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{{#Wiki_filter:NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1) License Renewal Application     NRC Staff Answer to Motion for Summary Disposition of Contention 4B     ATTACHMENT 4B-B State-of-the-Art Reactor Consequence Analyses (SOARCA) Report
{{#Wiki_filter:NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1)
License Renewal Application NRC Staff Answer to Motion for Summary Disposition of Contention 4B ATTACHMENT 4B-B


Office of Nuclear Regulatory Research NUREG-1935 xiii  selection process captured the more important internally and externally initiated core damage scenarios.
NUREG-1935 State-of-the-Art Reactor Consequence Analyses (SOARCA) Report Office of Nuclear Regulatory Research
SOARCA's analyses were performed with two computer codes, MELCOR for accident progression and the MELCOR Accident Consequence Code System, Version 2 (MACCS2) for offsite consequences. The NRC staff's preparations for the analyses included extensive cooperation from the licensees of Peach Bottom and Surry to develop high
-fidelity plant systems models, define operator actions including the most recently developed mitigation actions, and develop models for simulation of site
-specific and scenario
-specific emergency planning and response. Moreover, in addition to input for model development, licensees provided information on accident scenarios from their PRAs. Through tabletop exercises of the selected scenarios with senior reactor operators, PRA analysts, and other licensee staff, licensees provided input on the timing and nature of the operator actions to mitigate the selected scenarios. The licensee input for each scenario was used to develop assumed timelines of operator actions and equipment configurations for implementing available mitigation measures which include mitigation measures beyond those routinely credited in current PRA models.
A human reliability analysis, commonly included in PRAs to represent the reliability of operator actions, was not performed for SOARCA, but instead tabletop exercises, plant walkdowns, simulator runs and other inputs from licensee staff were employed to ensure that operator actions and their timings were correctly modeled.


SOARCA modeled mitigation measures, including those in emergency operating procedure s (EOPs), severe accident management guidelines (SAMGs), and Title 10 to the Code of Federal Regulations (10 CFR) 50.54(hh).
selection process captured the more important internally and externally initiated core damage scenarios.
The 10 CFR 50.54(hh) mitigation measures refer to additional equipment and strategies required by the NRC following the terrorist attacks of September 11, 2001, to further improve each plant's capability to mitigate events involving a loss of large areas of the plant caused by fire and explosions. To assess the benefits of 10 CFR 50.54(hh) mitigation measures and to provide a basis for comparison to the past analyses of unmitigated severe accident scenarios, the SOARCA project also analyzed each scenario without 10 CFR 50.54 (hh) equipment and procedures.
SOARCAs analyses were performed with two computer codes, MELCOR for accident progression and the MELCOR Accident Consequence Code System, Version 2 (MACCS2) for offsite consequences. The NRC staffs preparations for the analyses included extensive cooperation from the licensees of Peach Bottom and Surry to develop high-fidelity plant systems models, define operator actions including the most recently developed mitigation actions, and develop models for simulation of site-specific and scenario-specific emergency planning and response. Moreover, in addition to input for model development, licensees provided information on accident scenarios from their PRAs. Through tabletop exercises of the selected scenarios with senior reactor operators, PRA analysts, and other licensee staff, licensees provided input on the timing and nature of the operator actions to mitigate the selected scenarios. The licensee input for each scenario was used to develop assumed timelines of operator actions and equipment configurations for implementing available mitigation measures which include mitigation measures beyond those routinely credited in current PRA models. A human reliability analysis, commonly included in PRAs to represent the reliability of operator actions, was not performed for SOARCA, but instead tabletop exercises, plant walkdowns, simulator runs and other inputs from licensee staff were employed to ensure that operator actions and their timings were correctly modeled.
The analysis that credits successful implementation of the 10 CFR 50.54 (hh) equipment and procedures in addition to actions directed by the EOPs and SAMGs is referred to as the mitigated case.
SOARCA modeled mitigation measures, including those in emergency operating procedures (EOPs), severe accident management guidelines (SAMGs), and Title 10 to the Code of Federal Regulations (10 CFR) 50.54(hh). The 10 CFR 50.54(hh) mitigation measures refer to additional equipment and strategies required by the NRC following the terrorist attacks of September 11, 2001, to further improve each plants capability to mitigate events involving a loss of large areas of the plant caused by fire and explosions. To assess the benefits of 10 CFR 50.54(hh) mitigation measures and to provide a basis for comparison to the past analyses of unmitigated severe accident scenarios, the SOARCA project also analyzed each scenario without 10 CFR 50.54 (hh) equipment and procedures. The analysis that credits successful implementation of the 10 CFR 50.54 (hh) equipment and procedures in addition to actions directed by the EOPs and SAMGs is referred to as the mitigated case. The analysis without 10 CFR 50.54(hh) equipment and procedures is referred to as the unmitigated case (SAMGs were considered but not implemented in the unmitigated case). The unmitigated case of the Surry ISLOCA is an exception to this general principle because it was necessary to assume that at least one of the EOP actions failed to occur for the scenario to lead to core damage. Chapter 3 of NUREG/CR-7110, Volume 1, SOARCA Peach Bottom Integrated Analysis and Volume 2, SOARCA Surry Integrated Analysis, details the specific equipment and operator actions credited for each scenario.
The analysis without 10 CFR 50.54(hh) equipment and procedures is referred to as the unmitigated case (SAMGs were considered but not implemented in the unmitigated case)
For the LTSBO scenarios for both Peach Bottom and Surry (the most likely severe accident scenario for each plant considered in SOARCA) analyzed assuming no mitigation, core damage begins in 9 to 16 hours, and reactor vessel failure begins at about 20 hours. Offsite radiological release due to containment failure begins at about 20 hours for Peach Bottom (BWR) and at 45 hours for Surry (PWR). The SOARCA analyses therefore show that time may be available for operators to take corrective action and get additional assistance from plant technical support centers even if initial efforts are assumed unsuccessful. For the most rapid events (i.e., the xiii
. The unmitigated case of the Surry ISLOCA is an exception to this general principl e because it was necessary to assume that at least one of the EOP actions failed to occur for the scenario to lead to core damage.
 
Chapter 3 of NUREG/CR-7110, Volume 1 , "SOARCA Peach Bottom Integrated Analysis" and Volume 2, "SOARCA Surry Integrated Analysis", details the specific equipment and operator actions credited for each scenario
B-70}}
.
For the LTSBO scenarios for both Peach Bottom and Surry (the most likely severe accident scenario for each plant considered in SOARCA) analyzed assuming no mitigation, core damage begins in 9 to 16 hours, and reactor vessel failure begins at about 20 hours. Offsite radiological release due to containment failure begins at about 20 hours for Peach Bottom (BWR) and at 45 hours for Surry (PWR). The SOARCA analyses therefore show that time may be available for operators to take corrective action and get additional assistance from plant technical support centers even if initial efforts are assumed unsuccessful. For the most rapid events (i.e., the B-70}}

Latest revision as of 15:59, 4 November 2019

NRC Staff Answer to Motion for Summary Disposition of Contention 4B - Attachment 4B-B
ML13196A430
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 07/15/2013
From:
Atomic Safety and Licensing Board Panel
To:
SECY RAS
References
50-443-LR, ASLBP 10-906-02-LR-BD01, RAS 24821
Download: ML13196A430 (4)
v • d • e


Text

NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1)

License Renewal Application NRC Staff Answer to Motion for Summary Disposition of Contention 4B ATTACHMENT 4B-B

NUREG-1935 State-of-the-Art Reactor Consequence Analyses (SOARCA) Report Office of Nuclear Regulatory Research

selection process captured the more important internally and externally initiated core damage scenarios.

SOARCAs analyses were performed with two computer codes, MELCOR for accident progression and the MELCOR Accident Consequence Code System, Version 2 (MACCS2) for offsite consequences. The NRC staffs preparations for the analyses included extensive cooperation from the licensees of Peach Bottom and Surry to develop high-fidelity plant systems models, define operator actions including the most recently developed mitigation actions, and develop models for simulation of site-specific and scenario-specific emergency planning and response. Moreover, in addition to input for model development, licensees provided information on accident scenarios from their PRAs. Through tabletop exercises of the selected scenarios with senior reactor operators, PRA analysts, and other licensee staff, licensees provided input on the timing and nature of the operator actions to mitigate the selected scenarios. The licensee input for each scenario was used to develop assumed timelines of operator actions and equipment configurations for implementing available mitigation measures which include mitigation measures beyond those routinely credited in current PRA models. A human reliability analysis, commonly included in PRAs to represent the reliability of operator actions, was not performed for SOARCA, but instead tabletop exercises, plant walkdowns, simulator runs and other inputs from licensee staff were employed to ensure that operator actions and their timings were correctly modeled.

SOARCA modeled mitigation measures, including those in emergency operating procedures (EOPs), severe accident management guidelines (SAMGs), and Title 10 to the Code of Federal Regulations (10 CFR) 50.54(hh). The 10 CFR 50.54(hh) mitigation measures refer to additional equipment and strategies required by the NRC following the terrorist attacks of September 11, 2001, to further improve each plants capability to mitigate events involving a loss of large areas of the plant caused by fire and explosions. To assess the benefits of 10 CFR 50.54(hh) mitigation measures and to provide a basis for comparison to the past analyses of unmitigated severe accident scenarios, the SOARCA project also analyzed each scenario without 10 CFR 50.54 (hh) equipment and procedures. The analysis that credits successful implementation of the 10 CFR 50.54 (hh) equipment and procedures in addition to actions directed by the EOPs and SAMGs is referred to as the mitigated case. The analysis without 10 CFR 50.54(hh) equipment and procedures is referred to as the unmitigated case (SAMGs were considered but not implemented in the unmitigated case). The unmitigated case of the Surry ISLOCA is an exception to this general principle because it was necessary to assume that at least one of the EOP actions failed to occur for the scenario to lead to core damage. Chapter 3 of NUREG/CR-7110, Volume 1, SOARCA Peach Bottom Integrated Analysis and Volume 2, SOARCA Surry Integrated Analysis, details the specific equipment and operator actions credited for each scenario.

For the LTSBO scenarios for both Peach Bottom and Surry (the most likely severe accident scenario for each plant considered in SOARCA) analyzed assuming no mitigation, core damage begins in 9 to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, and reactor vessel failure begins at about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Offsite radiological release due to containment failure begins at about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> for Peach Bottom (BWR) and at 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> for Surry (PWR). The SOARCA analyses therefore show that time may be available for operators to take corrective action and get additional assistance from plant technical support centers even if initial efforts are assumed unsuccessful. For the most rapid events (i.e., the xiii

B-70

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