Slide 1

NRC Perspectives on Reactor Safety Course

Special Features of BWR Severe Accident Mitigation and Progression

L. J. OttOak Ridge National Laboratory

Appendix 2B-7Module 3 Section 7Module 4 Section 7

Slide 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

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

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

Slide 5

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

Slide 6

Station Blackout Involves Failure of AC Electrical Power Loss of offsite power Emergency diesel-generators do not start and load

Short-TermStation Blackout

Immediate Loss ofWater Makeup

Long-TermStation Blackout

Loss of Water Makeup Following Battery

Exhaustion

Slide 7

The Most Probable BWR Accident Sequence Involving Loss of Injection Is Station Blackout

Peach BottomShort-term 5%Long-Term 42%

Grand GulfShort-term 96%Long-Term 1%

Susquehanna*Short-term 52%Long-Term 10%

Station Blackout CoreDamage Frequencies

*From Plant IPE (NPE 86-003)

Slide 8

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

Grand GulfShort Term

Station Blackout without ADS

Actuation

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

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

Slide 11

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

Slide 12

Inside the Reactor Pedestal at Peach Bottom

Slide 13

Lower Drywell at Browns Ferry

Slide 14

BWR Evolution

Slide 15

Comparison of ESBWR and ABWR

Key parameters that increase core flow in ESBWR Shorter fuel Tall chimney Unrestricted downcomer

Slide 16

Safety Systems Inside Containment Envelope

Slide 17

Slide 18

Breakdown by Initiating Event