides for 3 hour principal fire barriers to be built around redundant and diversetrains of safety systems. In addition it provides for further 1 hour fire barriers within theprincipal barriers, in some cases the additional fire barriers are 3 hour rated fire barriers.
5. TECHNICAL DESCRIPTION AND DESIGN EVALUATION OF SYSTEMS,STRUCTURES AND COMPONENTS
5.1 Plant arrangement
Figure 2, Site layout shows the layout of the plant site. The site is about 500 m times 500 m, withthe reactor, turbine and radwaste buildings in the central part. The cooling water canals and thepump house are on the east side of the site.
5.2 Building and structures
Figure 3, Main plant layout identifies the main buildings within the site. Of particular interest isthe containment structure which encloses the primary circuit and the steam generators. SizewellB has a double-walled containment, the primary (inner) containment of prestressed concretebeing the main protective barrier. It is designed to withstand an overpressure of 0.345 MP (-3.4bar), a steel liner ensures airtightness. Pressure tests have shown that leakage is only about 25%of the prescribed limits, the elasticity of the containment is also as expected.
'Sea 5d»*KoV 1. Condeneate Tank A
I . Reactor Mcke-Up WoUr Tank3. Refuelling Water Storage Tank4. Hoik Prep Bay ttiter Storoje Tank5. Condensate Tank 8t. DxntarcEwl Water Tank7. Emergency ClwrglnQ System Storage Tat*& Generator Transformer 1t . Generator Transformer i
10. AuiSoiy Bohr Home11. HypocNnft* «w»ro11o« Bulding12. OlffiUl Oufllkd Store11 Water Ircalmml Plant14. Fin FlghGng Punplmaa15. f W figliUig Furi 01 TonkaIt, Hxtm ert)g> (Rod«Bt>/D«»!iitom)17. AunUoiy Trondomwr11 Station TransformerIB. Unit TlWKformerMl U m t U K<Hl TranfoniMr31. Curator Mtciorcr22. Muclor/tS CoolerJJ. ttxttwn Anneai24. HypochkxHa Sen Pionl S»9digaar Bulitng
ti. Hyijtao™ om Ammoaio Stora27. Cnatllc and Sdfrwie AcU Store25. Aujdlory Boleri TvM OH TwikeM . C02 Stan3a. W sun31. HP Cm Bottto Stan32. O n Settle Store3J. U t to Beat Store34. Uonorrw Drum Store35. Cuentiol Keeel Bd&ig34. AujUoiy SruUom SuiMiig37. SotMy Crargjng Die»<M Ri39. Dfcwl Fuel CM Tonke40. ft. Station
38
Figure 2, Site layout
re
n
I
00
Fuel buildingNew fuelFue! storePlant access hatchSecondary containmentSteam generators (four)Reactor pressure vesselSteam mainsVentilating plantReactor coolant pumpsPressuriserAuxiliary buildingControl buildingMain control roomFeed pumps (six)DeaeratorAuxiliary boiler houseTurbo-generatorLow pressure heatersAuxiliary plantLoading bayHigh pressure heatersLubricating oil plantReheatersPolishing plantSwitchgearCirculating water outletCirculating water inletTransformersStation transformersGenerator transformer
5.3 Reactor core and other reactor vessel internals
Apart from the fuel assemblies the core components consists of the rod cluster control assemblieswhich are used for core reactivity control during normal operation and for reactor shutdown, theburnable poison assemblies which are used for temporary reactivity control to compensate forthe excess reactivity of the new fuel, the primary and secondary source assemblies which givesufficient count rates on the neutron monitors outside the core at shutdown and the thimble plugassemblies which restrict coolant bypass leakage flow. It is only the rod cluster controlassemblies that are moveable in operation. See Table 1, Reactor core for details on the core andTable 2, Fuel for details on the fuel.
The rod cluster assemblies consists of 53 control assemblies each containing 24 absorber rodswhich again consists of two silver-indium-cadmium alloy rods encapsulated in a stainless steeltube. The assemblies are grouped into three control banks and six shutdown banks, see Figure 4,Core arrangement during cycle 1.
Number of fuel assembliesActive heightEquivalent diameterMass of UO2 in coreControl rod absorber materialNumber of assembliesAbsorber rods_per assemblyFuel rod heat rating averageFuel rod heat rating_maximumModeratorCore heatHeat transferred to steam
1933.66 m3.37 m101 te
Ag-In-Cd5324
17.8 kW/m41.3 kW/m
reactor coolant water (H2O)3411 MW3428 MW
Table 1, Reactor core
Fuel pellets
Fuel can
Fuel assemblies
i MaterialTDensitv1 Enriched (feed fuel) %! Material.Outside diameteri Thickness• Basic rod array• Fuel rods ger assemblyi Rod pitch
sintered UO2
9 5 %3.1
Zircaloy 49.5 mm
0.57 mm17 x 17
26412.6 mm
i_Nurnber_of_guide tubes for absorber; 24• Number of guide tubes for instrumentation 1i Number of grids 8
Table 2, Fuel
mm•S C
S 3
8 8
3 D
mm•KBSBiBiSg
illt l!
S D
, , , , „ • •
8E |
5
!
mmS C
X
Y
S F
Y
111SB
I 9B
P"Y
8 E
8 F
Z i
S F
.... . . !
j
SB
SB
X
Y
SF
Y
-
•SC
8D
mm
-
8 E
mm
HIII: «•f
SD |
SB I
[SB 1
•Rod olustor oontrol
X X-bank locationY Y-btnk locationZ 2-banfc location
SA A-bank EooationSB B-b«nk locationSC C-b*nk location8O Oteonk toeatlenBE E-bank tocotlonSF F-b«nk location
2 .1% (weight) •nrlchmant
2.6% CwalQftQ •rtftohmorrt
3 .1% (W«lght> •nrtchmeni
Figure 4, Core arrangement during cycle 1
The burnable poison assemblies are not used to any extent in fuel reloads but are requiredduring the first load consisting of fresh fuel only.
In the first core there are eight primary source assemblies which are identical to the burnablepoison assemblies except that a primary source rod consisting of californium-252 within astainless steel tube is replacing one of the thimble plug rodlets. Thereafter there are eightsecondary source assemblies replacing the primary source assemblies. Each secondary assemblyconsists of four secondary source rods consisting of a stack of pellets of 78% antimony and 22%beryllium by weight in a stainless steel tube.
A thimble plug assembly consists of 24 thimble plug rodlets which are made of a solid rod ofstainless steel, suspended from a holddown assembly or top plate.
10
5.3.1 Mechanical design
The vessel is made of ring forgings and at the heads single-piece dome forgings are used, seeFigure 5, Reactor vessel for details.
The drive mechanism is an electromagnetic jacking device moving the rods in small steps. If theelectric current is interrupted the control rods will drop by gravity into the core. All partsexposed to reactor coolant are made from three types of material; stainless steels, nickel-chromium-iron alloys and cobalt-based alloys. All pressure-retaining parts are made fromstainless steel or austeitic nickel-based alloys. Hard chrome plate and Stellite-6 have been usedfor bearings and surfaces subjected to wear. See Figure 6, Control rod drive mechanisms for howthe rod drive mechanism is arranged.
The absorber rods are kept in place by a spring at the top of the tube which is sealed at the topand bottom by stainless steel end plugs.
The poison assemblies consists of a number of burnable poison rods suspended from a hold-downassembly.
Both primary and secondary neutron source assemblies are located at the correct height forprimary/secondary protection system source range detectors or postfault monitoring neutrondetectors.
Penetration tubes are welded to the reactor vessel bottom head to allow for in-core flux mapping.The temperature of some fuel assemblies are measured by thermocouples above the core.
11
Figure 5, Reactor vessel
12
Missile Shield Seismic Sleeve
ElectricalConnector forLeadstoCRDMConnection
Rod PositionDetector Coils
ThermocoupleColumn Housing
Figure 6, Control rod drive mechanisms
5.3.2 Nuclear design
The core contains 193 fuel assemblies where the first fuel cycle have three fuel regions; 65assemblies with 2.1% by weight uranium-235, 64 assemblies with 2.6% by weight uranium-235and 64 assemblies with 3.1% by weight uranium-235. For details, see Figure 4, Corearrangement during cycle 1.
13
The fuel assemblies are built as a 17 x 17 square array with 264 fuel rods in each assembly, seeFigure 7, Fuel assembly. Each fuel rod is made of a Zircaloy-4 tube filled by a stack of fuelpellets, see Figure 8, Fuel rod. The pellets are small cylinders of uranium dioxide slightlyenriched in the uranium-235 isotope with a length of just over 13 mm and a diameter of around 8mm. A fuel rod is around 3.9 m high and 9.5 mm in diameter
Figure 7, Fuel assembly
14
.Top end plug
. Hold down spring
Plenum
. Fuel pellet
Fuel cladding
. Bottom end plug
Figure 8, Fuel rod
5.3.3 Thermal and hydraulic design
The core heat is 3411 MW and heat transferred to steam is 3428 MW (including pump losses).
Coolant inlet temperature is 292.4°C and outlet temperature is 323.4°C. The coolant pressure atinlet is 158.3 bar a and at outlet 155.1 bar a. The coolant flowrate is 19.2 te/sec and the volume ofwater in primary circuit is 334.5 m3. Of the coolant flow approximately 5.8% bypasses the core, ofwhich 2% is associated with the fuel assembly thimbles and 3.8% is associated with the internalstructures.
15
5.4 Reactivity control systems
Reactivity is controlled by the inherent negative power feedback (fuel temperature andmoderator temperature), by control rods and boric acid injection systems. The emergencyboration system contains a boron solution with a concentration of 7000 ppm.
The maximum time between the release of the drive rod and to the rods are fully inserted is 2.2seconds.
5.5 Reactor main coolant system
5.5.1 System information
The reactor coolant system consists of the reactor pressure vessel, four steam generators, fourreactor coolant pumps, the reactor internal structures and the reactor pressure control system,see Figure 9, Reactor coolant system. The normal flowrate is 19114 kg/s with all four pumpsrunning. At power levels below 15% the reactor is manually controlled.
At least two reactor coolant pumps are kept operable with one in operation during hot shutdownconditions to ensure uniform boron concentration, and the coolant system pressure andtemperature are held at no-load conditions.
The chemical and volume control system is connected to the cold legs of coolant loops 1 and 2 andto the reactor coolant pump seals. The letdown paths from the reactor to the system is from loop3 and 4 crossover leg. The make-up water system is connected via the chemical and volumecontrol system.
Suction for the residual heat removal system is taken from the hot legs of loop 1 and 4 andcoolant is returned to all four cold legs.
Each of the four accumulators of the emergency core cooling system is connected to a separatecold leg.
The emergency boration system is connected to each of the four cold legs and to each of the fourcrossover legs.
The nuclear sampling system is connected to hot leg of loop 1 and 3 and to the water space in thepressuriser.
16
Figure 9, Reactor coolant system
5.5.2 Reactor vessel
The vessel, see Figure 10, Reactor vessel and internals, has been made from ring forgings withsingle-piece dome forgings for the spherical sections of the upper and lower heads with inlet andoutlet nozzles symmetrically located around the vessel. The upper head can be removed and hastwo grooves cut in the flange to house hollow metallic O-rings and is penetrated by a vent lineand by Inconel-600 tubes used in the rod cluster control assembly drive mechanisms or forinstrumentation adapters. There is 58 penetrations for access for core instrumentation in thedome forging in the lower head of the vessel. All internal surfaces are clad with corrosionresistant stainless steel. The core barrel is a 7.87 m high cylinder with an internal diameter of3.76 m and with a thickness of 57 mm. The upper core support is a 305 mm thick forging and theupper core plate is 76 mm thick. All major components are fabricated from stainless steel with
17
some parts made from austenitic nickel-based alloys. The cobalt content of the stainless steel hasbeen limited to reduce the levels of activity in the primary coolant.
Thermocouple Column
Control Rod DriveMechanisms
Thermal Sleeve
Drive Rod
Hold Down Spring
Support Ledge
Upper Core Support
Hot Leg
Alignment Block
Lower LateralSupport Clevis Block
Rod Cluster ControlAssembly Guide Tubes
Alignment Pin
Cold Leg
Core Barrel
Reactor Vessel
Neutron Shield Pad
Lower Core Support
InstrumentationPenetration a'ndThimble Guide Tubes
InstrumentationThimbles
Figure 10, Reactor vessel and internals
18
Overall height
Inside diameter
Total thickness (opposite the core)
Material
Internal cladding
Cladding thickness
Inlet nozzle inside diameter
Outlet nozzle inside diameter
Dry weight
15.39 m
4.394 m
220 mm
low alloy steel
stainless steel
7 mm
704 mm
736 mm
435 te
Table 3, Reactor vessel details
5.5.3 Reactor coolant piping
The main reactor coolant system pipework consists of the outlet piping from the reactor, calledhot leg pipes, the piping linking the steam generator to the reactor coolant pumps, calledcrossover or intermediate leg, the reactor inlet piping, called the cold leg, and the pipeworkconnecting to the pressuriser. See Figure 11, Reactor coolant main pipework • plan, Figure 12,Reactor coolant main pipework - elevations and Table 4, Reactor coolant pipes details for details.
The main coolant pipes are made of seamless straight lengths of pipe and statically cast bends.The piping and fittings which make up the loop are austenitic stainless steel. The safeguardsystem and auxiliary lines are seamless straight lengths and either forged elbows or bendsforged from seamless pipe.
19
SteamGenerator
ReactorCoolantPump
ReactorPressure Vessel
Pressuriser
Figure 11, Reactor coolant main pipework - plan
Inside_djamet£r_lwtleg 7.5?. mm_
Inside_d_iamet£r_intermediate_leg^ 7j?2.5?5l
Inside_diame_ter_coldjeg ?99_mm
Inside d^a^mej;erj^essuiiser_surge_line ^l.™1!?.
Material carbon steel cladwithstainless steel
Table 4, Reactor coolant pipes details
20
Steam Generator
Hot leg
View A-A
Sieam Generator
Crossover Leg
Reactor CoolantPump
View S-B
Reactor Vessel Reactor CoolantPump
Cold Leg
\J
View C-C
Figure 12, Reactor coolant main pipework • elevations
5.5.4 Reactor coolant pumps
The pumps are single-stage, single-speed centrifugal pumps, see Figure 13, Reactor coolantpump. They are vertically mounted and driven by an above-mounted, air-cooled, three-phaseinduction motor. The pumps are each supported on three columns, see Figure 14, Reactor coolantpump support. Apart from specialised components all components are made of corrosion resistantaustenitic stainless steel. The flywheel ensures a prolonged coastdown following pump trip andensures an adequate transition between forced circulation and natural circulation in the case oftotal loss of forced circulation.
Two of the pumps are connected to one of the 11 kV unit boards and the two other to the otherboard. During start-up and shutdown at least one pump will work to prevent thermal and boronconcentration gradients.
21
Pump, motor shaft and frame vibration levels, motor bearing temperatures, oil sump levels,cooling water flow, pressure and temperatures and motor rotor and stator winding temperaturesare monitored continuously.
p a (ion device
ilbtor shaft
,. fiptpr core •
tbiuer guftfeBearing
oil cooler;
Oil liflpitmp
Air cooler
"Reirioveableisjioof-:
••Pump shaft;
3 slage seal
Radial bearing
ttierrtiai barrier;heat exchanger
pischarge nozzle Suction adapter
Suction nozzle
Figure 13, Reactor coolant pump
22
Cold leg
Tie Rods
Crossover Leg
Wide FlangeColumns
Figure 14, Reactor coolant pump support
Number ofpumps
Speed (synchronous)
Developed head
Flowrate
Coolant mass flow through jjumpjs
Pump bowl material
Motor rating_
Dry weight
Pressure rise across pump
4
1500 rpm
89.3 m
6.45 ma/sec
19114 k^sec
stainless steel
6MW
103 te
0.65 MPa
Table 5, Reactor coolant pumps details
23
5.5.5 Steam generators
Each steam generator is vertically mounted and consists of three main sections; a hemisphericalbottom head, the middle section which contains the inverted U-tubes and the upper section withmoisture separators, see Figure 15, Steam generator. The steam generator is supported by fourvertical wide flange columns, see Figure 16, Steam generator support. The steam outlet nozzlerestricts the maximum rate of heat removal in the case of a secondary side depressurization.Provisions are made to detect primary-to-secondary leakage.
The bottom head is a one-piece forging and clad with austenitic stainless steel. It is divided intotwo by a 50.8 mm thick divider plate.
The main feedwater is entering the steam generator through a 368 mm nominal bore nozzle.
;; Steam llowi***™™^-
:: i l l M
Ma ftway:r;JS-$Vv;;:MM
I:: :::::zmrUfeeffwafer inlet ^i^~$£g%
•\". nozztft::*: : ^ ^ p
En!
i i
til
if
-rr- Steam outlet; ; " ; ; -nozzle lo turbine
Hi' • 8 | r ; ' : " Steam dryers
; ^§ -^ -» Uppersihelt
|BflrtS& ditir^ udhb rt^rtf^
^ g ^separator
^ ^ p ~ ~ - Feed ring withH i 'J1. nozzles
stffspetfip pgr||
•Tiibeplatearttf-t i ;
Antiuifjratfon:b3r$:;
Qualrefoil lube:supports
-Ftow-Sistribuiioii"
.Shell.blowflown :
; Channel heatt
vParlilhin-plate
Primary coolant nozzles
Figure 15, Steam generator
24
Upper Lateral Support
Shubbers
Lower Lateral Support
Wide Flange Columns
Direction of ThermalExpansion
Figure 16, Steam generator support
25
Number
Overall height
Upper part diameter
Lower part diameter
Materials
Tubesheet thickness
U-tubes
Dry weight
Reactor coolant side
Secondary steam side
Secondary side shell
Tubes
Primary side shell
Numberr
Outside diameteri~
ThicknesshTotal heat transfer area
Inlet temperature
Outlet temperature
Flowrate
Feedwater temperaturer
Steam temperaturet~
Steam pressure
Steam flowrater
Steam condition
4
20.8 m
4.51m
3.48 m
low alloy steel
Inconel 690
low alloy steel cladwith stainless steel
631mm
5626
17.46 mm
1.03 mm
5110 m2
377 te
326.4°C
293.8°C
4.8 te/sec
227°C
284.8°C
69 bar a
477 kg/sec
0.25% wetness
Table 6, Steam generator details
26
5.5.6 Pressuriser
The pressuriser, see Figure 17, The pressuriser and Table 7, Pressuriser details, is a part of thereactor pressure control system which comprises among other the pressuriser, surge line,heaters, spray lines and valves and the pressuriser relief tank (see chapter 5.5.7), see Figure 18,Reactor pressure control system.
The pressuriser is a vertically mounted cylindrical vessel with water filling normally between25% and 60% of the volume. The pressure can be raised by using the heaters or reduced byspray, where the spray is taken from two of the cold legs or from an auxiliary spray line.
The pressuriser is supported by its skirt and four steel bumpers, see Figure 19, Pressurisersupports.
Figure 17, The pressuriser
27
Reactor coolantpi(X» (hot teg)
Pressurise?relief tank
Makeupwater
Gaseous wastamanagement
Liquid wastemanagwrmnt
Figure 18, Reactor pressure control system
28
"IPressuriserSkirt
Seismic Supports
Bearing Plate
Holding Down Bolts
Steel Support Structurein Pressuriser Cell
Figure 19, Pressuriser supports
29
Number
Overall height
Inside diameter
Material
Volumes
Heaters
Pilot-operated relief valves
Safety relief valves
Pressuriser relief tank
Totalp
Water at full power
Numberp
Total heater power
Numberl~Capacity (sat. steam at 172bar) per pair
Number
Opening pressurep
Capacity (at that pressure) pervalve
Total volumep
Normal liquid volume
Design pressure
Rupture disc burst point
1
16.1m
2.13 m
low alloy steel (clad)
51.0 m3
30.6 m3
78
1.8 MW
3
34 kg/sec
2
172.4 bar a
53 kg/sec
51m3
38 m3
7.9 bar a
6.3 bar a
Table 7, Pressuriser details
5.5.7 Pressuriser relief tank
The pressuriser relief tank is a horizontally mounted cylindrical vessel with semi-elliptical endsmade of stainless steel. It can take steam to 110% of the volume above the pressuriser waterlevel at full load conditions. Bursting discs, venting into the reactor building, have a capacitygreater than the total capacity of the pressuriser relief valves.
5.5.8 Safety and relief valve systems
Each pair of pilot operated safety relief valves is arranged in series with the first one acting as apressure relief valve and the second acting as an isolation valve. The primary protection system,see chapter 5.12.2 Protection system, monitor the reactor coolant temperature and uses the pilotoperated safety valves if the pressure is too high. At temperatures below 177°C the maximum
30
allowable pressure is reduced. Another safety relief valve is connected to the crossover leg of loop3 in case the primary protection system should fail.
The spring operated safety relief valves are arranged in parallel and operated by the fluidpressure. They acts as a diverse overpressure relief system for the pilot operated valves.
For capacities and other details see Table 7, Pressuriser details.
5.5.9 Valves
Different types of valves are used: Swing-type non-return valves for all non-return valves above50 mm, they have no body penetration; Piston lift-type non-return valves for sizes of 50 mmdiameter and below, they have a body guided piston; Gate valves with hardened seal faces, forsizes above 50 mm the stem sealing arrangement is a double packed stuffing box withintermediate lantern ring and leak-off connections; Globe valves with gland sealed stems forsizes above 50 mm; Globe valves with hermetically sealed stems for sizes of 50 mm and less;Minor valves for vent and drain lines.
Parts exposed to reactor coolant are made from stainless steel and corrosion resistant hardfacingalloys and all valves performing a safety function are duplicated within their system andpowered by segregated electrical supplies.
5.6 Residual heat removal systems
Suction is taken from the loop 1 and loop 4 hot legs, see Figure 20, Safety injection, residual heatremoval and spray system. It is a two way segregated system with two trains of equipment.
The decay heat after shutdown will typically be 1.5% of full load rate after one hour and 0.6%after one day. After around four hours on approach to cold shutdown the temperature would bebelow 177CC and pressure below 2.93 MPa and then the residual heat removal system would bestarted.
The residual heat removal system can also be used as a part of the emergency core coolingsystem, to provide back-up redundancy to the reactor building spray system, to provide aletdown path for coolant from the reactor when pressure is low, to ensure coolant flow when thereactor is shut down and the reactor coolant pumps cannot be run.
See Table 8, Residual heat removal pumps I low head safety injection pumps for some moredetails.
Number
Design flow
Design head
2 x 100 %
568 m3/hr
160 m
Table 8, Residual heat removal pumps/low head safetyinjection pumps
31
Enn
o.E.nas
o
aViT3
n
3
Key
>~4 Denotes normally closed
Denotes normaHy open
RC8 Reactor coolant »y*Um
CCW Component cooing water
RHR ReeMual heat removal
5.7 Emergency core cooling systems
The emergency core cooling system consists of the passive injection subsystem, the high headsafety injection subsystem and the low head safety injection subsystem. The chemical andvolume control system charging pumps have sufficient capacity to maintain the water followinga very small loss of coolant accident. See Figure 21, Engineered safety features and Figure 20,Safety injection, residual heat removal and spray system.
The low head safety subsystem comprises the residual heat removal system pumps and heatexchangers, see chapter 5.6 Residual heat removal systems for details.
The high head safety subsystem consists of four high head injection pumps, the reactor buildingrecirculation sumps and the refuelling water storage tank. Each pump is powered from one ofthe four trains of the essential electrical system and can be used instead of the low head injectionpumps. For details about the pumps see Table 9, High head safety injection pumps detail. Therefuelling water storage tank located outdoor contains borated demineralised water maintanedabove 21°C and has connections to the low head and high head injection pumps and the reactorbuilding spray system pumps, see Figure 2, Site layout for placement. The reactor buildingsumps are lined with carbon steel plate and are located at the bottom of the reactor building andthis is where the high head, low head and reactor building spray pumps take suction after thebulk of the refuelling water storage tank have been expended.
The passive injection subsystem consists of four nitrogen pressurised accumulator tankspartially filled with borated water connected to the cold legs via a line which includes two non-return valves and a motorised isolating valve. The opening of the non-return lines is all that isneeded for injection. For details of the accumulators see Table 10, Accumulators details.
Number
Design flow
Design head
Maximum delivery head
4 x 100 %
227 m3/hr
964 m
1250 m
Table 9, High head safety injection pumps detail
Numbers
Total volume per accumulator
Water volume_per accumulator
Nitrogen overpressure
Material
4 x 50%
57.3 m3
36.1 m3
45.8 bar g_
carbon steel pressure vessel, cladinternally with austenitic stainless steel
Table 10, Accumulators details
33
Figure 21, Engineered safety features
Water storage volume
Boric acid concentration
Material
1775 m3
2000 £pm
stainless steel
Table 11, Refuelling water storage tank details
34
5.8 Containment systems
5.8.1 Overall system information
The Sizewell B containment design reflects the stringent radiological requirements imposed byBritish authorities. As the ultimate barrier against releases to the environment it has towithstand the peak pressures and temperatures predicted for the most serious accidents indesign basis. Compared to the SNUPPS design the Sizewell containment has been modified toaccommodate the increased volume of equipment and to ensure that safety margins required byBritish standards were met. A notable addition is the secondary containment which envelops theprimary containment and act as a second barrier against releases.
5.8.2 Containment structure.
The containment is cylindrical with a half-sphere dome, radius 22.8 m and inside height of 64 mThe primary containment is made of prestressed concrete with very heavy bar reinforcements.The cylindrical shell is 1.375 m thick, the dome 1 m thick. A 6 mm thick steel liner covers theentire inside containment wall. A secondary containment encloses the primary containment.
5.8.3 Containment penetrations.
The containment has the standard penetrations: Personnel hatch, main steam line penetrationsand ventilation pipes etc.
5.8.4 Containment isolation system
For those process lines that connect to the reactor coolant system or containment atmosphereand goes through the primary containment ther are normally at least two isolation valves inseries. The valves are locked closed ones or automatic isolation ones. In the case of loss of powerthe valves goes to the position with greatest safety. At least one isolation valve is used for thoseprocess lines that do not connect to coolant or containment atmosphere but goes through theprimary containment.
Some parts of the main steam lines do not isolate automatically, among these are the reliefvalves and the steam lines to the auxiliary feedwater pump turbines.
The primary protection system controls all containment isolation valves through two phases ofsignals, where one phase is for closing valves that can be closed without increasing the potentialof damage to the reactor building equipment and still permit safeguards equipment to function,and the other phase is for isolating the component cooling water lines for the coolant pumps andthe heat exchangers for the reactor coolant drains tank and the excess letdown.
5.8.5 Pressure reducing and heat removal systems.
The spray system discharges borated water that reduces the pressure in the containment. Itconsists of two independent trains, see Figure 20, Safety injection, residual heat removal andspray system and Table 12, Spray pumps details. The residual heat system pumps can be used asspray pumps and the spray pumps as residual heat system pumps.
The reactor building heat removal system consists of several parts, see Figure 22, Reactorbuilding heat removal systems. During normal operation is three out of four fans sufficient
35
enough, and after a loss of coolant accident one fan together with one train of the spray systemare sufficient too cool the reactor building.
PresauHserVolve Gallery
PreMuriser • S -
St*am Ccrwretor
H8soa.
II
Steam Generator
- 1 . OiD
RPV 1u
Steam Ceneroter
Steam Generator
Figure 22, Reactor building heat removal systems
36
Number
Design flow
Design head
4 x 100%
795 m3/hr
145 m
Table 12, Spray pumps details
5.8.6 Containment gas control system.
The combustible gas control system that prevents the accumulation of hydrogen to reachunacceptable limits consists of the hydrogen mixing fans, the reactor building coolers, thehydrogen recombiners and the hydrogen monitoring subsystem and is capable of keeping thehydrogen concentration below 3% volume.
5.8.7 Secondary containment.
Sizewell B has a secondary containment of lightweight concrete enclosing the primarycontainment. The emergency exhaust system maintains a negative pressure in the case of arelease. Should any leakages occur from the primary containment these will then be collectedand filtered in the secondary containment before being released through the ventilation stacks.
5.9 Steam and power conversion systems
5.9.1 Overall system information
There are four steam generators delivering steam through two main loops to the two mainturbine generators. The output from the steam generators are cross-connected to balance theflow and pressure when the demand from the two turbines differ.
5.9.2 Turbine-generator
A turbine set consist of a high pressure turbine and three low pressure turbines. Both high andlow pressure types are double flow, for other details see Table 13, Turbine conditions and Figure23, Turbine, steam and condensate system.
The rotor and stator of the generator are cooled respectively by hydrogen and water. Thegenerator produces 629.5 MW power continuously at 23.5 kV.
Speed
Pressure at inlet
Temperature at inlet
Flowrate at inlet
3000 rpm
66.6 bar a
282 °C
955 kg/sec
Table 13, Turbine conditions
37
COCO
iH
89
I|
I3
in
Moistureseparatorrehoaler
« ; * ; * ~ * , l « .
Lu» steammanifold
HPturbme
1 ^ TTKNo1
• • • * « . . > •
No 2 No3 LP turbine
3 x SO-, condensateextraction pumps
Condensatepolishing plant
No2A" ^NoiALP heaters
5.9.3 Main steam supply system
Steam from the steam generators are transferred to the main turbine generators and theauxiliary steam system. If there should be a reactor trip, steam is dumped to the condensers orthe atmosphere. See Figure 25, One leg of the main steam system for more details of the steamflow.
5.9.4 Main condensers and evacuation system
The condenser is of the single pass transverse type, see Figure 24, Main condenser. It isprotected against overpressure by bursting diaphragms. There are three vacuum pumps toextract non-condensable gases and air from each condenser.
Figure 24, Main condenser
39
in
o
wS,
5"(ASto
Main t t M m wftty «<tv« to
Mmoophtra
yr»
OuilMftow
V Stoim dump to otmoop^ofo
Itwtin* bvpxo • b M l
\Slwn
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A
' T< w t m drivon pumft
SMm Bid fo^ Ml within dw «u>di«y buUina
No brut zom
Moirtn Mpwacor,
H h M K l d
condtrmf dump
(turtino bypaa «l>«r«wnl
To man turbint
I - highproMn
cyfcidoc
Mocnmfca' mnn of turtjira hod
5.9.5 Turbine gland sealing system
The turbine gland sealing system is a part of the turbine auxiliary systems and prevents air fromleaking into or out of the end glands of the turbine rotor and the governor valve spindle glands.
5.9.6 Turbine bypass system
The bypass system dumps steam to both condensers and to the atmosphere.
5.9.7 Condenser cooling system
Sea water is taken through a single intake tunnel and through the fine mesh revolving screens,see Figure 26, Main circulating water system and Table 14, Circulating water pumps.
Fieure 26. Main circulatine water svstem
41
Number per turbo-generator
Flowrate per pump
Developed head
Power of the motor
2 x 50%
12 m3/sec
9 m
1.4 MW
Table 14, Circulating water pumps
5.9.8 Condensate and feed water system
The condensate system consists of two trains where each train uses three 50% (of turbine flow)pumps to extract condesate for delivering to the polishing plant and low pressure heaters andfinally to the deaerator. The polishing plant consists of four mixed bed demineraliser units thatremoves corrosion products and condenser leakage impurities.
There are three main feedwater pumps for each turbine with one in stand-by, see Table 15, Mainfeedwater pumps. They take suction from the deaerators and sends the water through the highpressure heaters to the steam generators. If the main feed water pumps are not available theauxiliary feedwater pumps are used, see Table 16, Auxiliary feedwater pumps. For details of thelayout of the feed water system, see Figure 27, Main feed water system.
Number per turbine
Drive
Flowrate_per p_ump
Developed head
Speed
Temperature
3 x 50%
induction motor
0.548 m3/sec
935 m
variable
156°C
Table 15, Main feedwater pumps
Number - motor driven
Design flow
Design head
Number - turbine driven
Design flow
Design head
2 x 100%
135 m3/hr
1162 m
2 x 100%
189 m3/hr
1019 m
Table 16, Auxiliary feedwater pumps
42
Figure 27, Main feed water system
5.9.9 Condensate cleanup system
The pH and oxygen content of the condesate is controlled by injecting hydrazine or ammoniaafter the condesate polishing plant.
43
5.9.10 Steam generator blowdown system
The blowdown system controls the steam chemistry and purity in the steam generators, seeFigure 28, Steam generator blowdown system.
is,23
* 2 » fi tmi li
Figure 28, Steam generator blowdown system
44
5.9.11 Safety and relief valves
There is one quick-closing main isolating valve with bypass, five spring-operated safety valvesand an air-operated globe relief valve in each steam line, see Figure 25, One leg of the mainsteam system for more details.
5.9.12 Other turbine auxiliary systems
There are severally systems that support the main turbine generator in different ways; to supplysteam to the feed heaters and the deaerators, to lubricate the turbine generator end exciterbearings, to prevent air leakage into the glands, see 5.9.5 Turbine gland sealing system, and toprovide oil storage and transfer facilities for the lubricating oil systems.
5.10 Fuel and component handling and storage systems
Both new and used fuel is stored in the fuel building in the fuel storage racks which are locatedunder water in the fuel storage pond, see Figure 29, Fuel route and storage. The fuel storagepond is a stainless steel lined reinforced cavity. There is room for five year of used fuel and one-third of a core of new fuel plus space for a whole core of fuel. The space can be expanded withinthe existing pond. The fuel storage racks are stainless steel storage cells with sheets ofboronbearing polymer where the cells are arranged in a square lattice form. There is also boronin the fuel storage pond water.
All transport of fuel between the reactor building and the fuel building is done through thetransfer tube.
5.10.1 New fuel storage
One-third core of new fuel is stored in the fuel storage racks in the fuel building.
5.10.2 Spent fuel storage
The spent fuel is stored in the fuel storage racks in the fuel building until it is sent forreprocessing.
5.10.3 Handling and inspection systems
During refuelling is the refuelling pool cavity filled with borated water, the internal structureabove the core is lifted out and placed in the refuelling pool cavity, the fuel assemblies are liftedout, rotated and placed in the transfer tube and transferred to the storage pond.
45
6oVI
saz
Figure 29, Fuel route and storage
5.11 Radioactive waste systems
5.11.1 Liquid waste systems
Both reactor grade and non-reactor grade liquid waste are collected, processed and disposed ofthrough the liquid waste system. The main parts are the drain channel A which receives andprocesses clean tritiated reactor grade wastes, the drain channel B which collects and processesreactor coolant leakage plus active effluent from the dirty radwaste drain system, the activelaundry and hot shower subsystem which takes water from washing machines, showers, sinks
46
and such, the chemical drains tank subsystem which collects the waste in two tanks, monitoriesand discharges the waste and the secondary liquid waste system which collects waste from thesecondary systems (located in the turbine building). See the following figures for more details;Figure 30, Liquid waste channel A and B, simplified, Figure 31, Liquid waste system "Activelaundry and hot shower" plus "Chemical drains tank", both simplified and Figure 32, Thesecondary liquid waste system (simplified).
Aa*n
tank
Sow
F - WtsM
Figure 30, Liquid waste channel A and B, simplified
47
Cessationtank
Monitor
tsnkC
Drain chanMlB
i
MonitortankO
Dtaehorga
Aotiva laundry end hat gfeewwf M&gyrtgg
ItsnkA
I
Owmteri
Chtnloal dratw tank wibwgtew.
BoUndWBsta
M
^^Dbh^
F .M •
S • atraincr
Figure 31, Liquid waste system "Active laundry and hot shower" plus "Chemicaldrains tank", both simplified
48
Staam goneradx-
aumppump DraheharaMlB
CDBMtiiintank A
CofeotlonUnk»
oaseparator
1 7
M
• e — * -Dtioharga
EffluentUnk
Efflumttank
(tent
F * flteU • monitor
Figure 32, The secondary liquid waste system (simplified)
49
5.11.2 Gaseous waste systems
The gaseous waste systems consists of the station heating, ventilation and air-conditioningsystem and the gaseous radwaste system which main task is to delay the short-livedradionuclides to be reduced by decay before being released, see Figure 33, The gaseous radwastesystem (simplified).
]i
I!
Figure 33, The gaseous radwaste system (simplified)
50
5.11.3 Solid waste systems
The solid waste systems collect, process and package radioactive waste and store it until it istransported off site, see Figure 34, The solid waste processing (simplified).
Figure 34, The solid waste processing (simplified)
51
5.12 Control and instrumentation systems
5.12.1 Overall system information
The entire plant is built after the 'defence-in-depth' concept, see Figure 35, Defence-in-depth,including the control and instrumentation system.
The control and instrumentation system of Sizewell B is enhanced compared to the standardSNUPPS design. The system consists of the Integrated System for Centralised Operation (ISCO)which is a set of control and data acquisition equipment and computer data processors giving theoperators information about the plant, the Engineering Computer System (ECOS) whichprovides information for the engineers and maintenance staff, and the reactor protection system,see Figure 36, Instrumentation and control system.
ISCO is a distributed computer system based on a updated version of Westinghouse Electric'sDistributed Processing Family (WDPF) with two redundant Westnet highways, one used forsafety category 1 data and the other for safety category 2. In addition is an Ehternet informationhighway used for non-time-critical messages. The user interface of ISCO is called the DistributedComputer System (DCS) with Sun SPARC RISC workstations running Sun 'Solaris' UNIXoperating system performing the functions of DCS. The lover level of the WDPF is called theProcess Control System (PCS) and comprises the distributed I/O processors. It is conventionalprogrammable logic controllers programmed via ladderlogic diagrams. Some of the new featuresof WDPF for Sizewell is the ability to execute operator control from special dedicated controlpanel switches instead of using a keyboard and the use of modern man machine interface via theSPARC displays.
The High Integrity Control System (HICS) is a safety grade microprocessor I&C system from theWestinghouse Eagle family and provides the majority of the safety classified man-machineinterface. In addition to using the DCS displays it also uses four plasma screens in the maincontrol room which are non-breakable in the case of an earthquake destroying the normalcathode ray tube displays. HICS is split in four electrically and physically segregated redundantnetworks.
INCHQUALITY
REACTOR&SUI'PORTINC,
SYSTEMS
STATIONCONTROLSYSTEMS(WISCO)
OU KHOTKCITOX |SYSTHMS
v
SAFETYFEATURES
1KH HI I. ICONTAINMENT
m ii.DIM:
OPERATOR
Figure 35, Defence-in-depth
52
Adminoffices
Auriliary Auxiliaryshutdown monitoring
— * " J 1 facility
Main sutta Alarms Plant overvivw panel
Proms Control H i 9 h Integrity Controls (HIO)System (PCS) ***
Control room
Level 2
Data processingand displaygeneration
Level 1Protectionand control
Process plant
IMP-Isolated Measuramant Pod VIU - v»lv» InMrfau Unh SWG - Switch G«r SIDS - Saftty Information Display System DaUlinks Hardwire
Figure 36, Instrumentation and control system
5.12.2 Protection system
The protection system is made up of two mutually diverse technologies using two-out-of-fourvoting. The first system is the Primary Protection System (PPS) which is a microprocessorsystem based on Westinghouse Eagle system and the other system is the Secondary ProtectionSystem (SPS) which is based on analogue trip units and 'Laddie' magnetic logic (developed in the1960s for gas cooled reactors). There has been invested more than 500 man years into the design,manufacture, testing and assessment of the primary protection system.
The PPS system takes input from plant sensors, nuclear instrumentation, rod position sensorsand manual control from the control room. These input goes to a computer system whichgenerates output to the reactor trip system, safety actuation equipment and other systems likethe data processing system., see Figure 37, Primary Protection System (PPS) and the protectioncomputer system. The PPS reactor trip parameters are source range neutron flux, intermediaterange flux, power range flux, nitrogen-16 power measurement, core limit calculation for lowDNBR (Departure from Nucleate Boiling Ratio), linear power density (kW/m), rod cluster controlassembly misalignment calculations, rod cluster control assembly bank movement surveillancecalculations, rod cluster control assembly bank insertion calculations, pressuriser pressure,pressuriser level, reactor coolant system flow rate, reactor coolant cold leg narrow rangetemperature, steam generator narrow range water level, loss of 11 kV supply detected, bothmain turbines detected tripped, safety injection signal from the primary protection engineeredsafety features actuation system.
The SPS trip parameters are source range neutron flux, power range neutron flux, reactorcoolant system narrow range pressure, reactor coolant pump motor speed, reactor coolant pump
53
motor current, steam generator narrow range water level, main steam isolation valveinsufficiently open.
ProtectionComputer. System
Features^
Other
InputSignals
\
To/From ulhcr channels
OlhcrSystems
ReactorTrip
SwitchRear
Figure 37, Primary Protection System (PPS) and the protection computer system
5.12.3 Regulating system
The station automatic control system is a part of the high integrity control system. It is notnecessary with operator intervention for at least 30 minutes after a fault, due to the design ofthe plant, control and protection system. The HICS is safety category 1 and mainly controlssafety category 1 equipment. The process control system is safety category 2 and used forcategory 2 and 3 equipment.
54
5.12.4 Instrumentation system
Apart from the normal instrumentation there are also some special instrumentation systemsused for calibration of the reactor protection system, providing post-fault monitoring and to givethe operators information about the plant. Among these system we find the flux mappingsystem, the boron concentration monitoring system, the reactor vessel level instrumentationsystem, the core exit temperature measurement system, the loose parts and core vibrationmonitoring system, the fire detection system, the seismic instrumentation system, theradiological protection system, the health physics instrumentation and the main steam linenitrogen-16 gamma detectors. Post-fault neutron flux measurements are done by the secondaryprotection system source range detectors.
5.13 Electrical power systems
Sizewell B is connected to the external grid at three separate 400kV points, two at Bramford, oneat Norwich and one at Pelham.
5.13.1 Main transformer and connected equipment
The main generators produces electricity at 23.5 kV which is transformed up to 400 kV for theoff-site distribution net, a minor share is transformed down to 11 kV for station use, see Figure38, Main and essential electrical systems. The gross electrical output is 1258 MW while the netelectrical output is 1188 MW.
5.13.2 Plant distribution system
Power taken from both the external grid and the main generators are distributed around theplant at 11 kV, 3.3 kV and 415 V. The electrical system is split in two, the main electrical systemand the essential electrical system. The essential electrical system provides power to safetycategory 1 and 2 equipment and to some category 3 equipment.
There are two composite 11 kV boards which are split into two segregated boards; a station and aunit board. The unit auxiliary boards 1 and 2 serve the auxiliaries of the main turbinegenerators 1 and 2 at 3.3 kV and indirectly at 415 V. The station auxiliary boards 1 and 2supplies mainly the 3.3 kV to 415 V transformers for the reactor auxiliaries which do not need asupply from the essential system.
The essential electrical system is powered through the four essential auxiliary transformerswhich transforms the 11 kV from the four 11 kV boards down to 3.3 kV. The system is dividedinto four independent trains where the essential a.c. system is segregated by three-hour firebarriers. There is always a minimum of two-way segregation.
5.13.3 Standby power supply
Each train of the essential 3.3 kV system is connected to an essential diesel generator. In thecase of loosing both off-site and essential diesel generator the reactor can be kept at hotshutdown for 24 hours through the supplies from batteries and the battery charging diesels.
The essential d.c. power system which provides power to category 1 equipment will provide 120minutes of power after loss of a.c. power, off-site power and essential diesels if the chargingdiesels are available at request.
55
The essential uninterruptible power supply system provides 110 V a.c. single-phase to thereactor protection system. The system is category 1 and the batteries are rated to last for 120minutes.
1J2kV
• To A Station •
132kV U2kVBronx/ord 1 Norwich Bromford 2
40OI<V L L
Pelhom
EHV GridConnections
( 4O0kV (
L \{ Bus Coupler
GeneratorTransformer 1
4Generator 1
UnitTransformer
23.5W
UkV U81
StationVontformer 8
SB1 1 1StationTransformer 5
SB2
j Generator* Transformer 2
23.5W Generotor 2
f sUnit' Transformer 2
w uUnit
Auxiliory, Board 1
3.3M } A _ B
i I1 8
Turbine House ondAuxiliary Building
415V
Other Areas
Ewentiol ElectricalSystem
StationAuxiliaryBoards
TT£ i
} - UB2 n kvu Will8 8 1
UnitAuxiliaryBoard 2
B ^ A f J.JkV
Turbine House ondAuiiliary Building
TOiesef
Generatorsr Generot
I f ? ),_, f ?N E84^ f'EBS f* f" { ( EB3r/
I N _ _ Essential Pumps ' I Ejsentiol Pumps '
1 415V K15V I
r i r T I tN.^ Essentiol Pumps, Heoters, Fon Coolers etc —————'
I 1Battery Chargers ond Low Voltoge Systems
u i 2 £ n i EB Essentiol BoordSB Stotion BoardUS Unit Board
CW Circulating WaterRC Reactor Coolant
Figure 38, Main and essential electrical systems
56
6. FIRE PROTECTION
The fire protection system comprises detection and suppression systems that cover the entiresite. Several types of fire detectors are installed, chosen according to the type and nature of firerisk:
• Smoke detectors
• Flame detectors
• Heat detectors
Fire alarm can be raised from manual stations, nowhere at the site is the distance to the nearestfire alarm more than 30 meters.
Systems and equipment for fire fighting include:
• Fixed water suppression system. (Fire-fighting ring main. Sprinkler and spray systems.)
• Fire hydrant systems. (External hydrant system.)
• Gaseous suppression. (Fixed halon suppression.)
• Foam suppression
• Portable extinguishers.
6.1 Buildings, layout and materials
The fire protection system has a secondary role in ensuring nuclear safety at the plant.Primarily one aims at a construction that minimises the risk for fires and - in the event of a fire -passively restricts the damage to equipment and structures. Segregation and sectioning ofessential services equipment and fire resistant barriers are the principal means for fire riskminimisation and protection against fire induced events.
6.2 Fire-fighting equipment
The water reservoirs and the fire-fighting pumphouse supply the fixed fire-fighting ring main.The pumphouse contains three 100% diesel driven fire pumps and two 100% diesel drivenhydrant pumps. The ring main feeds all fixed water suppression subsystems and the hose reelsystem.
Automatic and manual sprinkler systems are widely used in the fixed water suppression system.These are mainly intended for suppressing fires in solid materials. Spray systems (delugesystems) are primarily intended for use against flammable liquid fires.
In the radwaste building Halon 1301 total flooding systems are installed. Hand-held portable fireextinguishers for manual fire-fighting are distributed throughout the site.
57
7. PLANT PERFORMANCE DURING NORMAL OPERATION
At the time of making this report (end 1996) Sizewell B has only been running for one cycle, butthe first cycle performance is an early indicator for its 40 year design life. Throughout the firstcycle plant performance and availability has been exceptional good
7.1 Phases of normal operation
Sizewell B is the first nuclear power station in UK to perform automatic frequency responseoperation to follow changes in demand.
7.2 Plant statistics
The first cycle statistics are excellent, even compared with similar plants. Sizewell B achieved200 days of continuously running at 100% on 4 April 1996.
Commercial operation
Capacity factor (MDC)*
Availability
95.07
91.5 %
97.0 %
* MDC - Expressed as a % in MW. Actualoutput compared to potential output.
Table 17, Plant statistics
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8. ACCIDENT ANALYSES
The accident analyses are based on event tree analysis for identified initiating events, combinedwith fault tree analysis of systems that may be called on to arrest a potentially dangeroussequence or mitigate its consequences. The starting point is the fault schedule, that is the list ofinitiating faults whose consequences are to be assessed. Sequences where safety systems andsafeguards function as expected do not lead to exceedance of plant safety limits, these are designbasis faults. There are about 180 initiating faults in the fault schedule, leading to app. 5000design basis faults. These design basis faults are supplemented by transient analyses toascertain that structural integrity of components is not challenged and that process parametersare within acceptable bounds.
Sequences leading to uncontrolled releases of radioactivity are beyond design basis faults. Thesefaults can be grouped in two classes: Benign beyond design basis and large uncontrolled releases.The benign beyond design basis faults do not lead to releases larger than one ERL, uncontrolledreleases may require initiation of emergency measures.
Plant damage sequences that are predicted to have roughly the same characteristics as regardsrelease and subsequent radiological effects are grouped together in four main types of plantdamage states:
• Core damage (Loss of coolable geometry, core melt.)
• Reactor building bypass (Direct leak path, such as incomplete containment isolation).
• Damaged reactor building (Sequences without core melt, but inadequate containment.)
• Ex-reactor plant damage states (Release from fuel pond, radwaste plant.)
Possible radiological effects of all sequences that may lead to uncontrollable releases are summedto give the total individual risk for death. The predicted limiting individual risk of death is 5.2 x10'8 per year by early effects, and 6.9 x 108 per year due to fatal cancer. In total a risk of deathassessed to 1.2 x 10~7 per year. This is nearly one order of magnitude less than the target of 10"6
per year.
8.1 Malfunctions in different systems
To reduce the many design basis faults to a manageable number for radiological analysis theywere grouped into nine categories comprising:
• Leaks into the auxiliary building at power
• Leaks into the auxiliary building at shutdown
• Reactor building loss of coolant accident at power
• Reactor building loss of coolant accident at power
• Intact circuit faults at power
• Intact circuit faults at shutdown
59
• Refuelling route faults
• Radwaste building faults
• Faults involving other sources of radioactivity
For each category those with the more severe radiological consequences were use to characteriseall sequences in that category, leading to a conservative assessment of the frequency of releases.The final result is within acceptable bounds, as also shown by the above death risk assessment.
8.1.1 Internal and external hazards
Various techniques were used to address hazards:
• Event tree analysis
• Fault tree analysis
• Engineering judgement
Hazards assess in detail were:
• Extreme winds
• Extreme temperatures
• Aircraft impact
• Seismic hazard
• Fires outside safety classified buildings
• Dropped loads
• Pressure part failure
• Turbine disintegration
In general the most significant contributions to the frequencies of plant damage were frominternal fires and earthquakes.
60
9. RADIATION PROTECTION
9.1 Basic radiation protection criteria and derived working rules
The radiological protection criteria applied in the design and operation of Sizewell B NPP reflectcompliance with governmental rules and regulations as given in Ionizing Radiation Regulationsand the recommendations of the National Radiological Protection Board. It is a stated goal thatthe actual exposure shall be kept well below statutory limits, and that the ALARP (As Low AsReasonably Practicable) principle is adhered to. The targets set shall apply to any individualmember of Sizewell B staff, and to temporary or contract workers.
The strategy used to achieve design targets can be summarised as follows:
• Obtain overview and calculate strength of radiological sources
• Limit radiological hazards at source
• Systematic implementation of dose reducing measures in design
• Final assessment of operator and public exposures
9.1.1 Dose limits
See Table 18, Dose limits for the statutory limits for doses.
Organ Annual dose equivalent
Effective dose equivalent _10 jnSv
Hands,_forearaaSj_feet,_ankles_ and_skin_
Eyes
_500mSv
150 mSv
Table 18, Dose limits
The design target for collective exposure is that it should not exceed 2 man Sv per year per GWeinstalled capacity. For Sizewell this transform into a design target of 2.4 man Sv/year.
For the target dose rates for direct radiation exposure of station staff, see Table 19, Directradiation exposure of station staff.
For short to very short exposures dose rates in excess of 10 mSv/h may be allowed given that theyearly limit is not exceeded.
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Access requirement
Continuous (more than 10 man-hour/week)
1 to 10 man-hour/week
Less than 1 man-hour/week
1 to 10 man-hour/year
Less than 1 man-hour/year
Dose equivalent
Mean
0.001
0.01
0.1
1.0
10
rate, mSv/h
Maximum
0.005
0.05
0.5
10
Table 19, Direct radiation exposure of station staff
9.2 Shielding
The design of Sizewell B has taken account of experiences gained from SNUPPS reference plantsand operation of other PWRs, particularly in France. Components or systems found to give largecontributions to annual staff doses have been redesigned or changed to provide better shieldingfor workers. The measures include:
• Improved shielding in the reactor pressure vessel refuelling cavity
• Additional shielding during maintenance of steam generators and reactor coolant pumps
• Substitution of Inconel-690 for Inconel-600 steam generator tubing
• Design of pipework to reduce build-up of active materials
Remote in-service inspection of equipment has been introduced where possible, notably forreactor pressure vessel welds. All primary side inspection of the of the steam generators is donewith automated equipment. Since refuelling activities contribute strongly to collective dose therehas been a determined effort to reduce doses during refuelling.
9.3 Radiation
The principal sources of radiation are the core, primary coolant and the depositions in piping andcomponents in the primary circuit. Induced sources, especially from cobalt and nickel isotopes,account for most of the occupational radiation exposure. A preliminary assessment showed thatapp. 85 % of occupational exposure is attributable to cobalt-60 and cobalt-58, the maincomponent sources being steam generator tubing, steel alloys in primary circuit and Inconelgrids in fuel assemblies.
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9.3.1 Radiation levels
An assessment of collective doses based on analyses of tasks - dose rates, duration, radiationlevels etc. was carried out to ascertain that contributions from different work functions wereincluded:
• In-service inspection
• Refuelling
• Scheduled maintenance
• Unscheduled maintenance
• Waste processing
• Health physics
• Operation and surveillance
Experience from PWR plants in Europe and USA shows that unscheduled maintenance accountsfor roughly the same annual collective dose as does scheduled maintenance. The task-basedanalysis gave the following contributions to annual collective dose, see Table 20, Annualcollective dose.
In-service inspection
Refuellingi
i
Scheduled ma in t enance •r~i
t~i
Unscheduled maintenance
Waste processing
Operations and surveillance
Mechanical
Electrical
In-service testing
(estimated)
230
260
377
101
49
527
191
237
Man mSv
Man mSv
Man mSv
Man mSv
ManmSv
Man mSv
Man mSv
Man mSv
11.7 9
13.2 9
19.19
5.19
2.5 9
26.7 9
9.7 9
12.0 9
& oftotal
& of total
h of total
h of total
h of total
h of total
'o of total
i> of total
Table 20, Annual collective dose
Health physics is included as a 10 % allowance in the separate work functions.The target is thatannual dose to an individual operator shall not exceed 10 mSv per year, which is achieved byrequiring that no single task shall result in a dose of more than 10 mSv. The highest individualtask dose for the most demanding tasks is given in Table 21, Highest individual task dose.
63
In-service inspection
Refuelling
Scheduled maintenance
7.6 mSv
8.5 mSv
6.6 mSv
Table 21, Highest individual task dose
As seen from the table there is no single task contributing more than 10 mSv.
9.3.2 Radiation monitoring
The process radiation monitoring systems and sampling systems provide the data required forproof of compliance with operational limits. Process monitors are located in active systems, non-active auxiliary systems and in heating and ventilation systems. Health physics monitors doserate levels, airborne activity and personnel contamination levels. The radwaste sampling systemcollects samples from the radwaste process systems and the airborne discharge ducts.
9.4 Monitoring systems and action levels for radioactive releases to theenvironment
9.4.1 Monitoring of releases to the atmosphere
The releases to the atmosphere are monitored by the measurement system for ventilation stackand ducts together with the sampling system for airborne activity.
9.4.2 Monitoring of releases to water
The liquid radwaste system controls the activity of discharges to the sea. Normally the effluentsare collected in hold-up tanks for sampling and filtration before being discharged to the sea, butthe system has facilities for further treatment of liquid waste if necessary.
64
10. OFFSITE DOSE ASSESSMENT
The off-site dose derives from airborne and liquid discharges, the main exposure pathways being:
• External and internal exposure from airborne activity
• External exposure due to ground contamination
• Internal exposure due to foodstuffs
• Handling of contaminated fishing gear
• Direct radiation from site structures
10.1 Releases under operating conditions
10.1.1 Radiation doses
The estimated collective doses to the UK population due to airborne and liquid discharges fromnormal operation of Sizewell B are (per year):
Liquid:
Airborne:
Total:
0.024 man Sv
0.10 man Sv
0.12 man Sv
The annual dose to the critical member of the public due to liquid discharges is 14.8 uSv.
The annual dose to the critical member of the public due to airborne discharges is 75 uSv.
The annual dose to the critical member of the public at Sizewell village due to both liquid andgaseous discharges is 85 uSv.
The predicted doses are below the ICRP limits and the target limit set by NE (170 uSv).
10.2 Releases under hazardous conditions
Abnormal conditions may arise due to faults in equipment and systems, operator error orexternal hazards such as storms, earthquakes or floods. A fault causing a release that mayendanger public health must be shown to be of sufficiently low probability that it does not give asignificant contribution to societal risk in general.
Design targets for Sizewell B are defined in terms of an Emergency Reference Level (ERL). Adose below ERL to a member of the public is unlikely to justify countermeasures. Releases largerthan 1 ERL are referred to as uncontrolled releases.
10.2.1 Contamination of the environment
To characterise off-site consequences there are defined 22 different release categories, primarilyin terms of effective doses to members of the public. These categories cover the whole range ofdesign basis and uncontrolled releases. For assessment purposes an ERL is set equal to 0.1 Sv
65
(100 mSv), and large releases are characterised in terms of doses at a distance of 3 km from the.plant.
10.2.2 Radiation doses
With ERL as unit reference there has been defined fault frequency targets related to fractions ofERL, see Table 22, Fault frequency targets.
Release in ERL
0.1 to 1ERL
0.01 to 0.1 ERL
0.001 to 0.01 ERL
Target frequency per year
10-4
10"3
10-2
Assessed frequency per year
1.6 x 10-4
4.9 x 10"3
1.6 x 10-2
Table 22, Fault frequency targets
For releases larger than 1 ERL there are two target frequencies:
• The summed frequency of uncontrolled releases shall be less than 10'^ per year.
• The contribution of any single fault shall be less than 10"' per year.
Uncontrolled releases resulting from large accidents are often related to core damage frequency(which for Sizewell B is assessed to be 9.54 x 106 per year), but core damage is not synonymouswith uncontrolled release. Mitigation systems and release barriers intervention have to be takeninto account to assess the probability for uncontrolled releases. Major releases do in generalimply that some core damage must have occurred, though spent fuel in temporary storage mayin theory cause a substantial source term for external release.
66
11. PLANNING, ORGANISATION AND ADMINISTRATIVE CONTROL
The responsibility for station management rests solely with the station manager, who is incharge of all activities on the station during plant operation. The station manager reports to theexecutive director of operations.
The management structure reflects the extensive operating experience accumulated at NuclearElectric, in addition one has drawn on experience with PWR operation world-wide andrecommendations of the IAEA. A structure with three main departments:
• Production department
• Management support department
• Technical support department
Each department is headed by a department manager.
11.1 Production department
The department has responsibility for
• plant operation and surveillance testing
• routine maintenance and minor repairs
• radiological and chemical monitoring
• work scheduling and short-term planning
• communication with company production co-ordinators
The department decides how the plant is to be operated day-to-day and that the technical aspectsof operation are in compliance with pertinent rules and regulations.
11.1.1 Management support department
The department has responsibility for
• business planning
• finance and procurement
• personnel and administration
• training co-ordination
• quality assurance
• emergency planning and industrial safety
• management information systems and communications
67
The department is responsible for maintaining an effective and clearly defined organisationalframework.
11.1.2 Technical support department
The department has responsibility for
• engineering, nuclear and other
• outage management
• safety case maintenance and maintenance strategy
• operational experience feedback
• chemistry and health physics service
The department is responsible for formulating engineering and technical policies, to plan andmanage outages and modifications, and to ensure compliance with safety requirements.
11.1.3 Training programme
There are station staff training programmes which are mainly directed towards qualifyingpersonnel for station operation. These programmes take advantage of the dedicated plantsimulator and other facilities for training.
Corporate training programmes include training in technical aspects to ensure that knowledge ofand insight in PWR technology is imparted to staff less directly involved in operation.
11.1.4 Operating procedures, instructions and orders
Operational documentation consists of
• Station manuals (Technical descriptions, operation manuals, technical specifications).
• Station emergency plan and handbook
• Nuclear Electric safety rules (Cover radiological, electrical and mechanical hazards).
• Quality assurance programme
• Station records (Comprehensive record of station operation).
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12. ORGANISATION OF THE AUTHORITIES
12.1 Local organisation
The Health and Safety Executive is the governmental agency in charge of nuclear safety. Theseresponsibilities are taken care of by the Nuclear Installations Inspectorate (Nil), which performsthe actual technical control and oversees the nuclear industry on behalf of the Government.Licences and permits are issued by Nil acting as representative for the authorities.
69
13. PROBABILISTIC SAFETY ASSESSMENT
A probabilistic safety assessment was performed for inclusion in the Pre-Construction SafetyReport which was the basis for granting a license for Sizewell B. For the Pre-Operational SafetyReport a more comprehensive and detailed PSA was carried out, covering a wide range ofoperational states and initiating events. The underlying requirement was to provide thoroughsafety analysis of the design and to demonstrate that the risk to the public is below an acceptablelevel, notably that the probability of large or fatal doses is below 10~6 per reactor year.
The analysis considered internal, external and beyond design basis initiating faults togetherwith non-core sources of radioactivity. Level 1 PSA is concerned with the state of process andplant equipment resulting from a fault chain, Level 2 with containment performance and Level 3with the release of radioactivity to the environment and radiological consequences. All 3 levelsare covered in the Sizewell B probabilistic safety assessment.
For the Level 1 analysis an initial list of about 170 initiators was drawn up, the possibility foreach to occur in 6 different operational states were to be considered. To reduce the analysis to amanageable size there was extensive binning of fault sequences. Still there were 58 systemsmodelled and 61 fault trees analysed. Each fault tree typically consisted of 1500 components and900 gates. There were several thousands cut-sets for each fault tree, a cut-off frequency of 1010
per year was used.
The Sizewell PSA has a wider scope than most PSAs and does not only look at the core meltfrequency. If one considers internal initiators at power the core melt frequency is likely to beapp. 2-3 106 per reactor year.
For external initiators a list of 50 was reduced to 9 that were quantified: Seismic events, internalfires, fires external to the main building, temperature extremes, extreme wind, pressure partfailure, turbine disintegration, aircraft crash and dropped load. The same plant damage stateswere used as for internal initiators. Of particular interest are the fault chains resulting fromfires and seismic events. The analysis shows that seismic events are not major contributors tothe overall results.
The main contributors to the risk of death to an individual are:
Small releases 38%
Beyond Design Basis Initiating Faults 28.5%
These results show that the risk associated with credible faults have been reduced to the pointwhere the less credible faults (that is Beyond Design Basis faults) contribute nearly as much asDesign Basis faults.
The target frequency for the analysis was that a release resulting in 100 mSv whole body dose atthe plant perimeter should have a probability less than 106 per reactor year. The analysis showsthe probability to be somewhat higher, but the risk is still acceptably low. In round figures theSizewell B plant poses a risk to the public which is two to three orders of magnitude less thanthe total everyday risk caused by more familiar sources of risk.
70
14. REFERENCES
Sizewell B Safety Statement. SXB-IP-772002/1. Nuclear Electric pic. January 1994.
Proceedings of a Forum on Safety Related Systems in Nuclear Applications. The Royal Academy ofEngineering. December 1992. ISBN 1 871634 24 5
ATOM 433 March/April 1994
Sizewell B technical outline, Nuclear Electric pic.
New Civil Engineer Sizewell supplement, New Civil Engineer October 1994
Sizewell B Power Station, Supplement to Civil Engineering vol. 108, special issue 1, February 1995
71
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