Transcript LEVEL 2 PSA SUPPORT TO SEVERE ACCIDENT MANAGEMENT
AN INTEGRATED APPROACH TO LIVING LEVEL 2 PSA
R. Himanen and H. Sjövall Teollisuuden Voima Oy, FIN-27160 Olkiluoto, Finland Presented at: INTERNATIONAL WORKSHOPONLEVEL 2 PSA AND SEVERE ACCIDENT MANAGEMENT COLOGNE, GERMANY 29TH TO THE 31ST OF MARCH 2004
Severe Accident Management in Olkiluoto 1 and 2 NPP • Asea-Atom BWR • Reactor thermal power 2500 MW • Net electric power 840 MW • Reactor pressure 7 MPa • Safety systems 4x50 % • Automatic liquid boron system for ATWS and ATWC • Pressure suppression containment • Containment inerted with nitrogen during normal operation • Drywell gas volume 4300 m 3 • Wetwell gas volume 3000 m 3 • Condensation pool volume 2700 m 3
Severe Accident Management in Olkiluoto 1 and 2 NPP • Containment design pressure 0.47 MPa • Containment ultimate capacity 1.01 MPa at 100 containment o C (95 % non exceedance probability) • Primary containment surrounded by reactor building acting as secondary •Severe accidents were not included in the original design.
•The provisions for severe accident management were installed in Olkiluoto 1 and 2 BWRs during the SAM project, which was finished in 1989.
•The SAM approach is hardware oriented.
•Plant modifications in order to prevent/withstand severe accident loads and minimize environmental consequences.
Severe Accident Management in Olkiluoto 1 and 2 NPP • Emergency Operating Procedure for severe accidents The Emergency Operating Procedure for severe accidents contains instructions for severe accident management and covers all phases of severe accident including a full core melt: - Primary system depressurisation - Flooding of lower drywell - Containment water filling - Procedures for filtered containment venting - Instructions to recover active core and containment cooling systems
PDS CBP RCO ROP COP HPL HPT LPL LPT RHL RHT VLL VEN FCF CM
Plant damage states and their frequences (Jan 2004)
10 -6 /ry 0.41
1.3
0.13
0.0072
0.045
3.6
0.61
8.5
0.22
2.2
0.00005
(51.) (11.) 17.
Description Containment by-pass (refuelling only) Reactivity control lost.
Very early reactor overpressurization Very early containment overpressurization LOCA initiated core melt begins early at high pressure Transient initiated core melt at high pressure LOCA initiated core melt at low pressure Transient initiated core melt at low pressure LOCA initiated late core melt due to loss of RHR Transient initiated late core melt due to loss of RHR Unsuccessful RHR using containment venting Successful RHR using containment venting (no CD) Fuel cladding failure (no CD) Total core damage frequency
Severe accident phenomena studied in level 2 PSA In-vessel issues: Steam explosion and other in-vessel fuel coolant interactions Recriticality Hydrogen generation Modes of vessel failure Ex-vessel issues: Containment issues: Direct containment heating Steam explosion and other ex-vessel fuel coolant interactions Generation of noncondensible gases Debris coolability in the lower drywell Core-concrete interaction Non-inert containment during start-up Direct containment bypass Containment venting, leakage and failure Basemat penetration
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Integrated simulation of physical and probabilistic models
Simple graphical presentation of CET
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”if–then–else” –statements inside the branching points
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Physical parameters transferred and modified in accident sequences
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Simulation of the phenomenon at branching point
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as a function of the input parameter set production of the output parameter set for next b.p.
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Simulation of the probability at branching point
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conditional probability of the branch as a function of the result of the simulation of the physical model
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Integration of accident progression and nuclide transportation models (1)
The analysis of source term and transportation of radio nuclides integrated into the simulation of each accident sequence
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No need for binning the CET sequences for this analysis
Integration of accident progression and nuclide transportation models (2)
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Time dependent transportation model
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Four dynamically sized control volumes
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LDW, UDW, WW gas volume, and reactor building
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Time dependent gas flow between volumes
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input parameters from MAAP
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Decontamination factors with uncertainty distributions
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pools filter containment spray deposition on surfaces
2,50
The strength of containment weak points
2,00 1,50 1,00 0,50 0,00 0 50 100 150 200
Temperature o C
250
Figure 8.5.4.1-1b: Containment weak points' break pressure as a function of temperature (median and confidence limits)
300 350 400 5% Door Door 95% Door 5%Eq Equipm 95%Eq 5%Ho Hoop 95%Ho 5%DoBY DomeBoltY 95%DoBY 5%DoSI DomeSeInst 95%DoSI 5%DoS70 DomeSe70h 95%DoS70 5% 361 361 95% 361 5% 362 362 95% 362
Severe Accident Management in Olkiluoto 1 and 2 NPP Level 2 PSA showed that the containment may break due to sum pressure of steam and noncondensible gas Modification in procedures: - Venting line isolation valve to be left open after initiating event.
- Possibility to fast automatic venting through the rupture disk line
Figure 5: Impact of modifications, summary
Figure 1: Venting line to be left open after IE(1997). Total LERF 7.9E-6/ry, unfiltered 7.0E-6/ry (89%)
E_VENT_U 11% L_VENT_U 0% REFUEL 5% L_CF 0% VE_VB_O2 2% E__VB_O2 18% E_CF 8% VE_UD_ 0% VE_CBP 0% E_CF 8% E_NFL 55% E__VB_O2 18% VE_VB_O2 2% L_CF 0% REFUEL 5% L_VENT_U 0% E_VENT_U 11% L_VENT_W 0% E_NFL 55%
Severe Accident Management in Olkiluoto 1 and 2 MODE PROJECT • Energetic ex-vessel fuel coolant interactions The range of the dynamic loading of steam explosions is estimated to be 10 to 30 kPas.
Regarding steam explosion loads the concrete structures are relatively stiff, particularly during the short period when the pressure waves are reflected.
Severe Accident Management in Olkiluoto 1 and 2 MODE PROJECT The median ultimate load impulse for the containment concrete structures, i.e. for the liner in the lowermost drywell wall sections corresponds to a rigid wall impulse of 54 kPas. The median ultimate load impulse for the personnel access lock was 6.3 kPas. The lower drywell access lock of Olkiluoto 1 was modified in 2001 and Olkiluoto 2 in 2002 so that it will sustain a steam explosion of 54 kPas. The personnel lock tube is fixed to the concrete wall so that the connection can resist a steam explosion.
Figure 5: Impact of modifications, summary
Figure 2: Lower containment air lock strenghtened (2001). Total LERF 7.4E-6/ry, unfiltered 5.8E-6/ry (79%)
L_VENT_W 0% E_VENT_U 21% VE_UD_ 0% VE_CBP 0% E_CF 4% L_VENT_U 0% REFUEL 6% L_CF 0% VE_VB_O2 2% E__VB_O2 19% E_NFL 47% VE_UD_ 0% VE_CBP 0% E_CF 4% E_NFL 47% E__VB_O2 19% VE_VB_O2 2% L_CF 0% REFUEL 6% L_VENT_U 0% E_VENT_U 21% L_VENT_W 0%
Severe Accident Management in Olkiluoto 1 and 2 SIMULATOR TRAINING • Failure to flood the LDW in time has almost 50% contribution to the LERF.
• Full scope simulator on site • All shifts were trained on the simulator once, and the flooding seems to succeed in time (2001) • Flooding of LDW trained also to the emergency organization in full scope emergency exercise (2002)
Figure 5: Impact of modifications, summary
Figure 3: LDW flooding – operators trained (2001). Total LERF 6.6E-6/ry, unfiltered 3.6E-6/ry (54%)
VE_UD_ 0% L_VENT_W 0% VE_CBP 0% E_CF 9% E_NFL 14% VE_UD_ 0% VE_CBP 0% E_CF 9% E_NFL 14% E__VB_O2 21% VE_VB_O2 3% L_CF 0% REFUEL 6% L_VENT_U 0% E_VENT_U 46% L_VENT_W 0% E_VENT_U 46% E__VB_O2 21% L_VENT_U 0% VE_VB_O2 3% L_CF 0% REFUEL 6%
What if?
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Inert start-up from refueling Several negative effects, like more difficult leakage check at start-up Benefit rather small
Figure 5: Impact of modifications, summary
Figure 4: Inert cmnt when start-up (option). Total LERF 6.4E-6/ry, unfiltered 2.9E-6/ry (46%)
E_VENT_U 54% E_CF 11% E_NFL 26% VE_UD_ 0% VE_CBP 0% E_CF 11% E_NFL 26% E__VB_O2 0% VE_VB_O2 0% L_CF 1% REFUEL 6% L_VENT_U 0% E_VENT_U 54% L_VENT_W 0% VE_VB_O2 0% E__VB_O2 0% REFUEL 6%
Summary of parts of level 2 PSA
Structural
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analysis of the strength of the containment details, strength against static and dynamic loads uncertainties before cut off Physics
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thermal hydraulics, phenomena, loads sequence specific source terms use of several codes, comparison of results not to be limited in ”representative” or ”worst” cases uncertainties before cut off Probabilistic
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accident sequences
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treatment of uncertainties (not cut off) importance ranking
Summary
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Structural Omission of detailed and realistic analyses with uncertainties may lead to biased risk profile
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Physical Omission of detailed plant and accident sequence specific analyses with sensitivity studies may lead to misunderstanding of uncertainty and biased risk profile
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Probabilistic Next page
Summary
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Probabilistic Level 2 PSA in SAM is like map and compass in orienteering Without them one can
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loose his way in the forest of structures or
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