LEI Experience on modelling spent fuel pools during severe accident conditions T. Kaliatka, V. Vileiniskis, A. Kaliatka and E. Uspuras Laboratory of Nuclear Installations Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas, Lithuania Technical Meeting on the Phenomenology, Simulation and Modelling of Accidents at Spent Fuel Pools September 2-5, 2019 – IAEA, Vienna, Austria 1
Content • Introduction • LEI experience on the modelling of spent fuel pools during severe accident conditions: • Ignalina NPP (RBMK-1500) – The evaluation of the worst possible consequences, assuming the maximal amount of Spent Fuel Assemblies (SFA) in the pools and the maximal possible residual heat of nuclear fuel. – Situation with three years after the permanent shutdown of Unit 2 reactor of Ignalina NPP. • BWR (BWR-4, Mark I) – LOCA accident in the spent fuel pool based on a generic model of the BWR-4 boiling water reactor. • Main conclusions on the modelling 2
Introduction • Tsunami that followed the earthquake at the Fukushima Daiichi nuclear plants in Japan showed that a loss of heat removal in the spent fuel pools may lead to very serious consequences. • Consequences of such an accident can be very serious creating a possibility of significant amount of radioactive material release to the environment. This is because spent fuel pools are in general not housed in a containment with the same integrity as the containment around the reactor core and primary pressure boundary. Thus, the loss of water, which leads to the loss of heat removal in SFP may lead to very serious consequences. 3
Spent fuel pools of RBMK-1500 type reactor • At the Ignalina NPP (Lithuania) two Russian design channel-type graphite-moderated boiling water reactors (RBMK-1500) are shutdown for decommissioning (in 2004 and 2009). • The SFP of both units have been full of spent fuel for a long time due to the long delay (more than 5 years) in the construction of the new interim storage facility, which was commissioned only in 2018. Therefore, analysis of accidents in the SFP is a very important task at Ignalina NPP. 4
Spent fuel pools of RBMK-1500 type reactor • The RBMK fuel assembly consists of two fuel bundles 3,5 m long, placed one above the other. • The reloaded from the RBMK-1500 reactor fuel assemblies remain in the pool for at least a year, after which they may be removed to be cut in a “ hot ” cell. During this procedure the two fuel bundles are separated and placed into the special shipping casks. • The shipping casks with spent fuel assemblies are stored in the storage pools until they are loaded into the protective casks CASTOR or CONSTOR to be further transported to the dry spent fuel storage facility. 5
Spent fuel pools of RBMK-1500 type reactor ROOM LAYOUT IN STORAGE PONDS HALL (ROOM 632) Layout at elevation +25.20 m Reactor Hall Storage Ponds Hall Rooms 157; 235 - Transfer canyon Room 234 - Transfer canyon of Cutting Bay Room 236/1,2 - SFA Cooling Ponds Rooms 336; 337/1,2; 339/1,2 - Storage Ponds for fuel loaded in 102 pcs. Baskets Room 338/1,2 - Ponds for loading of transport containers Room 625 - HC Protective chamber Room 627 - HC Control Room 6
Spent fuel pools of RBMK-1500 type reactor The spent fuel storage and handling system consists of 12 pools: • 2 pools used for cooling uncut spent fuel extracted from the reactor ( Sections 236/1, 236/2 , bottom – at elevation+7.20, top- at elevation+25.20 ); • 5 pools used for storing spent fuel in 102 pcs baskets after cutting ( Sections 336, 337/1,2, 339/1,2 bottom – at elevation +13.00 m ); • 1 pool used for collecting spent fuel assemblies prepared for cutting, cutting hanger from fuel assembly; transporting spent fuel assembly to the hot cell and 102 pcs. baskets in the hot cell and backward to pools ( Section 234, bottom – at elevation +7.70 m ); • 2 pools (338/1 ( bottom – at elevation +13.0/+17.0 m ) and 338/2) intended to transport spent fuel assemblies and transport baskets loaded with spent fuel assemblies between the pools and 2 transport canals (235 and 157 ( at elevation 0 - +24.75. Volume – 773m3. Length – 26m, width – 1.5m )) used for handling with fresh ant spent FA, 102 pcs. baskets and casks. 7
Spent fuel pools of RBMK-1500 type reactor Wet storage and cooling in pools 8
LOCA accident in the spent fuel pools at maximal possible residual heat level • Analysis of loss of heat removal accident in the SFPs of Ignalina NPP was performed using the codes for severe accident analysis ASTEC and RELAP/SCDAPSIM. – Total 7901 SFAs with total decay heat of SFP (4253 kW) – Total volume of water in the SFP is 5070 m 3 – Initial temperature of water in SFP equal 50 o C – Water leakage from SFPs was assumed equal to maximal possible (21.11 kg/s) 20 18 Fuel rods (RELAP/SCDAPSIM Water level from pool bottom, m all SFA are flooded by water model) 16 14 start of uncovering of fuel assemblies 12 ROD4 (ASTEC model) 10 8 ROD1,2,3 (ASTEC model) 6 start of water supply 4 2 0 -2 0 25 50 75 100 125 Time, h 9
The sequence of analysed accident will be supply of water with Fuel-bottom R-SCDAPSIM constant flow rate 27.80 kg/s 2500 Fuel-top R-SCDAPSIM the following: ROD1 ASTEC ROD4 ASTEC • 2000 t= 0 s – initiation of water leakage in Temperature, o C the SFP; 1500 discharge of water with • constant flow rate 21.11 kg/s t= 16.67 h – water level decreases 1000 down to the top of SFA (fuel uncovering and heat up in air starts); 500 • t= 59.72 h – water level decreases 0 down to the bottom of SFA (all SFAs 0 25 50 75 100 125 Time, h are fully uncovered); 10000 • t= 65.28 h – water level decreases ASTEC 8000 RELAP5/MOD3 down to the bottom of SFP (stop of Hydrogen, kg. water leakage from SFP); 6000 • t= 83.33 h – water injection starts; 4000 • t=87.50 h – water level increases up to start of water supply 2000 the bottom part of SFA; • t=116.67 h – water level increases up 0 0 25 50 75 100 125 to the top of SFA (all SFAs and molten Time, h material are cooled down). 10
LOCA accident in the spent fuel pools 3 years after the shutdown of reactor • Analysis of loss of heat removal accident in the SFPs of Ignalina NPP was performed using ASTEC code. – Total decay heat of SFP - 810 kW – Total volume of water in the SFP is 5070 m 3 – Initial temperature of water in SFP equal 50 o C – Water leakage from SFPs was assumed equal to maximal possible (21.11 kg/s) – Water is not injected into the SFP, the spent fuel assemblies are cooled by air after pools are empty 20 500 Water level from pool bottom, m. 18 start of uncovering of fuel assemblies 16 400 14 Temperature, o C top of fuel assemblies in heat structure "12221" (ROD4) 12 300 10 top of fuel assemblies in heat structures "12111", "12211", "12121" (ROD1-3) 8 ROD1 ROD2 200 6 4 ROD3 ROD4 100 2 0 0 -2 0 50 100 150 200 250 300 350 400 450 500 550 0 30 60 90 120 150 Time, h Time, h 11
Conclusions • The results of the RBMK-1500 analysis showed that: – assuming theoretically maximal possible residual heat of fuel assemblies in SFP (4253 kW), the late operator action can lead to the generation of huge amount of hydrogen, failure of fuel claddings and release of radioactive isotopes to the environment; – for the situation – more as three years after the permanent shutdown of Unit 2 reactor of Ignalina NPP. The total decay heat of spent nuclear fuel in SFP of Unit 2 decreased down to 810 kW. In the case of total loss of water, the increase of fuel temperature is very slow. The preliminary analysis using ASTEC code showed that SFAs can be cooled by the air circulation (in the case if all SFP steal covers will be removed/opened); 12
LOCA accident in the SFP of BWR type reactor • The basic data of a BWR-4 type reactor with Mark-I type containment were used preparing this generic input deck. • Generic input deck was developed in the framework of the CESAM project (Euratom 7th FP under grant agreement No 604965). • For modelling of the processes in the SFP during a severe accident, the RELAP5 and RELAP/SCDAPSIM codes are used. 13
LOCA accident in the SFP of BWR type reactor • SFP of the generic NPP is capable to Parameter store 3055 spent fuel assemblies, with total decay heat power 3.565 MW. Pool length, m 12.2 • Assumed air temp. in the pool is 303 Pool width, m 9.9 K and water 313 K. Water height, m 11.5 • The initiating event is a leak through Pool height, m 11.8 broken 0.025 m diameter pipe (maximum leakage is equal to ~3 Steel liner thickness, m 0.0635 kg/s), connected to the bottom part Concrete wall thickness, m 2.0 of pool, together with total loss of electrical power. Pool bottom thickness, m 2.0 14
RELAP5 and RELAP/SCDAPSIM model of BWR SFP 15
RELAP5 calculation results Behavior of pool water collapsed level Comparison of integral water release and integral steam loss 16
RELAP5 calculation results Behavior of pool coolant temperature (top of active fuel) Temperature of SFP wall at bottom, middle and upper part of the pool 17
Recommend
More recommend
