LEI Experience on modelling spent fuel pools during severe accident - - PowerPoint PPT Presentation

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

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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

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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.

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• 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.

Spent fuel pools of RBMK-1500 type reactor

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• 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.

Spent fuel pools of RBMK-1500 type reactor

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ROOM LAYOUT IN STORAGE PONDS HALL (ROOM 632)

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

Layout at elevation +25.20 m

Spent fuel pools of RBMK-1500 type reactor

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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.

Spent fuel pools of RBMK-1500 type reactor

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Wet storage and cooling in pools

Spent fuel pools of RBMK-1500 type reactor

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• 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 m3 – Initial temperature of water in SFP equal 50 oC – Water leakage from SFPs was assumed equal to maximal possible (21.11 kg/s)

LOCA accident in the spent fuel pools at maximal possible residual heat level

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2 4 6 8 10 12 14 16 18 20 25 50 75 100 125 Time, h Water level from pool bottom, m

ROD4 (ASTEC model) start of uncovering of fuel assemblies start of water supply all SFA are flooded by water ROD1,2,3 (ASTEC model) Fuel rods (RELAP/SCDAPSIM model)

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The sequence of analysed accident will be the following:

• t= 0 s – initiation of water leakage in

the SFP;

• t= 16.67 h – water level decreases

down to the top of SFA (fuel uncovering and heat up in air starts);

• t= 59.72 h – water level decreases

down to the bottom of SFA (all SFAs are fully uncovered);

• t= 65.28 h – water level decreases

down to the bottom of SFP (stop of water leakage from SFP);

• t= 83.33 h – water injection starts;
• t=87.50 h – water level increases up to

the bottom part of SFA;

• t=116.67 h – water level increases up

to the top of SFA (all SFAs and molten material are cooled down).

500 1000 1500 2000 2500 25 50 75 100 125 Time, h Temperature, oC

Fuel-bottom R-SCDAPSIM Fuel-top R-SCDAPSIM ROD1 ASTEC ROD4 ASTEC

discharge of water with constant flow rate 21.11 kg/s supply of water with constant flow rate 27.80 kg/s

2000 4000 6000 8000 10000 25 50 75 100 125 Time, h Hydrogen, kg. ASTEC RELAP5/MOD3 start of water supply

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• 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 m3 – Initial temperature of water in SFP equal 50 oC – 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

LOCA accident in the spent fuel pools 3 years after the shutdown of reactor

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2 4 6 8 10 12 14 16 18 20 30 60 90 120 150 Time, h Water level from pool bottom, m.

top of fuel assemblies in heat structure "12221" (ROD4) top of fuel assemblies in heat structures "12111", "12211", "12121" (ROD1-3) start of uncovering of fuel assemblies

100 200 300 400 500 50 100 150 200 250 300 350 400 450 500 550 Time, h Temperature, oC ROD1 ROD2 ROD3 ROD4

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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

• f 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);

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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.

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LOCA accident in the SFP of BWR type reactor

• SFP of the generic NPP is capable to

store 3055 spent fuel assemblies, with total decay heat power 3.565 MW.

• Assumed air temp. in the pool is 303

K and water 313 K.

• The initiating event is a leak through

broken 0.025 m diameter pipe (maximum leakage is equal to ~3 kg/s), connected to the bottom part

• f pool, together with total loss of

electrical power.

Parameter Pool length, m 12.2 Pool width, m 9.9 Water height, m 11.5 Pool height, m 11.8 Steel liner thickness, m 0.0635 Concrete wall thickness, m 2.0 Pool bottom thickness, m 2.0

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RELAP5 and RELAP/SCDAPSIM model of BWR SFP

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RELAP5 calculation results

Behavior of pool water collapsed level Comparison of integral water release and integral steam loss

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RELAP5 calculation results

Behavior of pool coolant temperature (top of active fuel) Temperature of SFP wall at bottom, middle and upper part

• f the pool

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RELAP/SCDAPSIM calculation results

Water level in the pool Water and steam released from the pool

0 – t1 – time up to the time when the water level decreases to the top part of the active fuel; t1 – t2 – temperature increase in fuel rods (the duration of t1 – t2 interval is ~16 h); at the end of this interval first material relocation (1189 kg mass of liquefied material) to lower part of the SFP occurs; t2 – t3 – water level decreases down to the bottom of the fuel assemblies; the duration of t1 – t3 interval is ~35 h; t3 – t4 – water level decreases down to the bottom of pool – SFP is empty; the duration of t1 – t4 interval is ~49 h;

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The calculation using RELAP/SCDAPSIM started when the water level in the pool is at ~4m (top of the active fuel) – this condition is assumed as the starting point (t = 0 s).

RELAP/SCDAPSIM calculation results

Cladding temperatures in the first group upper and lower parts of the fuel assemblies

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Conclusions

• Modelling of a LOCA in a BWR spent fuel pool using RELAP5 and

RELAP/SCDAPSIM showed that:

– Calculation results showed that after initiation of the LOCA, the amount of evaporated water is lower than the water lost through the leakage. However, when the water level decreases to the top of active fuel water lost through the leakage is lower than amount of evaporated water. Only when the water level decreased to the bottom of the fuel assemblies, steam generation decreased and the dominant water release from the pool is due to water discharged through the break. However, steam is also produced even though there is no water in contact with the fuel assembly. This is due to the molten fuel and metal slumping to the lower part of the spent fuel pool and the water in the bottom of the pool comes in contact with the hot surfaces of the walls. – first molten material relocation occurred ~76 h from initiation of the LOCA (~16 h later when water level reaches top of active fuel ). The water level decreased downward to the bottom part of the fuel assemblies after ~95 h from initiation

• f the LOCA (~35 h later when water level reaches top of active fuel ). The

water level decreased down to the bottom of pool after ~109 h from initiation

• f the LOCA (~49 h later when water level reaches top of active fuel )

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General Conclusions

• This presentation the calculation results of the hypothetical beyond

design basis accident in spent fuel pools at RBMK-1500 (Ignalina NPP) and BWR-4 type reactors – loss of cooling water, which leads to the loss of heat removal. For the analysis the SFPs models were developed using the RELAP5, RELAP/SCDAPSIM and ASTEC codes. The developed models allowed to model different phenomena: uncovering and heat-up of fuel rods, steam–zirconium reaction, quenching of hot fuel rods by water, etc.

• The performed analyses are useful for the evaluation of different

accident mitigation measures.

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Tadas Kaliatka E-mail: Tadas.Kaliatka@lei.lt

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