Effect of passive safety systems on typical beyond-design accidents - - PowerPoint PPT Presentation

Effect of passive safety systems

• n typical beyond-design accidents

for WWER-1000/V-392 reactor plant

N.V.Boukine, L.N.Borisov, A.L.Gromov, N.S.Fil, A.M.Shumsky

EDO ”Gidropress”, Podolsk, Russia

Sixth International Information Exchange Forum “Safety Analysis for Nuclear Power Plants of VVER and RBMK Types” 8-12 April 2002; Kyiv, Ukraine

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Introduction (1/2)

The Russian regulatory documents for nuclear power plant safety (OPB-88/97) contain the requirement on the necessity of the beyond-design- basis accidents (BDBA) consideration as the events and scenarios participating in the formation of the relevant safety systems design basis. In particular, the list of such accidents have to be composed, the acceptance criteria are to be formulated and the realistic analysis of BDBAs have to be made.

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Introduction (2/2)

The designer should tend to the estimated probability of the limiting radioactivity release less than 10-7 per reactor-year, and the estimated probability of severe core damage derived on the PSA basis should not exceed 10-5 per reactor-year.

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Main characteristics of power unit

Core rated power 3000 MW Coolant pressure at core outlet 15,7 MPa Coolant flow rate through reactor 86000 m3/h Steam pressure at steam generator outlet 6,27 MPa

Examples advancements in the safety increasing area are as follows:

– advanced reactor WWER-1000; – passive system of residual heat removal (SPOT); – passive system for core flooding under loss-of-coolant accidents (HA-2); – passive system of quick boron supply to reactor; – primary coolant pump preventing coolant leak under long-term station blackout.

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Brief description of new passive systems

(HA-2)

The HA-2 system (JNG50-80) is intended to supply the boron solution to reactor with the purpose of long term (up to 24 h) cooling of the fuel during LOCAs of different size with active ECCS failure. The HA-2 system consists of four groups (four channels) of the tanks with 16 g/kg boron solution being under atmospheric pressure.

1 - ECCS hydroaccumulator

(HA-1)

2 - HA-2 tank (2 pcs.) 3 - reactor

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Brief description of new passive systems

(SPOT)

The passive heat removal system (JNB50-80) is intended for the long term residual heat removal under condition with complete loss of feedwater supply to SG in case of intact primary circuit. This system can also facilitate to the residual heat removal under certain scenarios of a loss of coolant accident.

1 - reactor; 2 - MCP; 3 - steam generator; 4 - air heat exchanger

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Beyond-design accidents without new passive systems

The following typical beyond-design accidents that essentially determine the design basis of the above passive systems are considered in this paper:

• station blackout with intact primary circuit;
• LB LOCA (double-ended cold leg break 850 mm

diameter) with 24 h station blackout. The “best estimate” approach is used when performing the calculations.

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Station blackout (1/3)

1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 t c 14.0 15.0 16.0 17.0 18.0 19.0 P 1 2 3 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 t c 290 300 310 320 330 340 350 360 T 1 2 3

Pressure at the core outlet Coolant temperature at the reactor outlet

1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – DINAMIKA-97

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Station blackout (2/3)

Collapsed level in upper plenum Maximum temperature of fuel rod claddings

1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – DINAMIKA-97

1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 t c 0.0 2.0 4.0 6.0 8.0 1 2 3 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 t c 200 400 600 800 1000 1200 1400 T 1 2 3

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Station blackout (3/3)

Time, s Event DINAMIKA-97 RELAP5/ MOD3.2 ATHLET 1.2A Beginning of the PRZ SV operation 1920 2550 2240 Steam generators emptying 7500 6400 6200 Beginning of the upper plenum boiling 4830 6600 5900 Termination of the natural circulation 6000 7200 6600 The maximum cladding temperature reached 1200 ° ° ° °С 8280 9500 8680

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Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (1/2)

Water inventory in the reactor Maximum temperature of fuel rod claddings

TECH-M-97 code (without HA-2 and SPOT)

50 100 150 200 250 300 t s 20 40 60 80 100 120 V m3 50 100 150 200 250 300 t s 200 400 600 800 1000 1200 1400 1600 T

• C

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Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (2/2)

Water inventory in the reactor Maximum temperature of fuel rod claddings

RELAP5/MOD3.2 code (without HA-2 and SPOT)

50 100 150 200 250 300 t s 20 40 60 80 100 120 V m3 50 100 150 200 250 300 t s 200 400 600 800 1000 1200 1400 1600 T

• C

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Beyond-design accidents with new passive systems

The results of the typical BDBA considered above indicate the necessity to provide for additional engineered features, intended to prevent the progression of a BDBA into severe accident. In the present chapter, the calculation results of the same typical BDBAs, but with new passive systems (HA-2, SPOT)

• peration are shown. It was assumed that all four channels of

this systems in operation. SPOT during the first period works in the control mode, and after 1800 s is switched over by operator to cooldown mode. The optimized (taking into account the pre-determined containment pressure change) dependence of the water flowrate from HA-2 was used in calculation.

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Station blackout (1/2)

Pressure at the core outlet Coolant temperature at the reactor outlet

1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – DINAMIKA-97

2000 4000 6000 8000 10000 12000 14000 t c 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 P 1 2 3 2000 4000 6000 8000 10000 12000 14000 t s 200 220 240 260 280 300 320 340 T

• C

1 2 3

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Station blackout (2/2)

Collapsed level in SG Maximum temperature of fuel rod claddings

1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – ДИНАМИКА-97

2000 4000 6000 8000 10000 12000 14000 t c 200 220 240 260 280 300 320 340 360 380 T 1 2 3 2000 4000 6000 8000 10000 12000 14000 t c 0.0 0.4 0.8 1.2 1.6 2.0 2.4 H 1 2 3

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Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (1/3)

Water inventory in the reactor 1 – pressure at the core outlet 2 – pressure in SG

TECH-M-97 code (with HA-2 and SPOT)

900 1800 2700 3600 4500 5400 6300 7200 t c 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 P 1 2 10800 21600 32400 43200 54000 64800 75600 86400 t c 20 40 60 80 100 120 V

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Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (2/3)

Maximum temperature of fuel rod claddings Mass of the primary coolant

TECH-M-97 code (with HA-2 and SPOT)

10800 21600 32400 43200 54000 64800 75600 86400 t c 200 400 600 800 1000 T 10800 21600 32400 43200 54000 64800 75600 86400 t c 20000 40000 60000 80000 100000 120000 140000 160000 180000 200000 220000 240000 M

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Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (3/3)

Maximum temperature of fuel rod claddings

RELAP5/MOD3.2 code (with HA-2 and SPOT)

10800 21600 32400 43200 54000 64800 75600 86400 t c 200 400 600 800 T 10800 21600 32400 43200 54000 64800 75600 86400 t c 20 40 60 80 100 120 V

3

Water inventory in the reactor

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Conclusion

Operation of the new passive systems (SPOT and HA-2) in considered beyond- design accidents provides a possibility of effective cooling of the core during required 24 hours of accident. This ensures the essentially decreased probability of severe core degradation.