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

Effect of passive safety systems on 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

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. 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 2

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. 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 3

Main characteristics of power unit Core rated power 3000 MW Coolant pressure at core outlet 15,7 MPa Coolant flow rate through reactor 86000 m 3 /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. 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 4

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 1 - ECCS hydroaccumulator channels) of the tanks with 16 g/kg (HA-1) boron solution being under 2 - HA-2 tank (2 pcs.) 3 - reactor atmospheric pressure. 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 5

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 1 - reactor; 2 - MCP; 3 - steam generator; loss of coolant accident. 4 - air heat exchanger 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 6

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. 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 7

Station blackout (1/3) P T 19.0 360 350 18.0 340 17.0 330 320 16.0 1 1 2 310 2 3 15.0 3 300 t t 14.0 290 c c 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 Pressure at the core outlet Coolant temperature at the reactor outlet 1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – DINAMIKA-97 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 8

Station blackout (2/3) T 8.0 1400 1 2 3 1200 6.0 1000 4.0 800 600 2.0 1 400 2 3 t t 0.0 200 c c 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 Collapsed level in upper plenum Maximum temperature of fuel rod claddings 1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – DINAMIKA-97 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 9

Station blackout (3/3) Event Time, s DINAMIKA-97 RELAP5/ ATHLET MOD3.2 1.2A Beginning of the PRZ SV operation 1920 2550 2240 Steam generators emptying 7500 6400 6200 Beginning of the upper plenum 4830 6600 5900 boiling Termination of the natural 6000 7200 6600 circulation The maximum cladding temperature 8280 9500 8680 reached 1200 ° ° ° ° С 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 10

Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (1/2) V T m 3 o C 120 1600 1400 100 1200 80 1000 60 800 40 600 20 400 t t 0 200 s s 0 50 100 150 200 250 300 0 50 100 150 200 250 300 Water inventory in the reactor Maximum temperature of fuel rod claddings TECH-M-97 code (without HA-2 and SPOT) 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 11

Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (2/2) V T m 3 o C 120 1600 1400 100 1200 80 1000 800 60 600 40 400 20 200 t t 0 0 s s 0 50 100 150 200 250 300 0 50 100 150 200 250 300 Water inventory in the reactor Maximum temperature of fuel rod claddings RELAP5/MOD3.2 code (without HA-2 and SPOT) 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 12

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) operation 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. 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 13

Station blackout (1/2) P T o C 18.0 340 1 1 2 2 16.0 320 3 3 14.0 300 12.0 280 10.0 260 8.0 240 6.0 220 t t 4.0 200 c s 0 2000 4000 6000 8000 10000 12000 14000 0 2000 4000 6000 8000 10000 12000 14000 Pressure at the core outlet Coolant temperature at the reactor outlet 1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – DINAMIKA-97 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 14

Station blackout (2/2) H T 2.4 380 1 1 2 360 2 3 2.0 3 340 1.6 320 300 1.2 280 260 0.8 240 0.4 220 t t 200 0.0 c c 0 2000 4000 6000 8000 10000 12000 14000 0 2000 4000 6000 8000 10000 12000 14000 Collapsed level in SG Maximum temperature of fuel rod claddings 1 – ATHLET 1.2A; 2 – RELAP5/MOD3.2; 3 – ДИНАМИКА-97 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 15

Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (1/3) P V 3 16.0 120 1 2 14.0 100 12.0 80 10.0 8.0 60 6.0 40 4.0 20 2.0 t t 0 0.0 c c 0 10800 21600 32400 43200 54000 64800 75600 86400 0 900 1800 2700 3600 4500 5400 6300 7200 1 – pressure at the core outlet Water inventory in the reactor 2 – pressure in SG TECH-M-97 code (with HA-2 and SPOT) 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 16

Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (2/3) M T 240000 1000 220000 200000 800 180000 160000 600 140000 120000 400 100000 80000 200 60000 40000 t t 0 20000 c c 0 10800 21600 32400 43200 54000 64800 75600 86400 0 10800 21600 32400 43200 54000 64800 75600 86400 Mass of the primary coolant Maximum temperature of fuel rod claddings TECH-M-97 code (with HA-2 and SPOT) 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 17

Main coolant pipeline break at reactor inlet (2x100% CL LOCA) with station blackout (3/3) V T 3 120 800 100 600 80 400 60 40 200 20 t t 0 0 c c 0 10800 21600 32400 43200 54000 64800 75600 86400 0 10800 21600 32400 43200 54000 64800 75600 86400 Water inventory in the reactor Maximum temperature of fuel rod claddings RELAP5/MOD3.2 code (with HA-2 and SPOT) 8-12 April 2002; Kyiv,Ukraine . Sixth International Information Exchange Forum 18