POLICY ISSUE 

NOTATION VOTE 

July 5, 2012 SECY-12-0092

FOR: The Commissioners 

FROM: R. W. Borchardt 

Executive Director for Operations 

SUBJECT: STATE-OF-THE-ART REACTOR CONSEQUENCE ANALYSES -

RECOMMENDATION FOR LIMITED ADDITIONAL ANALYSIS 

PURPOSE: 

The purpose of this memorandum is to respond to Staff Requirements Memorandum (SRM) 

SRM-SECY-05-0233, “Plan for Developing State-of-the-Art Reactor Consequence Analyses,” 

dated April 14, 2006 (WITS 200600236/200600237), and SRM-COMSECY-06-0064, “State-ofthe-Art Reactor Consequence Analyses Communications Plan,” dated April 2, 2007 (WITS 

200700117/200800373). Specifically, the Commission asked the staff to provide a 

recommendation based on initial studies as to whether continuing the State-of-the-Art Reactor 

Consequence Analyses (SOARCA) project as originally described in SECY-05-0233, “Plan for 

Developing State-of-the-Art Reactor Consequence Analyses,” dated December 22, 2005, is 

necessary to achieve its objectives. The staff is seeking Commission approval to perform 

limited additional analyses to meet the objectives of SOARCA. 

SUMMARY: 

The staff has completed the SOARCA project for two pilot plants, the Peach Bottom Atomic 

Power Station (Peach Bottom), which is a boiling water reactor (BWR) with a Mark I 

containment and the Surry Power Station (Surry), which is a pressurized water reactor (PWR) 

with a large dry (subatmospheric) containment. In order to provide the Commission a 

recommendation on whether and how to continue the SOARCA project, the staff sought input 

from a number of sources. Based on its review, the staff recommends performing limited 

severe accident consequence analyses for a PWR plant with an ice condenser containment 

CONTACT: Patricia Santiago, RES/DSA Jonathan Barr, RES/DSA 

 301-251-7982 301-251-7538 

SECY NOTE: THIS SECY PAPER TO BE RELEASED TO THE PUBLIC IN 10 WORKING 

DAYS WITH THE EXCEPTION OF THE ENCLOSURE WHICH IS SENSITIVE INTERNAL 

INFORMATION. 
The Commissioners - 2 - 

because these analyses will provide additional knowledge of severe accident progression and 

consequences in a plant with a unique containment design. The results of this analysis could 

inform agency activities such as the implementation of the Near Term Task Force (NTTF) 

recommendations related to hydrogen control, reliable hardened vents, or filtered vents for 

reactor designs other than BWRs with Mark I and Mark II containments. The staff also 

recommends performing an uncertainty analysis (UA) for a PWR station blackout (SBO) 

scenario. The ongoing UA for a Peach Bottom scenario, expected to be completed in late 2012, 

is providing important insights regarding how uncertainties in the most influential parameters 

associated with severe accident progression and consequences in a BWR affect the results. A 

UA for a Surry SBO scenario would be expected to provide similar types of insights for PWR 

severe accident progression and this would inform the Site Level 3 Probabilistic Risk 

Assessment (PRA) project for the Vogtle Electric Generating Plant (Vogtle) Units 1 and 2, which 

are also PWRs. 

DISCUSSION: 

The staff provided the Commission a memorandum titled, “State-of-the-Art Reactor 

Consequence Analyses – Peach Bottom and Surry Results” (ML12132A213), dated June 22, 

2012, which enclosed the completed SOARCA analyses of two pilot plants, Peach Bottom and 

Surry. The objectives of the SOARCA project as described in SECY-05-0233 are the following: 

• Using a methodology based on state-of-the-art analytical tools, determine best estimates 

of the radiological dose consequences (including early and latent fatalities and land 

contamination) for each U.S. operating reactor site and present those results using risk 

communication techniques; and 

• Achieve informed public understanding of the following factors: 

– Extent and value of defense-in-depth features of plant design and operation, 

including mitigative strategies that are employed to reduce risk 

– Most significant influential assumptions. 

The staff believes that the analyses of the important scenarios for the two pilot plants have 

generally met the objectives stated above. The NUREG reports (NUREG-1935, “State-of-theArt Reactor Consequence Analyses Report,” NUREG/CR-7110 Volume 1, “Peach Bottom 

Integrated Analysis,” and Volume 2, “Surry Integrated Analysis,” and NUREG/BR-0359, 

“Modeling Potential Reactor Accident Consequences”) that staff recently completed on the 

SOARCA project document the following: (1) the state-of-the-art analytical models and 

methods used in the SOARCA project, (2) the best estimate radiological dose consequences, 

including early and latent fatality risks, for important scenarios for the Peach Bottom and Surry 

plants, (3) the results presented using risk communication techniques, for example through a 

public information brochure and public meetings, and (4) the most influential assumptions such 

as failure criteria for BWR safety relief valves. The UA of a Peach Bottom SBO scenario 

currently in progress will add to staff and public understanding of the most significant influential 

assumptions in the SOARCA project. The Peach Bottom UA will be completed in late 2012, as 

well as the MELCOR and MELCOR Accident Consequence Code System, Version 2 (MACCS2) 

Best Practices NUREG/CRs. 
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Regarding public understanding of defense-in-depth features including mitigative strategies, 

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) equipment and strategies required following the terrorist 

attacks of September 11, 2001. 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. The unmitigated case of the 

Surry interfacing systems loss-of-coolant accident 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 and radioactive release. Chapter 3 of NUREG/CR-7110, 

Volumes 1 and 2 details the specific equipment and operator actions credited for each scenario. 

In addition, the SOARCA public information brochure (NUREG/BR-0359) includes a discussion 

of mitigation and the results of the mitigated and unmitigated scenarios. 

SOARCA analyses indicate that successful implementation of existing mitigation measures can 

prevent reactor core damage or delay or reduce offsite releases of radioactive material. All 

SOARCA scenarios, even when unmitigated, progress more slowly and release much less 

radioactive material than in earlier studies (e.g., NUREG/CR-2239, “Technical Guidance for 

Siting Criteria Development,” commonly referred to as the 1982 Siting Study). The SOARCA 

analyses show that emergency response programs implemented as planned and practiced, 

would reduce the scenario-specific risk of health consequences among the public during a 

severe reactor accident at the plants evaluated. As a result, the calculated risks of public health 

consequences from severe accidents modeled in SOARCA are very small. 

The SOARCA project’s scope initially included analysis of all operating reactors; however 

SRM-COMSECY-06-0064 reduced the scope to a subset of eight plants with first assessments 

of one BWR and one PWR. The staff has concluded it is not necessary to analyze all plant sites 

or even all plant types to achieve the objectives of the SOARCA project. The analyses for 

Peach Bottom and Surry updated the understanding of severe accident progression, mitigation, 

and consequences for two major plant designs—the BWR with a Mark I containment and the 

PWR with a large dry (subatmospheric) containment. These two designs encompass 80 

percent of the designs of the current operating fleet (83 of 104 operating reactors). 

In order to provide the Commission a recommendation on whether and how to continue the 

SOARCA project, the staff sought input from a number of sources. These sources include the 

historical record of staff interactions with the Commission, the external SOARCA peer review 

committee, the internal SOARCA steering committee, the Advisory Committee on Reactor 

Safeguards (ACRS), and members of the public. Further, staff also used Fukushima lessons 

learned to identify potential open issues that would necessitate additional research. 

Based on this review, the staff identified the following items for potential further work or to 

address open issues raised during the pilot plant studies. To avoid duplication and ensure 

efficient use of agency resources, the staff evaluated whether the items require NRC action, are 

being addressed by other agency activities, or whether follow-on SOARCA-related research 

would be the best alternative to address the item. For each of the items except the last two, the 
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staff determined that either other agency activities (noted in parentheses) are being conducted 

or planned to address the issue, the item is no longer needed, or appropriate information is not 

currently available to analyze the item. The following paragraphs provide details on the staff’s 

disposition of these items. 

• Spent fuel pool risk (Fukushima Tier 3 spent fuel transfer recommendation, Site Level 3 

PRA) 

• Multi-unit and site risk (Site Level 3 PRA) 

• Low power, shutdown, and other modes (Site Level 3 PRA) 

• Human reliability analysis (Site Level 3 PRA) 

• Economic consequences and land contamination (Commission paper under 

development) 

• Additional comparisons to Fukushima (DOE/NRC Forensic Analysis) 

• Large seismic event and soil liquefaction (Site Level 3 PRA and Seismic research plan) 

• Development of an integrated, predictive computer-based tool (no longer needed) 

• Cancer incidence rates (requires updated EPA risk coefficients) 

• Other plant and containment types – e.g., Ice condenser, Mark II, III, CE, B&W 

• UA for a Surry scenario 

Many stakeholders hold the view that the spent fuel pools, multiple units, and operating modes 

other than full power should be analyzed. Stakeholders also feel that a human reliability 

analysis for mitigation measures would be informative. SOARCA was not intended to be the 

equivalent of a full-scope level 3 PRA. Rather, it provides a state-of-the-art body of knowledge, 

models, and methodologies about severe accident consequences in two representative reactor 

designs that can be applied as needed to agency activities. The ongoing Site Level 3 PRA of 

the Vogle Units 1 & 2, which will analyze spent fuel pool risk, multi-unit and site risk, all 

operating modes, and human reliability, will further add to this body of knowledge, models, and 

methodologies. The Site Level 3 PRA plans to use the updated modeling and methods that 

were developed for SOARCA and is scheduled for completion in 2016. Additionally, SOARCA 

models and methods will inform the staff’s Tier 3 lessons-learned activity related to the potential 

transfer of spent fuel from pools to casks. 

As described in SECY-05-0233, the SOARCA project’s original scope included calculating 

offsite consequences of severe accidents in terms of (1) health effects and (2) land 

contamination. However, in SRM-COMPBL-08-002/COMGBJ-08-0003, “Economic 

Consequence Model” dated September 10, 2008, the Commission directed the staff to complete 

the SOARCA project to assess offsite health consequences and to not delay the project in order 

to include an assessment of economic consequences and land contamination. 

The Fukushima accident occurred when SOARCA was nearing completion. Because the Peach 

Bottom plant studied in the SOARCA project has a similar design to the Fukushima plants, 

stakeholders posed many questions on how SOARCA Peach Bottom calculations compared 

with actual Fukushima outcomes. The SOARCA report (NUREG-1935) includes a qualitative 

discussion comparing the Peach Bottom analysis to Fukushima in five areas—(1) operation of 

the reactor core isolation cooling system, (2) hydrogen release and combustion, (3) 48-hour 

truncation of releases in SOARCA, (4) multi-unit risk, and (5) spent fuel pool risk. Additionally, 

Sandia National Laboratories is conducting a joint NRC and U.S. Department of Energy (DOE) 

Fukushima Forensic Analysis targeted for completion later this year. This analysis should 

provide additional insights into the Fukushima accident. It is also expected that international 
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cooperative efforts sponsored by the Organization for Economic Cooperation and 

Development’s Nuclear Energy Agency and the International Atomic Energy Agency over the 

next few years will provide ongoing analysis of Fukushima with NRC and other countries’ codes 

as information becomes available. 

The SOARCA study analyzed SBOs assumed to be initiated by low probability seismic events. 

SOARCA excluded more extreme seismic events of even lower probability which could 

conceivably cause immediate containment failure followed by core damage. Seismic fragility 

quantification for these extreme and rare seismic events, in particular quantification of the size 

of a hole or amount of leakage, is currently subject to considerable uncertainty. More research 

is needed before undertaking a realistic, best-estimate analysis of such rare events. Soil 

liquefaction analysis will be included in the Site Level 3 PRA project. 

In SECY-05-0233, the staff proposed developing an integrated, predictive, computer-based tool 

to assist decisionmaking in the event of a severe reactor accident. Staff envisioned that this tool 

could model and predict accident progression faster than real time and thereby aid the NRC’s 

incident response center in the event of a severe reactor accident. Resources needed to 

develop this tool were redirected to higher priority work including completion of the SOARCA 

pilot plant analyses for Peach Bottom and Surry. Increases in computing capability allow 

existing codes to perform analyses much faster than in the past. As a result of the slow 

progression of the Fukushima accident and our demonstrated ability to use existing tools 

including MELCOR, MACCS2, and the Radiological Assessment System for Consequence 

Analysis (RASCAL) code during the Fukushima response, the staff concludes that the 

development of another predictive computer-based tool is not necessary. However, staff is 

evaluating enhancements to existing tools based on the Fukushima experience, and they would 

be deployed as necessary in the agency’s incident response program. 

Some stakeholders noted that a state-of-the-art report should report cancer incidence as well 

as, or instead of, cancer mortality since that is state-of-the-art in epidemiology. The SOARCA 

project reported offsite consequences in terms of individual average latent cancer fatality risk to 

provide appropriate context of the risk in relation to other risks and to enable comparison to the 

NRC Safety Goal. SOARCA used latent cancer expression coefficients for the United States 

population based on Biological Effects of Ionizing Radiation (BEIR) V risk projection models, as 

detailed in the U.S. Environmental Protection Agency’s (EPA’s) publication “Estimating 

Radiogenic Cancer Risks” (EPA 402-R-93-076, 1994) and implemented in the EPA’s Federal 

Guidance Report 13, “Cancer Risk Coefficients for Environmental Exposure to Radionuclides” 

(FGR-13). The BEIR V report used cancer mortality as a metric because, at that time, most 

epidemiological studies were based on cancer mortality, not cancer incidence. The EPA has 

not yet incorporated cancer incidence data from the BEIR VII report into a revision of FGR-13, 

so there are no new updated cancer risk coefficients currently available. 

SECY-05-0233 envisioned applying the SOARCA methodology to all operating reactors. 

Subsequently, SRM-COMSECY-06-0064 directed the staff to analyze just eight reactors 

representing the different plant/containment types, starting with one BWR and one PWR. Three 

plants had volunteered for the project—Peach Bottom, Surry, and the Sequoyah Nuclear Plant 

(Sequoyah). The SOARCA team started analysis of these three plants but then stopped its 

analysis of Sequoyah in order to focus resources on the Peach Bottom and Surry analyses. 

The staff evaluated the SOARCA pilot plant results and the improved models and methods for 
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severe accident and consequence analysis for each of the remaining six plant/containment 

designs to determine whether further analyses are necessary to meet the project’s objectives. 

Due to the similarities of the other six designs to the designs of Peach Bottom and Surry, staff 

believes that these analyses would provide limited additional insight and thus would not be cost 

effective, as discussed below. However, a severe accident consequence analysis of SBOs at a 

PWR with an ice condenser containment would inform ongoing agency activities and be cost 

effective since staff has already initiated the analysis of Sequoyah. 

The ice condenser containment is unique relative to other containment types including those 

studied in the SOARCA project (BWR Mark I at Peach Bottom and large dry at Surry) and that 

to be studied in the Site Level 3 PRA project (large dry at Vogtle 1 & 2). It has aspects worth 

exploring regarding buildup of hydrogen during an SBO since it has the lowest design pressure 

among containments in the United States and relies on hydrogen igniters. Sequoyah provided 

plant information for the original SOARCA analyses, and staff would continue the analysis with 

some resource savings. The NRC currently has a partially developed MELCOR model of 

Sequoyah. Further analysis would be limited to a seismically induced SBO event since the 

SOARCA project, Fukushima, and other studies have shown SBOs to be the most important 

scenarios. 

The BWR with a Mark III containment design has similar aspects regarding hydrogen during an 

SBO as the PWR with an ice condenser containment. However, staff does not recommend 

analyzing the BWR Mark III. The BWR Mark III has a larger containment volume and greater 

containment design pressure than the ice condenser containment, so the ice condenser 

consequence analysis is considered more likely to reveal a more extreme accident progression. 

Also, the BWR Mark III design is less prevalent than the PWR ice condenser design (4 BWR 

Mark III units vs. 9 PWR ice condenser units currently operating1

). 

The staff does not recommend conducting severe accident consequences for a BWR Mark II 

design either. The BWR Mark II design is similar in concept to the Mark I design but with a 

simplified suppression pool design and a more unified containment structure. Containment heat 

removal systems (sprays and suppression pool cooling) and nitrogen inerting strategies are the 

same as for the Mark I containments. Containment venting can also be performed in a similar 

fashion to the Mark I containments. Due to these similarities, the staff believes that its analysis 

of Peach Bottom bounds the accident progression and consequences expected for an SBO at a 

BWR Mark II plant and therefore an analysis of Mark II plant would not be cost effective. 

The staff also believes that severe accident consequence analyses of other PWR designs with 

large dry containments beyond Surry would not be cost effective. Although Surry has a 

subatmospheric large dry containment, the difference between its containment and a standard 

large dry containment is not expected to impact severe accident progression and offsite 

consequences. The major difference is that the containment is maintained at a negative 

pressure with respect to the outside atmosphere. This negative pressure provides some 

additional margin for response to design basis accidents but is insignificant for severe 

 

1 Currently operating PWR plants with an ice condenser containment: Catawba Nuclear Station Units 1 and 2, 

Donald C. Cook Nuclear Power Plant Units 1 and 2, McGuire Nuclear Station Units 1 and 2, Sequoyah Nuclear Plant 

Units 1 and 2, and Watts Bar Nuclear Plant Unit 1 

Currently operating BWR Mark III plants: Clinton Power Station Unit 1, Grand Gulf Nuclear Station Unit 1, Perry 

Nuclear Power Plant Unit 1, and River Bend Station Unit 1
The Commissioners - 7 - 

accidents. The staff also recommends against analyzing a Babcock and Wilcox (B&W) 

designed PWR with a large dry containment or a Combustion Engineering (CE) designed PWR 

with a large dry containment. While the B&W and CE nuclear steam supply system (NSSS) 

designs are different than the Westinghouse NSSS design in the Surry plant, the differences are 

mostly in the design and size of the steam generators. A severe accident consequence 

analysis of B&W and CE designs could reveal some limited insights regarding accident 

progression, but the staff believes these would not be cost effective. 

Beyond the goal of more completely meeting the SOARCA project’s original objectives, the 

consequence analysis of an ice condenser plant would provide insights and results to inform 

agency activities such as responses to NTTF recommendations 5.1 (reliable hardened vents for 

Mark I and Mark II containments) and 5.2 (reliable hardened vents for containment designs 

other than Mark I and II), 6 (hydrogen control and mitigation inside containment or in other 

buildings), and 12.2 (enhance training on severe accidents). 

The SOARCA project included within its scope a UA for a severe accident scenario at one of the 

two pilot plants studied. The unmitigated long-term SBO at Peach Bottom was selected 

because its release timing and consequences are characteristic of the majority of unmitigated 

scenarios analyzed in SOARCA. Since this was a first-of-a-kind study, significant resources 

were required to develop the methodology to conduct the UA. Based on progress to date, the 

Peach Bottom UA has revealed many important insights about accident progression in a BWR 

with a Mark I containment. For example, the Peach Bottom UA has shown that the failure 

criteria for safety relief valves (SRV) is critical to determine whether the reactor coolant system 

fails by main steam line creep rupture or by SRV failure. A UA for a Surry scenario would be 

expected to reveal similar insights for PWR severe accident progression and these 

insights would directly support the Site Level 3 PRA project. For example, the Surry UA could 

provide data regarding reactor coolant system and containment failure modes and subsequent 

releases that could support the choices for split fractions and release fractions in the Level 2 

component of the PRA. 

Pending available resources, the staff considers the ice condenser plant analysis higher priority 

than the Surry UA. This is because the ice condenser analysis will provide additional 

knowledge of severe accident progression in another containment design. 

In the May 15, 2012 letter (ML12135A385) from J. S. Armijo to Chairman Jaczko, the ACRS 

commended the staff’s analyses of Peach Bottom and Surry and endorsed a follow-on study to 

analyze severe accident progression in a plant with an ice condenser containment, noting that it 

should be a lower priority than completion of the Site Level 3 PRA project. Staff would 

coordinate the Sequoyah analysis and Surry UA with the Site Level 3 PRA project as well as 

other activities supporting the Fukushima lessons learned such that schedules for all activities 

are not adversely impacted. The ACRS also recommended staff perform a UA for Surry. Staff 

anticipates performing a UA for selected Surry parameters that would inform the Site Level 3 

PRA as the Vogtle MELCOR model is being completed. These analyses would also add to the 

body of knowledge on severe accident progression and consequences developed in SOARCA. 
The Commissioners - 8 - 

RECOMMENDATION: 

The staff recommends that the Commission approve limited continuation of the SOARCA 

project. Based on the review of open items related to severe accident research, the staff 

recommends the following two analyses: 

 

(1) Severe accident consequence analysis of an SBO at a PWR with an ice condenser 

containment (Sequoyah) based on the insights from the Peach Bottom and Surry 

analyses. 

(2) UA for a severe accident scenario at Surry. 

RESOURCES: 

Resources proposed to complete the activities discussed above are described in the non-public 

Enclosure. 

COORDINATION: 

The Office of the Chief Financial Officer has reviewed this paper for resource implications and 

has no objections. The Office of the General Counsel has reviewed this paper and has no legal 

objection. 

 /RA/ 

R. W. Borchardt 

 Executive Director 

 for Operations 

Enclosure: 

Projected Resource Needs for SOARCA 

 Recommendations for Limited Additional 

 Analyses
The Commissioners - 8 - 

RECOMMENDATION: 

The staff recommends that the Commission approve limited continuation of the SOARCA 

project. Based on the review of open items related to severe accident research, the staff 

recommends the following two analyses: 

 

(1) Severe accident consequence analysis of an SBO at a PWR with an ice condenser 

containment (Sequoyah) based on the insights from the Peach Bottom and Surry 

analyses. 

(2) UA for a severe accident scenario at Surry. 

RESOURCES: 

Resources proposed to complete the activities discussed above are described in the non-public 

Enclosure. 

COORDINATION: 

The Office of the Chief Financial Officer has reviewed this paper for resource implications and 

has no objections. The Office of the General Counsel has reviewed this paper and has no legal 

objection. 

/RA/ 

R. W. Borchardt 

 Executive Director 

 for Operations 

Enclosure: 

Projected Resource Needs for SOARCA 

 Recommendations for Limited Additional 

 Analyses 

 WITS 200700117/EDATS: SECY-2010-0267 

ADAMS Accession No.: ML12132A274 (pkg) WITS 200800373/EDATS: SECY-2012-0275 

OFFICE RES/DSA/AAB BC: RES/DSA Tech Editor D: RES/DSA D: RES/DRA 

NAME JBarr PSantiago QTE KGibson RCorreia 

DATE 05/11/12 05/11/12 05/14/12 05/15/12 05/23/12 

OFFICE D: RES/DE Acting D: NRO D: NRR D: NSIR A: RIII 

NAME MCase GHolahan 

(C. Ader for via 

email) 

ELeeds 

(J. Giitter for via 

email) 

JWiggins 

(via email) 

CCasto 

(C. Pederson for 

 via email) 

DATE 05/23/12 06/6/12 06/6/12 06/6/12 06/6/12 

OFFICE D: OPA OGC OCFO D:RES EDO 

NAME EBrenner 

(S. Burnell for via 

email) 

MZobler 

(B. Harris for via 

email) 

JDyer 

(K. Kline for via 

email) 

BSheron RBorchardt 

DATE 06/6/12 06/12/12 06/18/12 06/22/12 7/5 /12 

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