state-of-the-art reactor consequence analyses (soarca) project accident analysis
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State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Accident Analysis. Presented at the USNRC 20 th Regulatory Information Conference Washington, DC March 11, 2008 Randall Gauntt, Sandia National Laboratories Charles Tinkler, USNRC. - PowerPoint PPT PresentationTRANSCRIPT
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
200
300
400
500
600
700
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
0.4
0.5
0.6
0.7
0.8
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
0.6
0.7
0.8
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
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
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
0.200
0.300
0.400
0.500
0.600
0.700
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
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1
0 8 16 24 32 40 48
time [hr]
Fra
ctio
n o
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