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L. J. Ott Oak Ridge National Laboratory

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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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slide1

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

bwr severe accident studies were conducted at oak ridge national laboratory 1980 1999
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
bwr severe accident technology activities at ornl
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
predicted bwr severe accident response is different from that expected of a pwr in several aspects
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
station blackout involves failure of ac electrical power
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

the most probable bwr accident sequence involving loss of injection is station blackout
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)

slide8
If the Reactor Vessel Remains Pressurized, Relocating Core Debris Falls into Water above the Core Plate

Grand GulfShort TermStation Blackout without ADS Actuation

release of debris liquids through penetration internals has been extensively analyzed
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
slide10

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

the drywell floor area is small and the drywell shell is within ten feet of the pedestal doorway
The Drywell Floor Area Is Small and the Drywell Shell Is Within Ten Feet of the Pedestal Doorway
comparison of esbwr and abwr
Comparison of ESBWR and ABWR
  • Key parameters that increase core flow in ESBWR
    • Shorter fuel
    • Tall chimney
    • Unrestricted downcomer
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