Design, Safety Technology and Operability Features of Advanced VVERs

Transcription

1 Technical Cooperation Project INT/4/142 Interregional Workshop on Advanced Nuclear Reactor Technology for Near Term Deployment IAEA Headquarters, Vienna, Austria, 4-8 July 2011 Design, Safety Technology and Operability Features of Advanced VVERs Presented by Nikolay Fil OKB GIDROPRESS Russian Federation

2 2 Introduction OKB GIDROPRESS has been established in 1946 to develop designs of nuclear steam supply system (NSSS) and separate equipment for nuclear installations. Some milestones: 1947 Research heavy water reactor 1954 Steam generator for 1st NPP 1963 Lead-bismuth NSSS for submarine First VVER Boiling water reactor VK Fast sodium reactor BOR First VVER First VVER-1000 Now VVER-440 and VVER-1000 are in operation in Russia, Ukraine, Armenia, Finland, Bulgaria, Hungary, Czech Rep., Slovakia, China. Now VVERs of MWe, VVER-SCP, sodium BN-800/1200 steam generator, lead-bismuth SVBR-100 are under development. Visit for details.

3 Introduction What is VVER ВВЭР: водо-водяной энергетический реактор WWER: water-cooled water-moderated energy reactor Pressurized light water reactor. Loop-type reactor plant. Horizontal steam generators. Hexagonal fuel assemblies. High level of inherent safety. About 1500 reactor-years of operating time 3

4 VVER evolution Low and medium power designs ( ) RP design RP type Site (units) Status V-1 VVER-210 NVNPP (1) DCOM V-2 VVER-70 NPP «Rheinesberg» (1) DCOM V-3М VVER-365 NVNPP (2) DCOM V-179 VVER-440 NVNPP (3,4) SLE V-230 VVER-440 Kola NPP (1,2), NPP «Kozloduy» (1-4), NPP «Nord» (1-4), NPP «Bohunice» (1,2) SLE DCOM V-270 VVER-440 Armenian NPP UO V-213 VVER-440 Kola NPP (3,4), Rovno NPP (1,2), UO NPP «Loviisa» (1,2), NPP «Paks» (1-4), NPP «Bohunice» (3,4), NPP «Mochovce» (1,2), NPP «Dukovany» (1-4), NPP «Nord» (5) DCOM 4

5 VVER evolution High power designs ( ) RP design RP type Site (units) Status V-187 VVER-1000 NVNPP (5) SLE V-302 VVER-1000 South Ukraine NPP (1) UO V-338 VVER-1000 South Ukraine NPP (2), Kalinin NPP UO (1,2) V-320 VVER-1000 Balakovo NPP (1-4), Zaporozhe NPP UO (1-6), Rovno NPP (3,4), Khmelnitsk NPP (1,2), South Ukraine NPP (3), Rostov NPP (1,2), NPP «Temelin» (1,2), Kalinin NPP (3), NPP «Kozloduy» (5,6) V-392 VVER D V-446 VVER-1000 NPP «Bushehr» (1) COM 5

6 VVER evolution Current designs ( ) RP RP type Site (units) Status design V-392B VVER-1000 Balakovo NPP (5) D V-412 VVER-1000 NPP «Kudankulam» (1,2) UC V-428 VVER-1000 Tianwan NPP (1,2) UO V-407 VVER D V-466B VVER-1000 NPP «Belene» (1,2) D V-392М VVER-1200 NVNPP-2 (1,2) UC V-491 VVER-1200 LNPP-2 (1-4), BaltNPP (1,2) UC D V-448 VVER D VVER-1200 is also a starting point for future VVER designs of MWe power. 6

7 VVER evolution VVER-1200 Basic design VVER-1200A 300 MWe per loop 600 MWe per loop VVER-600 VVER

8 VVER evolution Future designs RP design RP type Site (units) Status V-498 VVER CtP V-501 VVER-1200А - CtP VVER CtP V-393 VVER-SCP - CtP 8 Abbreviations used: SLE service life extended; COM commissioning; CtP concept proposal; D design is developed; DCOM decommissioned; UC NPP under construction; UO under operation.

9 NPP-2006 basis Now VVER-1200 (also known as NPP-2006) is the main design for near term nuclear power program in Russia and for bidding at international market. NPP-2006 is the evolution of VVER-1000s by improving plant performance and increasing plant safety. Plant performance Rated power 1170 (1000) MWe Primary pressure 16.2 (15.7) MPa Secondary pressure 7.0 (6.3) MPa Coolant temperature 329 (320) C Refueling period 24 (12) months Burn-up in FA 70 (50) MWD/kgU Plant safety Passive reactor scram system Passive core flooding system Passive decay heat removal system Passive containment cooling system Passive hydrogen removal system Passive corium catcher 9

10 NPP-2006 basis Reference plants for NPP-2006 design: Operating standard VVER-1000 with V-320 AES-91 with V-428 Tianwan NPP AES-92 with V-392 Kudankulam NPP (V-412) LNPP-2 (VVER-1200/V-491) NVNPP-2 (VVER-1200/V-392M) EUR organization has certified in 2007 that AES-92 design has passed all the steps of the analysis of compliance vs. EUR. Following this analysis, a specific subset of the EUR volume 3 dedicated to AES-92 project has been published by EUR organization. 10

11 NPP-2006 performance Parameter Value NSSS equipment lifetime, years 60 Load factor, % 92 NSSS availability factor 99 Efficiency factor, net % 35.7 Length of fuel cycle, years 4-5 Frequency of refueling, months 12 (18-24) FA maximum burn-up, MW day/kgu 70 Inter-repair period length, years 4-8 Refueling length, days 16 Number of unplanned reactor shutdowns per year 1 11 For more details

12 Reactor NPP-2006 equipment 7 1 In-core instrumentation detectors 2 Top head unit 3 Protective tube unit 4 Core barrel 5 Core baffle 6 Core 7 Reactor pressure vessel 12 No welds in core region No joints below cold leg

13 NPP-2006 equipment Core Fuel assembly Top nozzle Spacer grid Fuel rods fuel assemblies of to 5% enrichment. To 121 control rods. Bottom nozzle 312 fuel rods. 18 guide channels. 13 spacer grids.

14 NPP-2006 equipment Steam generator Cylindrical vessel of 4200 mm internal diameter, mm long with welded elliptical bottoms HX U-tubes of 16x1.5 mm. 14

15 NPP-2006 equipment Pressurizer Parameter Value Total volume, m 3 79 Water volume, m 3 55 Pressure, MPa 16,1 Temperature, C

16 NPP-2006 equipment Main coolant pump Head 0.59 Mpa Flow rate m 3 /h Speed 1000/750 rot/min Current 50 Hz Massive flywheel ensures smooth reduction of flow rate in case of loss of power. No coolant leakage in case of loss of power min 24 (72) hours. 16

17 NPP-2006 safety Parameter Frequency of severe core damage, 1/year Frequency of the limiting emergency release, 1/year Autonomy time of passive safety systems, hour Design basis earthquake, MSK-64 scale Safe shutdown earthquake, MSK-64 scale Value < 10-6 < (72) 6 7(8) 17

18 NPP-2006 safety Novovoronezh NPP-2 Leningrad NPP-2 ECCS active part Emergency boron injection system High and low pressure combined twochannel system with ejector pumps with internal redundancy of main safety functions Two-channel system with channel redundancy 2 х 100 % and internal channel redundancy 2 х 50 % High and low pressure separate four-channel systems with channel redundancy 4 х 100 %, each Four-channel system with channel redundancy 4 х 50 % Emergency feedwater system SG emergency cooldown system Core passive flooding system (HA- 2) Passive heat removal system (PHRS) Closed two-channel system with redundancy 2 х 100 % Passive four-channel system with channel redundancy 4 х 33 % with two accumulators in each channel Passive four-channel system with channel redundancy 4 х 25 % with two heat exchangers, cooled by air, in each channel Not available Four-channel system with channel redundancy 4 х 100 % with emergency feedwater storage tanks Not available Not available Passive four-channel system with channel redundancy 4 х 33 % with 18 heat exchangers, cooled by water, in each channel 18 Passive systems are provided to back-up the active systems for all main safety functions (reactivity control, fuel cooling and confinement of radioactivity) during min 24 (72) hours.

19 NPP-2006 safety Emergency boron injection system ГЕ ГЕ ГЕ ГЕ - injection into PRZ in case of primary-to-secondary leak for fast decrease of primary pressure - injection into primary circuit to shutdown reactor in case of failure of reactor scram (ATWS) - pumps of 24.5 Mpa shut-off head are connected to tanks with 40 g/kg boric acid concentration 150 m 3 each

20 NPP-2006 safety Active ECCS in NVNPP-2 High and low pressure combined system with ejector pumps is intended for boric acid solution supply into reactor coolant system within the whole spectrum of design basis LOCAs up to DEGB of 850 mm main coolant pipeline. 4 3 ГЕ ГЕ ГЕ ГЕ Parameters Value P HPIS max. head MPa 8 P LPIS max. head MPa 2,5 G LPIS max. supply, m 3 /h 900 C Н 3 ВО 3 concentration,g/kg 16

21 NPP-2006 safety Active ECCS in LNPP-2 Separate 4-trains high and low pressure injection systems are intended for boric acid solution supply into reactor coolant system within the whole spectrum of design basis LOCAs up to DEGB of 850 mm main coolant pipeline. channel channel 3 ГЕ ГЕ ГЕ ГЕ channel channel 2 21 Parameter Value P HPIS max. head, MPa 7,9 G HPIS max. supply, m 3 /h 260 P LPIS max. head, MPa 2,5 G LPIS max. supply, m 3 /h 900 C Н 3 ВО 3 concentration, g/kg 16

22 NPP-2006 safety Passive part of ECCS Passive part of emergency core cooling system is intended for boric acid solution supply into the reactor at the primary pressure less than 5,9 MPa under LOCA till start of active ECCS. 4 3 HA ГЕ HA ГЕ HA ГЕ HA ГЕ Parameters Value P Nominal pressure, MPa 5,9 C Н 3 ВО 3 concentration, g/kg 16 V Accumulator/solution volume, m 3 60 / 50

23 NPP-2006 safety Additional passive ECCS in NVNPP-2 design prevents progression of a LOCA to severe core damage in case of failure of active ECCS during min 24 hours. Flow rate law verified by scaled Experiments. ГЕ ГЕ ГЕ ГЕ Time, s Flow, kg/s To 5430 To To To ,0 5,0 3,3 1, Water inventory 120x8 m 3 Design pressure 3.0 MPa Pressure of connection 1.5 MPa

24 NPP-2006 safety Emergency cool down system via steam generators in NVNPP-2 design is intended for removal of decay heat and RP cool down in case of loss of normal and auxiliary feed water supply. 4 3 канал Parameter Value Channel power, MW 80 Temperature of condensate, С Max. flow rate to one SG, t/h 200 Control of cool down rate Available

25 NPP-2006 safety Emergency cool down system via steam generators in LNPP-2 design is intended for removal of decay heat and RP cool down in case of loss of normal and auxiliary feed water supply. 4 3 канал 3 канал 4 канал 1 канал Parameter Value G Flow rate from one EFWP, t/h 150 T Emergency feed water temperature, С 5-40 N Number of storage tanks V Tank volume, m D Control cool down rate via BRU-A Available

26 NPP-2006 safety Passive decay heat removal system via steam generators in NVNPP-2 design is intended for reactor core decay heat removal to ultimate heat sink (atmosphere) through the secondary side during beyond design basis accidents (e.g., station blackout) 4 3 канал 3 канал Parameter Value System power, MW 64(at +38 С) Number of HX in each channel 2 Control of cooling medium flow rate Available Air temperature range, С

27 NPP-2006 safety Passive decay heat removal system via steam generators in LNPP-2 design is intended for decay heat removal through the secondary side during beyond design basis accidents (e.g., station blackout) Parameters Value Q System power, MW 200 V Water inventory, m 3 4 х 540 τ Time of operation, hour 24 27

28 NPP-2006 safety Passive containment cooling system in LNPP-2 design is intended for long-term condensation of steam from containment atmosphere. 1 This system shares the offcontainment water storage tanks with passive decay heat removal system via steam generators tank; 2 steam line; 3 condensate line; 4 SG PHRS valve; 5 HX of containment PHRS; 6 steam generator; 7 cut-off valve

29 NPP-2006 safety The core catcher in NPP-2006 design provides corium confinement and exclude corium discharge outside the containment in any scenario. - Protects the reactor cavity against thermal and mechanical impact of corium -Takes in and accommodates solid and liquid corium constituents -Ensures formation of optimal structure and properties of the melt pool and subsequent solidification of corium. - Provides heat sink from corium to cooling water passively supplied min 24 h without any coolant makeup. - Provides corium retention - Minimises hydrogen and radionuclide release into containment on ex-vessel phase of a core melt scenario. 29

30 NPP-2006 safety Core catcher long term cooling provisions 1 - reactor; 2 core catcher; 3 fuel pool; 4 inspection vault for in-vessel components; 5 sump tanks; 6 pipeline supplying water onto corium surface; 7 pipeline supplying water to core catcher heat exchanger 30

31 NPP-2006 safety Passive catalytic recombiners are provided in NPP-2006 design to remove hydrogen from containment atmosphere 31

32 Conclusion VVER-1200 is developed as the main design for near term nuclear power program in Russia and for bidding at international market. VVER-1200 is the evolution of operating VVER-1000s by improving plant performance and increasing plant safety. VVER-1200 design is a basis for future VVER designs of wide power range. Passive systems are designed for VVER-1200 to back-up fulfillment of all main safety functions. These systems are capable to prevent typical BDBA progression to Fukushima-like scenarios. The viability of new passive systems implemented in VVER-1200 design is confirmed by extensive R&D works. A few nuclear power plants are already in operation and under construction with these systems implemented. 32