AN INTEGRATED APPROACH TO LIVING LEVEL 2 PSA R. Himanen and H. Sjövall Teollisuuden Voima Oy, FIN 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 o C (95 % non- exceedance probability) Primary containment surrounded by reactor building acting as secondary containment 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 The SAM approach is hardware oriented. Plant modifications in order to prevent/withstand severe accident loads and minimize environmental consequences.

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 Severe Accident Management in Olkiluoto 1 and 2 NPP

PDS10 -6 /ryDescription CBP0.41Containment by-pass (refuelling only) RCO1.3Reactivity control lost. ROP0.13Very early reactor overpressurization COP0.0072Very early containment overpressurization HPL0.045LOCA initiated core melt begins early at high pressure HPT3.6Transient initiated core melt at high pressure LPL0.61LOCA initiated core melt at low pressure LPT8.5Transient initiated core melt at low pressure RHL0.22LOCA initiated late core melt due to loss of RHR RHT2.22.2Transient initiated late core melt due to loss of RHR VLL Unsuccessful RHR using containment venting VEN(51.)Successful RHR using containment venting (no CD) FCF(11.)Fuel cladding failure (no CD) CM17.Total core damage frequency Plant damage states and their frequences (Jan 2004)

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: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 Containment issues: Non-inert containment during start-up Direct containment bypass Containment venting, leakage and failure Basemat penetration

Integrated simulation of physical and probabilistic models Simple graphical presentation of CET –”if–then–else” –statements inside the branching points Physical parameters transferred and modified in accident sequences Simulation of the phenomenon at branching point –as a function of the input parameter set –production of the output parameter set for next b.p. Simulation of the probability at branching point –conditional probability of the branch –as a function of the result of the simulation of the physical model

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 No need for binning the CET sequences for this analysis

Integration of accident progression and nuclide transportation models (2) Time dependent transportation model Four dynamically sized control volumes –LDW, UDW, WW gas volume, and reactor building Time dependent gas flow between volumes –input parameters from MAAP Decontamination factors with uncertainty distributions –pools –filter –containment spray –deposition on surfaces

The strength of containment weak points

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 Severe Accident Management in Olkiluoto 1 and 2 NPP

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%)

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%)

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%)

What if? 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%)

Summary of parts of level 2 PSA Structural –analysis of the strength of the containment –details, strength against static and dynamic loads –uncertainties before cut off Physics –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 –accident sequences –treatment of uncertainties (not cut off) –importance ranking

Summary Structural Omission of detailed and realistic analyses with uncertainties may lead to biased risk profile Physical Omission of detailed plant and accident sequence specific analyses with sensitivity studies may lead to misunderstanding of uncertainty and biased risk profile Probabilistic Next page

Summary Probabilistic Level 2 PSA in SAM is like map and compass in orienteering Without them one can –loose his way in the forest of structures or –go deep to the endless morass of physical phenomena