Presented at the Presented at the USNRC 20USNRC 20thth Regulatory Information Conference Regulatory Information Conference

Washington, DCWashington, DCMarch 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 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 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 of 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 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 hrsCore 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 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

0

100

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800

0 2 4 6 8 10 12 14 16 18 20 22 24

time (hr)

Tw

o P

has

e M

ixtu

re L

evel

[in

]

In-ShroudDowncomerTAFBAFMain Steam Nozzle

RPV Water Level

Automatic RCICactuation

Operator takes manualcontrol of RCIC

RCIC steamline floods

Initial debrisrelocation intolower 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

0.1

0.2

0.3

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0.9

0 5 10 15 20 25 30 35 40 45 50

time [hr]

Fra

ctio

n o

f In

itia

l C

ore

In

ven

tory

Release toenvironment (3.7%)

Captured in Suppression Pool

Deposited/Airbornewithin RPV

Drywell (mostly airborne)

Slide 16 of 26

Peach Bottom Atomic Power Station Long-term Station Blackout Without Mitigation

Cesium Fission Product Distribution

0

0.1

0.2

0.3

0.4

0.5

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0.9

0 5 10 15 20 25 30 35 40 45 50

time [hr]

Fra

cti

on

of

Init

ial C

ore

Inv

ento

ry

Release toenvironment (1.8%)

Captured in Suppression Pool

Deposited/Airbornewithin 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 LevelsLTSBO - Mitigation with Portable Equipment

-4

-2

0

2

4

6

8

10

0 3 6 9 12 15 18 21 24

Time (hr)

Tw

o-P

ha

se L

evel

(m

)

Accumulators

Start RCS cooldown

BAF

TAF

Lower head

Vessel top

Start RCS injection with portable pump

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

-4

-2

0

2

4

6

8

10

0 1 2 3 4 5 6 7 8

Time (hr)

Tw

o-P

has

e L

evel

(m

)

Accumulators

BAF

TAF

Lower head

Vessel top

Vessel failure

Slide 20 of 26

Fission Product Release to the EnvironmentSTSBO - Mitigated with portable equipment

0

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0.9

1

0 1 2 3 4

Time (days)

Fra

ctio

n r

ele

as

e (-

)

NGICs

< 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

-4

-2

0

2

4

6

8

10

0 3 6 9 12 15 18 21 24

Time (hr)

Tw

o-P

ha

se

Le

vel

(m)

Accumulators

Start RHR

BAF

TAF

Lower head

Vessel top

Shift to HL Injection

Slide 22 of 26

Surry Power Station ISLOCA With Mitigation

ISLOCA mitigated using Second Unit RWST

RWST Water VolumesISLOCA - Mitigated with Unit #2 Equipment

0

50000

100000

150000

200000

250000

300000

350000

400000

0 6 12 18 24 30 36

Time (hr)

Vo

lum

e (

ga

l)

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 hrsContainment failure at 45 hrs (increased containment

leakage)Public evacuation begins at 2.5 hrs

– Short term SBOCore damage at 3 hrsContainment failure at 25 hrsPublic evacuation begins at 2.5 hrs

– ISLOCARelease scrubbed in flooded Aux building roomNon-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 LevelsLTSBO - No Mitigation with Portable Equipment

-4

-2

0

2

4

6

8

10

0 3 6 9 12 15 18 21 24

Time (hr)

Tw

o-P

has

e L

evel

(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 25 of 26

Surry Power Station Long-term Station Blackout Without Mitigation

Fission Product Release to the EnvironmentLTSBO - No Mitigation, Calculated RCP Seal Failure

0.000

0.002

0.004

0.006

0.008

0.010

0 1 2 3 4

Time (days)

Fra

ctio

n r

ele

as

e (-

)

I

Cs

Slide 26 of 26

Surry Power Station Short-term Station Blackout Without Mitigation

Fission Product Release to the EnvironmentUnmitigated STSBO

0.00

0.01

0.02

0.03

0.04

0.05

0 1 2 3 4

Time (days)

Fra

ctio

n r

ele

as

e (-

)

I

Cs

~1% at 4 days

Slide 27 of 26

Surry Station BlackoutsCompared to SST-1

Surry SOARCA Environmental Release Fractions

0.000

0.100

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0.800

0.900

1.000

0 1 2 3 4

Time (days)

Rel

ease

Fra

ctio

n (

-)

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

0 1 2 3 4

Time (days)

Rel

ease

Fra

ctio

n (

-)

LTSBO - I

LTSBO - Cs

STSBO - I

STSBO - Cs

SST-1 Iodine

SST-1 Cesium

Slide 28 of 26

Peach Bottom Long Term Station BlackoutCompared to SST-1

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0 8 16 24 32 40 48

time [hr]

Fra

ctio

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f In

itia

l Co

re In

ven

tory

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