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State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Accident Analysis Presented at the Presented at the th Regulatory Information Conference USNRC 20 USNRC 20 th Regulatory Information Conference Washington, DC Washington, DC

SLIDE 1

Presented at the Presented at the USNRC 20 USNRC 20th

th Regulatory Information Conference

Regulatory Information Conference Washington, DC Washington, DC March 11, 2008 Randall Gauntt, Sandia National Laboratories Charles Tinkler, USNRC

Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project

Accident Analysis

Slide 2 of 26

SOARCA Objectives

Perform a state-of-the-art, realistic evaluation of severe

accident progression, radiological releases and offsite consequences for important accident sequences

– Phenomenologically based, consistent, integral analyses of radiological source terms

Provide a more realistic assessment of potential offsite

consequences to replace previous consequence analyses

– 1982 Siting Study

Slide 3 of 26

SOARCA Accident Progression Modeling Approach

Full power operation
Plant-specific sequences with a CDF>10-6 (CDF>10-7 for bypass

events)

External events included
Consideration of all mitigative measures, including B.5.b
Sensitivity analyses to assess the effectiveness of different

safety measures

State-of-the-art accident progression modeling based on 25

years of research to provide a best-estimate for accident progression, containment performance, time of release and fission product behavior

SLIDE 2 Slide 4 of 26

1982 Siting Study

Evaluated potential consequences relevant to generic

siting criteria

Used hypothesized, generalized, source term

categories

– Based on limited knowledge and bounding rationale – Uncoupled from specific plant design or specific sequences

Consequences dominated by

– Source term magnitude and timing – Population density – Emergency response

Slide 5 of 26

Radiological Source Terms

1982 Siting Study results were dominated by the SST1 source

term – Loss of safety features – Large FP release from core – Severe early reactor and containment failure or bypass

1982 SST1 characterization (magnitude, timing and frequency)

reflected then state of understanding and modeling – Early containment failure modes contemporaneously cited included alpha mode (steam explosion) failure, direct containment heating, hydrogen combustion

Research and plant improvements over 25 years have

dramatically altered our view of the early failure modes

Slide 6 of 26

Severe Accident Improvements

Research/plant improvements provided bases to conclude

that some presumed early containment failure modes have been shown to be

– negligible/highly improbable

In-vessel steam explosion and alpha mode failure
SERG, Sizewell PRA, Experiments (FARO, KROTOS, TROI)
direct containment heating due to high pressure melt ejection
DCH Issue Resolution, experiments at SNL, ANL, Purdue

– or can be prevented by accident management

BWR Mark I liner melt through
Hydrogen control systems
For large dry concrete containments, increased containment

leakage is failure mode (vs catastrophic failure of the containment)

SLIDE 3 Slide 7 of 26

Preliminary SOARCA Findings

No sequences could be identified which resemble the

characteristics of the dominant sequence from the 1982 study sequences – Sequences which were identified have lower frequencies than that assigned to SST1 in 1982 study

All sequences identified could be prevented or significantly

mitigated by existing or recently developed plant improvements – Important to realistically treat plant features/capabilities and include in probabilistic assessments – Confirmed by MELCOR analyses and served as the basis for evaluating plant/operator response including the TSC

Slide 8 of 26

Preliminary SOARCA Findings

Containment failure or bypass sequences are still identified in

some plant specific PRA but even in those instances severity

f conditions are significantly reduced

– Reactor vessel lower head failure delayed even for the most severe (and most remote) of sequences (~ 7- 8 hrs) and much delayed for more likely severe sequences ( ~20+ hrs) – Bypass events are delayed beyond timing of SST1, bypass events also reflect scrubbed releases due to submergence of break (consistent, mechanistic modeling) or fission product deposition in the system piping

These conditions while identified as important in current/past

PRA, may now be considered to be more amenable to mitigation because of timing (revealed by integral analyses) and plant capabilities

Slide 9 of 26

Preliminary SOARCA Findings

Without those mitigation strategies, sensitivity studies indicate

a radiological release fraction which is significantly smaller than earlier studies.

Unmitigated sensitivities also result in a delayed release

SLIDE 4 Slide 10 of 26

Peach Bottom Atomic Power Station Emergency (B.5.b) Equipment

Portable power source for SRVs and level indication
Manual operation of RCIC without dc power
Portable diesel driven pump (250 psi, 500 gpm) to makeup to RCS,

drywell, CST, Hotwell, etc. and provide external spray

Portable air supply to operate containment vent valves
Off-site pumper truck can be used in place of portable diesel driven

pump

Slide 11 of 26

Peach Bottom Atomic Power Station

Long-term Station Blackout Without Mitigation

Without B.5.b mitigation

– Accident progression

Core uncovery in 9 hrs Core damage in 10 hrs RPV and containment failure in 20 hrs, start of radioactive release, (liner melt-through or containment head flange leakage) Time between start of evacuation and radioactive release: ~17 hrs

– Offsite radioactive release is relatively small

1 – 4 % release of volatiles, except noble gases Release is much less severe than 1982 Siting Study

– Accident progression timing and emergency evacuation significantly reduce potential consequences

Slide 12 of 26

Peach Bottom Atomic Power Station

Long-term Station Blackout With Mitigation Swollen Vessel Water Level Response

SLIDE 5 Slide 13 of 26

Preliminary Findings Summary

B.5.b measures have potential to prevent or significantly delay core

damage

Without B.5.b mitigative measures

– Releases are significantly lower than 1982 study – Releases can be significantly delayed

Accident progression timing (long time to core damage and

containment failure) and mitigative measures significantly reduce the potential for core damage and/or containment failure

Slide 14 of 26

Peach Bottom Atomic Power Station

Long-term Station Blackout Without Mitigation Swollen Vessel Water Level Response

100 200 300 400 500 600 700 800 2 4 6 8 10 12 14 16 18 20 22 24 time (hr) Two Phase Mixture Level [in] In-Shroud Downcomer TAF BAF Main Steam Nozzle RPV Water Level Automatic RCIC actuation Operator takes manual control of RCIC RCIC steam line floods Initial debris relocation into lower head +5 to +35" Batteries exhaust

SRV recloses

Slide 15 of 26

Peach Bottom Atomic Power Station

Long-term Station Blackout Without Mitigation

Iodine Fission Product Distribution 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 5 10 15 20 25 30 35 40 45 50 time [hr] Fraction of Initial Core Inventory Release to environment (3.7%) Captured in Suppression Pool Deposited/Airborne within RPV Drywell (mostly airborne)

SLIDE 6 Slide 16 of 26

Peach Bottom Atomic Power Station

Long-term Station Blackout Without Mitigation

Cesium Fission Product Distribution 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 5 10 15 20 25 30 35 40 45 50 time [hr] Fraction of Initial Core Inventory Release to environment (1.8%) Captured in Suppression Pool Deposited/Airborne within RPV Drywell negligible

Slide 17 of 26

Surry Nuclear Station Emergency (B.5.b) Equipment/Procedures

2 diesel-driven high-pressure skid-mounted pumps for

injecting into the RCS

1 diesel-driven low-pressure skid-mounted pump for

injecting into steam generators or containment

Portable power supply for restoring indication
Portable air bottles to operate SG PORVs
Manual operation of TDAFW
Spray nozzle (located on site fire truck) for scrubbing fission

product release

Slide 18 of 26

Surry Power Station

Long-term Station Blackout With Mitigation Swollen Vessel Water Level Response

Vessel Water Levels LTSBO - Mitigation with Portable Equipment

2 4 6 8 10 3 6 9 12 15 18 21 24 Time (hr) Two-Phase Level (m) Accumulators Start RCS cooldown BAF TAF Lower head Vessel top Start RCS injection with portable pump

SLIDE 7 Slide 19 of 26

Surry Power Station

Short-term Station Blackout With Mitigation (Emerg. CS) Swollen Vessel Water Level Response

Vessel Water Level STSBO -Mitigation with Portable Equipment

2 4 6 8 10 1 2 3 4 5 6 7 8 Time (hr) Two-Phase Level (m) Accumulators BAF TAF Lower head Vessel top Vessel failure Slide 20 of 26 Fission Product Release to the Environment STSBO - Mitigated with portable equipment 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1 2 3 4 Time (days) Fraction release (-) NG I Cs < 0.003%

Surry Power Station

Short-term Station Blackout With Mitigation (Emerg. CS)

Slide 21 of 26

Surry Power Station

ISLOCA With Mitigation Swollen Vessel Water Level Response

Vessel Water Level ISLOCA- Mitigation with Unaffected Unit's Equipment

2 4 6 8 10 3 6 9 12 15 18 21 24 Time (hr) Two-Phase Level (m) Accumulators Start RHR BAF TAF Lower head Vessel top Shift to HL Injection

SLIDE 8 Slide 22 of 26

Surry Power Station

ISLOCA With Mitigation ISLOCA mitigated using Second Unit RWST

RWST Water Volumes ISLOCA - Mitigated with Unit #2 Equipment 50000 100000 150000 200000 250000 300000 350000 400000 6 12 18 24 30 36 Time (hr) Volume (gal) RWST #1 RWST #2 Refilled by 150 gpm make-up Isolate RWST at 1.75 hr Slide 23 of 26

Mitigative Measures Sensitivity Analysis

Without mitigative measures

– Long term SBO

Core damage at 16 hrs Containment failure at 45 hrs (increased containment leakage) Public evacuation begins at 2.5 hrs

– Short term SBO

Core damage at 3 hrs Containment failure at 25 hrs Public evacuation begins at 2.5 hrs

– ISLOCA

Release scrubbed in flooded Aux building room Non-mitigated analysis ongoing

– SGTR

Unsuccessful mitigation not considered credible >40 hrs to core damage and offsite release

Slide 24 of 26

Surry Power Station

Long-term Station Blackout Without Mitigation Swollen Vessel Water Level Response

Vessel Water Levels LTSBO - No Mitigation with Portable Equipment

2 4 6 8 10 3 6 9 12 15 18 21 24 Time (hr) Two-Phase Level (m) Accumulators Hot leg creep rupture failure Batteries exhausted S/G dryout Start RCS cooldown ECST Empty RCP Seal Failures PORVs open Accumulators BAF TAF Lower head Vessel top

SLIDE 9 Slide 25 of 26

Surry Power Station

Long-term Station Blackout Without Mitigation

Fission Product Release to the Environment LTSBO - No Mitigation, Calculated RCP Seal Failure 0.000 0.002 0.004 0.006 0.008 0.010 1 2 3 4 Time (days) Fraction release (-) I Cs Slide 26 of 26

Surry Power Station

Short-term Station Blackout Without Mitigation

Fission Product Release to the Environment Unmitigated STSBO 0.00 0.01 0.02 0.03 0.04 0.05 1 2 3 4 Time (days) Fraction release (-) I Cs ~1% at 4 days Slide 27 of 26

Surry Station Blackouts Compared to SST-1

Surry SOARCA Environmental Release Fractions 0.000 0.100 0.200 0.300 0.400 0.500 0.600 0.700 0.800 0.900 1.000 1 2 3 4 Time (days) Release Fraction (-) LTSBO - I LTSBO - Cs STSBO - I STSBO - Cs SST-1 Iodine SST-1 Cesium Surry SOARCA Environmental Release Fractions 0.000 0.002 0.004 0.006 0.008 0.010 0.012 0.014 0.016 0.018 0.020 1 2 3 4 Time (days) Release Fraction (-) LTSBO - I LTSBO - Cs STSBO - I STSBO - Cs SST-1 Iodine SST-1 Cesium

SLIDE 10 Slide 28 of 26

Peach Bottom Long Term Station Blackout Compared to SST-1

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 8 16 24 32 40 48 time [hr] Fraction of Initial Core Inventory

Cesium to Environment Iodine to Environment SST-1 Iodine SST-1 Cesium Slide 29 of 26

Summary

SOARCA study completing evaluation of

Surry and Peach Bottom plants

Releases for unmitigated accident vastly

reduced and delayed in time compared to SST-1

Mitigation shown to capable of terminating

accidents

Sequoyah analysis getting underway
Uncertainty analysis and peer review planned