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Slide 1 NRC Perspectives on Reactor Safety Course Special Features of BWR Severe Accident Mitigation and Progression L. J. Ott Oak Ridge National Laboratory.

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Presentation on theme: "Slide 1 NRC Perspectives on Reactor Safety Course Special Features of BWR Severe Accident Mitigation and Progression L. J. Ott Oak Ridge National Laboratory."— Presentation transcript:

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

2 Slide 2 BWR Severe Accident Studies Were Conducted at Oak Ridge National Laboratory  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 Slide 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 Slide 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 Slide 5 Boiling Water Reactor Contributors to Core Damage Frequency – NUREG-1150

6 Slide 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 of Water Makeup Long-Term Station Blackout Loss of Water Makeup Following Battery Exhaustion

7 Slide 7 The Most Probable BWR Accident Sequence Involving Loss of Injection Is Station Blackout Peach Bottom Short-term5% Long-Term42% Grand Gulf Short-term96% Long-Term1% Susquehanna* Short-term52% Long-Term10% Station Blackout Core Damage Frequencies *From Plant IPE (NPE )

8 Slide 8 If the Reactor Vessel Remains Pressurized, Relocating Core Debris Falls into Water above the Core Plate Grand Gulf Short Term Station Blackout without ADS Actuation

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

12 Slide 12 Inside the Reactor Pedestal at Peach Bottom

13 Slide 13 Lower Drywell at Browns Ferry

14 Slide 14 BWR Evolution

15 Slide 15 Comparison of ESBWR and ABWR  Key parameters that increase core flow in ESBWR  Shorter fuel  Tall chimney  Unrestricted downcomer

16 Slide 16 Safety Systems Inside Containment Envelope

17 Slide 17

18 Slide 18 Breakdown by Initiating Event


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