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L. J. Ott Oak Ridge National Laboratory

NRC Perspectives on Reactor Safety Course Special Features of BWR Severe Accident Mitigation and Progression. Appendix 2B-7 Module 3 Section 7 Module 4 Section 7. L. J. Ott Oak Ridge National Laboratory. BWR Severe Accident Studies Were Conducted at Oak Ridge National Laboratory 1980-1999.

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L. J. Ott Oak Ridge National Laboratory

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  1. NRC Perspectives on Reactor Safety CourseSpecial Features of BWR Severe Accident Mitigation and Progression Appendix 2B-7 Module 3 Section 7 Module 4 Section 7 L. J. Ott Oak Ridge National Laboratory

  2. BWR Severe Accident Studies Were Conducted at Oak Ridge National Laboratory 1980-1999 • Follow-on to NRC Severe Accident Sequence Analysis (SASA) Programs initiated late 1980 • Response to Three Mile Island • PWR studies • SNL • INL • LANL • BWR studies ORNL • Evaluations of Owners Group Emergency Procedure and Severe Accident Guidelines for NRR

  3. BWR Severe Accident Technology Activities at ORNL • Accident progression • Event sequence • Timing • Code application and model development • Analytical support of experiments • Pretest planning • Posttest analyses • Diverse locations • ACRR (Sandia) • NRU (Chalk River) • CORA (Karlsruhe) • Accident management strategies • Preventive • Mitigative • Extension to advanced reactor designs

  4. Predicted BWR Severe Accident Response Is Different from That Expected of a PWR in Several Aspects • Much more zirconium metal • Isolated reactor vessel • Reduction in power factor in the outer core region • Effects of safety relief valve actuations • Progressive relocation of core structures • Importance of core plate boundary • Steel structures in vessel • Large amount of water in vessel lower plenum

  5. Boiling Water Reactor Contributors to Core Damage Frequency – NUREG-1150

  6. Station Blackout Involves Failure of AC Electrical Power • Loss of offsite power • Emergency diesel-generators do not start and load Short-Term Station Blackout Immediate Loss ofWater Makeup Long-Term Station Blackout Loss of Water Makeup Following Battery Exhaustion

  7. The Most Probable BWR Accident Sequence Involving Loss of Injection Is Station Blackout Peach Bottom Short-term 5% Long-Term 42% Grand Gulf Short-term 96% Long-Term 1% Susquehanna* Short-term 52% Long-Term 10% Station Blackout Core Damage Frequencies *From Plant IPE (NPE 86-003)

  8. If the Reactor Vessel Remains Pressurized, Relocating Core Debris Falls into Water above the Core Plate Grand GulfShort TermStation Blackout without ADS Actuation

  9. Release of Debris Liquids through Penetration Internals Has Been Extensively Analyzed • Control rod drive mechanism penetrations: secure • Vessel drain: very improbable • Instrument tube: most likely internal path

  10. The BWR Control Rod Drive Mechanism Assemblies Are Held in Place by Upper Stub Tube Welds; the Incore Instrument Tubes Are Supported by Welds at the Vessel Wall

  11. The Drywell Floor Area Is Small and the Drywell Shell Is Within Ten Feet of the Pedestal Doorway

  12. Inside the Reactor Pedestal at Peach Bottom

  13. Lower Drywell at Browns Ferry

  14. BWR Evolution

  15. Comparison of ESBWR and ABWR • Key parameters that increase core flow in ESBWR • Shorter fuel • Tall chimney • Unrestricted downcomer

  16. Safety Systems Inside Containment Envelope

  17. Breakdown by Initiating Event

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