ML20100J883

Research Information Letter 20-03, Benefits and Uses of the State-of-the-Art Reactor Consequence Analyses (Soarca) Project

ML20100J883
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Issue date:	03/31/2020
From:	Christina Leggett NRC/RES/DSA
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RIL 20-03
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RIL20-03 BENEFITS AND USES OF THE STATE-OF-THE-ART REACTOR CONSEQUENCE ANALYSES (SOARCA)

PROJECT Date Published: March 2020 Prepared by:

C. Leggett Reactor Systems Engineer Division of Systems Analysis Office of Nuclear Regulatory Research Research Information Letter Office of Nuclear Regulatory Research

Disclaimer Legally binding regulatory requirements are stated only in laws, NRC regulations, licenses, including technical specifications, or orders, not in Research Information Letters (RILs). A RIL is not regulatory guidance, although NRCs regulatory offices may consider the information in a RIL to determine whether any regulatory actions are warranted.

EXECUTIVE

SUMMARY

The State-of-the-Art Reactor Consequence Analyses (SOARCA) project was a major research study undertaken over the last decade by the U.S. Nuclear Regulatory Commission (NRC) and its contractor, Sandia National Laboratories. The project objectives were to develop an updated body of knowledge on the realistic outcomes of severe reactor accidents and update the assessment of severe accidents in previous NRC studies that were believed to be conservative.

The projects scope evolved over time and ultimately includes detailed accident progression and source term calculations using the MELCOR computer code and detailed accident consequence calculations using the MELCOR Accident Consequence Code System for three pilot plants:

Peach Bottom Atomic Power Station, a boiling-water reactor with a Mark I containment in Pennsylvania; Surry Power Station, a pressurized-water reactor with a large dry subatmospheric containment in Virginia; and Sequoyah Nuclear Plant, a pressurized-water reactor with an ice condenser containment in Tennessee. For each plant, the project team conducted a set of deterministic best estimate calculations and a detailed uncertainty analysis for one accident scenario. These calculations were completed in 2017. The results of these analyses consistently predict essentially zero individual early fatality risk for the modeled scenarios, very low long-term cancer fatality risks, and smaller radiological releases than those predicted from previous studies. The project has involved extensive communication through technical documentation, presentations, and public meetings with diverse stakeholders.

In addition to satisfying the project objectives, the SOARCA project has (1) developed staff expertise in a variety of important technical areas, including accident progression and source term analysis, offsite consequence analysis, parametric uncertainty analysis, and risk communication; (2) identified improvements in NRC analytical tools, such as computer codes and associated severe accident analysis methodologies; (3) provided readily available, detailed site- and plant-specific computer code models that could be used for additional analyses; and (4) been used to support risk-informed decisions that in turn supported safe and economical operating decisions. The improvement of tools, methodologies, and technical expertise has enhanced the NRCs ability to efficiently and effectively carry out its mission to protect public health and safety and the environment.

The projects results, insights, computer code models, and modeling best practices have supported NRC rulemaking, licensing, and oversight efforts and facilitated international cooperation and knowledge management. For example, computer code models and modeling best practices from SOARCA enabled the NRC to perform timely calculations to support its technical basis for issues identified as a result of the accident at the Fukushima Daiichi nuclear power plant in Japan. These include filtered containment venting for boiling-water reactors with Mark I and Mark II containments (see NUREG-2206, Technical Basis for the Containment Protection and Release Reduction Rulemaking for Boiling Water Reactors with Mark I and Mark II Containments, issued March 2018) and expedited transfer of spent fuel from pools to dry cask storage (see NUREG-2161, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, issued September 2014). SOARCA results and insights were used during the accident at Fukushima to support rapid, time-sensitive emergency response. The outcomes of NRC and external stakeholders uses of SOARCA include a substantial number of risk-informed decisions that have enhanced safety and security, while also supporting operational flexibilities, in the U.S.

and abroad. This research information letter (RIL) describes these and many other applications of SOARCA project insights, models, and methodologies.

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The SOARCA project has produced nine NUREG-series publications and has been used or cited in over 325 publications in the open literature, including technical reports, conference papers, journal articles, and dissertations. These publications cover a broad range of research areas, including but not limited to accident-tolerant fuel, reactor safety (including advanced designs), societal risk, and spent fuel storage and transportation, demonstrating the diverse areas in which researchers have referenced or used aspects of the SOARCA project.

The purpose of this RIL is to formally document the numerous benefits and uses of the project beyond its original objectives as well as its uses by the NRC, reactor licensees and applicants, domestic and international regulatory and research organizations, academia, and other stakeholders. This RIL focuses on the benefits and uses of SOARCA to date, and although there are many potential additional future benefits, the RIL does not speculate on them. It summarizes the SOARCA project, including motivation, approach, and results, followed by a high-level description of how SOARCA information has supported various projects and analyses related to nuclear power safety. The appendix lists publications from diverse research areas that have used or cited the SOARCA project.

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TABLE OF CONTENTS EXECUTIVE

SUMMARY

...........................................................................................................III 1 OVERVIEW OF THE SOARCA PROJECT ....................................................................... 1 2 BENEFITS AND USES OF THE SOARCA PROJECT ...................................................... 5 2.1 Regulatory Support for NRC Post-Fukushima Activities........................................... 6 Evaluation of Filtered Containment Venting and Containment Protection and Release Reduction Strategies for Boiling-Water Reactors with Mark I and Mark II Containments ............................................................................ 6 Evaluation of Tier 3 Issues Related to Containment Venting for Other than Mark I and Mark II Containments and to Hydrogen Control and Mitigation ..................................................................................................... 8 Evaluation of Expedited Transfer of Spent Fuel to Dry Cask Storage........... 8 2.2 Licensing and Design Certification Application Reviews .......................................... 9 NuScale Application and the NRCs Review ................................................ 9 APR1400 Design Certification Review and SOARCA-Like Analysis ............10 Contentions on Severe Accident Mitigation Alternatives during Indian Point Operating License Renewal Proceedings...........................................10 AREVA European Pressurized Reactor Application ....................................10 2.3 Domestic and International Cooperation in Response to the Fukushima Daiichi Accident .................................................................................................................11 Support for Federal Government Response to Fukushima ..........................11 Fukushima Incident Response ....................................................................11 Fukushima Forensic Analysis ......................................................................11 Fukushima Uncertainty Analyses ................................................................12 Benchmark Study of the Accident at Fukushima .........................................12 Benchmarking of Fast-Running Software Tools to Inform Offsite Decisionmaking...........................................................................................12 2.4 Risk-Informing NRC Programs and Projects ...........................................................13 Probabilistic Risk Assessment Technical Development...............................13 Site Level 3 PRA Project .............................................................................13 Emergency Preparedness Significance Quantification Process ..................14 Seismic Probabilistic Risk Assessment Relief Request ...............................14 Identification of Missing Emergency Action Level ........................................14 NEI Use of SOARCA in Discussions on Margins.........................................15 2.5 Enhancement of Existing Computer Models and Methodologies for Technical Analyses .................................................................................................................15 v

Analysis of Cyber-Attack Vectors on Nuclear Power Plants ........................15 Atucha II Level 2 and 3 Probabilistic Risk Assessment Technical Review ...15 Boiling-Water Reactor Owners Group Severe Accident Guidance ..............15 Integrated System Response Modeling .......................................................16 MELCOR Integration with the Emergency Response Data System .............16 Enhancement of the RASCAL Code for Incident Response ........................16 Modern Radiological Source Terms for MACCS Model Testing and Benchmarking .............................................................................................16 2.6 Knowledge Management ........................................................................................17 Cooperative Severe Accident Research Program .......................................17 NRC-Sponsored Training ............................................................................17 Boiling-Water Reactor Owners Group and Nuclear Energy Expert Group Training .......................................................................................................17 Sandia National Laboratories Training to Korea Hydro and Nuclear Power .........................................................................................................18 3 CONCLUSIONS ..............................................................................................................18 4 REFERENCES ................................................................................................................20 APPENDIX A LIST OF TECHNICAL LITERATURE CITING THE SOARCA PROJECT .. A-1 vi

1 OVERVIEW OF THE SOARCA PROJECT The U.S. Nuclear Regulatory Commission (NRC) initiated the State-of-the-Art Reactor Consequence Analyses (SOARCA) project to provide a state-of-the-art, more realistic evaluation of severe accident progression, radionuclide release, and offsite consequences for risk-significant severe reactor accident scenarios. The analyses leveraged insights from several decades of research on severe accident phenomenology and radiation health effects, information that was captured in two modern codes: MELCOR, an integrated severe accident progression code, and the MELCOR Accident Consequence Code System (MACCS), an accident consequence analysis code. One of the objectives was to update the quantification of offsite consequences found in earlier NRC publications, particularly NUREG/CR-2239, Technical Guidance for Siting Criteria Development, issued December 1982, as well as WASH-1400, Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, issued October 1975, and NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, issued October 1990. The project team did so by incorporating (1) significant plant improvements and changes that were not reflected in earlier assessments, including security-related enhancements issued in Title 10 of the Code of Federal Regulations (10 CFR) 50.54(hh)(2), 1 in the wake of the terrorist attacks of September 11, 2001, (2) emergency response, (3) reactor power uprates, and (4) higher core burnup. The MELCOR and MACCS models used the most up-to-date site- and plant-specific information. An additional objective of the SOARCA project was to enable the NRC to more effectively communicate severe accident-related aspects of nuclear safety to diverse stakeholders, including the public; Federal, State, and local authorities; and nuclear power plant licensees.

The SOARCA project analyzed severe accidents for Peach Bottom Atomic Power Station (Pennsylvania), which is a U.S. boiling-water reactor (BWR) with a Mark I containment, and Surry Power Station (Virginia), a U.S. pressurized-water reactor (PWR) with a large dry (subatmospheric) containment. The three historical studies mentioned above also evaluated these reactors. The project team analyzed two groups of reactor accident scenarios: (1) long-term station blackouts (LTSBOs) and short-term station blackouts (STSBOs) for both Peach Bottom and Surry, and (2) containment bypass scenarios, including thermally induced steam generator tube rupture (SGTR) and interfacing systems loss-of-coolant accident, for Surry only.

The project evaluated all scenarios, with and without the successful implementation of 10 CFR 50.54(hh)(2) mitigation equipment and procedures.

The process used to perform these evaluations included five steps:

(1) Select accident scenarios to model. The project used core damage frequencies (CDFs) to select accident scenarios. The SOARCA project analysts selected accident scenarios with a CDF higher than 10-6 per reactor year to allow them to analyze the most likely, yet very remotely possible, severe accident scenarios. These include the station blackout (SBO) scenarios described above. They also selected some lower probability accident scenarios (e.g., containment bypass scenarios) because of their potential to result in higher consequences. These accident scenarios used a lower screening criterion of 10-7 per-reactor-year.

1 SOARCA did not evaluate FLEX, since the FLEX strategies were not yet formulated at the time of the Peach Bottom and Surry studies and were still under development at the start of the study for Sequoyah Nuclear Plant.

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(2) Model accident progression and mitigation measures. The team used MELCOR to analyze accident progression, plant response, and mitigation measures for each of the scenarios described above.

(3) Model offsite release of radioactive material. The MACCS code used site-specific weather conditions to model atmospheric transport and dispersion of released radionuclides and site-specific population and land use data to model radiation exposure to the population.

(4) Model emergency response. In conjunction with step 3, the team modeled the evacuation of the public using site-specific emergency plans and evacuation time estimate studies.

(5) Model health effects. The team used MACCS to calculate radiation exposure to the population. It subsequently used the code to determine the resulting early and latent cancer fatality risks to the public. Analysts used multiple dose-response models to calculate individual latent cancer fatality risks.

The NRC completed the Peach Bottom and Surry SOARCA studies in 2012 and documented them in NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA), issued November 2012; NUREG/CR-7110, State-of-the-Art Reactor Consequence Analyses Project:

Volume 1: Peach Bottom Integrated Analysis, Revision 1, issued May 2013; and State-of-the-Art Reactor Consequence Analyses Project: Volume 2: Surry Integrated Analysis, Revision 1, issued August 2013. The NRCs Advisory Committee on Reactor Safeguards reviewed the methodology, assumptions, and results published in these studies, which were peer reviewed by an independent panel of external scientific and technical experts in the fields covered by the analysis. In addition, the staff received feedback from the public when it released draft NUREG-1935 for public comment. The staff addressed these comments before the final publication of NUREG-1935. To facilitate dissemination of severe accident consequence results to the general public, the NRC issued NUREG/BR-0359, Modeling Potential Reactor Accident Consequences, Revision 1, in December 2012.

The Peach Bottom and Surry SOARCA studies (2012 publications) had the following summary results:

• Radiological releases are considerably smaller than those reported in NUREG/CR-2239 for its siting source term 1 case 2, which led to the highest consequences in that study.

• Successful implementation of existing mitigation measures can prevent core damage or delay or reduce offsite releases of radionuclides.

• The individual early fatality risk for the modeled scenarios is essentially zero for both sites.

2 One of five source terms evaluated in NUREG/CR-2239, siting source term 1 is the source term resulting from a postulated severe accident scenario in which there is severe core damage, loss of all installed safety features, and a severe breach of containment.

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• The calculated individual long-term cancer fatality risks for the accident scenarios analyzed are millions of times lower than the general U.S. cancer fatality risk from all causes.

Before the conclusion of the Peach Bottom and Surry evaluations, the staff began an uncertainty analysis (UA) of the SOARCA unmitigated LTSBO severe accident scenario for Peach Bottom with the following objectives:

• Assess the overall sensitivity of SOARCA results to uncertainties in inputs.

• Identify the input parameters that most strongly influence releases and consequences.

• Demonstrate a UA methodology that could be used for subsequent studies.

This analysis used the same SOARCA model and software that were used for the deterministic 3 analyses documented in NUREG-1935 and NUREG/CR-7110, but it varied a set of key uncertain MELCOR and MACCS input parameters. The specific parameters chosen captured important influences on potential releases of radioactive materials and on offsite consequences.

Importantly, the results for the Peach Bottom UA (NUREG/CR-7155, State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Long-Term Station Blackout of the Peach Bottom Atomic Power Station, issued May 2016) corroborated the conclusions obtained from the Peach Bottom deterministic SOARCA study.

After completing the Peach Bottom and Surry SOARCA studies, the staff recommended to the Commission in SECY-12-0092, State-of-the-Art Reactor Consequence Analyses Recommendation for Limited Additional Analysis, dated July 5, 2012, that it perform a UA for Surry and a consequence analysis for Sequoyah focused on issues unique to the ice condenser containment design and limited to SBO scenarios. In Staff Requirements Memorandum (SRM)-SECY-12-0092, State-of-the-Art Reactor Consequence Analyses (SOARCA)

Recommendation for Limited Additional Analysis, dated December 6, 2012, the Commission approved the staffs recommendation and stated that the analyses should complement and support the ongoing Site Level 3 probabilistic risk assessment (PRA) project (referred to hereafter as the Site Level 3 PRA project; see Section 2.4.2) and regulatory activities following the accident at the Fukushima Daiichi nuclear power plant in Japan in March 2011. The staff performed a Surry UA of an unmitigated STSBO accident using the same approach as for the Peach Bottom UA. Notably, the UA results corroborated the results obtained from the Peach Bottom and Surry SOARCA studies. The NRC summarized the models, results, and insights in the draft report 4, State-of-the-Art Reactor Consequences Analyses Project: Uncertainty Analysis of the Unmitigated Short-Term Station Blackout of the Surry Power Station, issued August 2015 (ADAMS Accession No. ML15224A001).

As noted above, the Sequoyah deterministic (and uncertainty) analyses differed from the Peach Bottom and Surry analyses in that it focused only on LTSBO and STSBO accidents, specifically 3 The term deterministic is commonly used to refer to an analysis in which the model form is fixed, such that the same set of initial and boundary conditions will always lead to the same output. Even though SOARCA used selected accident sequences based on plant probabilistic risk analyses, the accident progression and consequences were not modeled probabilistically, with the exception that a range of potential weather conditions is sampled even in deterministic MELCOR Accident Consequence Code System calculations.

NUREG-2122, Glossary of Risk-Related Terms in Support of Risk-Informed Decisionmaking, issued November 2013, defines best estimate as used for deterministic calculations, in which best estimate designates inputs or results obtained by using the most realistic assumptions available to the analyst (i.e., not biased by conservatism or optimism).

4 The updated final analyses will be published in the forthcoming NUREG/CR-7262 report.

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on issues unique to ice condenser containments, including hydrogen generation and combustion. Moreover, the Sequoyah SOARCA analysis included an integrated UA, whereas the project team conducted the Peach Bottom and Surry UAs after the deterministic analyses.

The Sequoyah UA evaluated an unmitigated STSBO using insights from the Surry UA for the MELCOR analysis as well as the same input parameters that were varied in the Surry UA MACCS analysis. The results from the Sequoyah UA with respect to radionuclide release, individual early fatality risk, and individual latent cancer fatality risk also support those from previous SOARCA analyses for Peach Bottom and Surry. NUREG/CR-7245, State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Sequoyah Integrated Deterministic and Uncertainty Analyses, published October 2019, documents the Sequoyah SOARCA UA. The table below summarizes the SOARCA studies described above.

Report No. Title Reactor Type Scenarios Evaluated STSBO, LTSBO, NUREG- State-of-the-Art Reactor BWR-4 (Mark I thermally induced 1935 Consequence Analyses containment), 3-loop PWR SGTR, interfacing (2012) (SOARCA) (Parts 1 and 2) (large dry containment) systems loss of coolant accident*

NUREG/BR- BWR-4 (Mark I STSBO, LTSBO, 0359, containment), 3-loop PWR thermally induced Modeling Potential Reactor Revision 2 (large dry containment), SGTR, interfacing Accident Consequences (2016) 4-loop PWR (ice systems loss of condenser containment) coolant accident NUREG/CR- State-of-the-Art Reactor 7110, vol. 1, Consequence Analyses Project: BWR-4 (Mark I STSBO, LTSBO*

Revision 1 Volume 1: Peach Bottom containment)

(2013) Integrated Analysis STSBO, LTSBO, NUREG/CR- State-of-the-Art Reactor thermally induced 7110, vol. 2, Consequence Analyses Project: 3-loop PWR (large dry SGTR, interfacing Revision 1 Volume 2: Surry Integrated containment) systems loss of (2013) Analysis coolant accident State-of-the-Art Reactor Consequence Analyses Project:

NUREG/CR-Uncertainty Analysis of the BWR-4 (Mark I 7155 STSBO Unmitigated Long-Term Station Containment)

(2016)

Blackout of the Peach Bottom Atomic Power Station State-of-the-Art Reactor NUREG/CR- Consequence Analyses 4-loop PWR (ice 7245 (SOARCA) Project: Sequoyah STSBO, LTSBO condenser containment)

(2019) Integrated Deterministic and Uncertainty Analyses State-of-the-Art Reactor Consequence Analyses Project:

NUREG/CR- Uncertainty Analysis of the 3-loop PWR (large dry STSBO 7262 (Draft) Unmitigated Short-Term Station containment)

Blackout of the Surry Power Station

• A third scenario, loss of vital AC bus E-12, was evaluated for Peach Bottom but was determined not to lead to core damage.

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2 BENEFITS AND USES OF THE SOARCA PROJECT The SOARCA studies have increased the technical expertise of the staff in several areas, including UA, accident progression and accident consequence analyses, and atmospheric transport and dispersion. In accordance with its original objectives, the SOARCA project has also enabled enhanced communication of severe nuclear accident safety to external stakeholders. In addition to the numerous SOARCA reports referenced in the previous section, the NRC published a brochure, NUREG/BR-0359, with the specific intention of communicating up-to-date information about realistic severe accident consequences to the public. Other ways in which the staff has communicated SOARCA results include hosting public meetings near the plants studied in the SOARCA project to solicit public comments, as documented in Appendix C, SOARCA Public Comments Summary, of NUREG-1935, and presenting at NRC Regulatory Information Conference sessions, Cooperative Severe Accident Research Program (CSARP) meetings, International MACCS User Group (IMUG) meetings, and special SOARCA UA sessions organized for international Probabilistic Safety Assessment and Analysis and Probabilistic Safety Assessment and Management conferences. Additionally, the staff has communicated SOARCA insights to interested Federal partners, including the Federal Emergency Management Agency and the Federal Bureau of Investigation.

The SOARCA project has also identified improvements in NRC analytical tools, such as computer codes and associated severe accident analysis methodologies, including parametric UA. The MELCOR and MACCS computer codes and their modeling best practices have evolved significantly throughout the SOARCA project. NUREG/CR-7008, MELCOR Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA)

Project, and NUREG/CR-7009, MACCS Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project, both issued August 2014, document the MELCOR and MACCS best practices from the Peach Bottom and Surry deterministic studies, respectively. These reports document model improvements, modeling approaches, and parameter selection and explain the significance of the modeling improvements and approaches, as of 2012. The best practices further evolved during the later SOARCA studies and post-Fukushima severe accident risk evaluations that relied on SOARCA methods. These improvements are reflected in subsequent reports such as NUREG/CR-7245 for the Sequoyah SOARCA study. This documentation of best practices has helped provide the basis for the modeling choices made for accident analyses that followed, including the Site Level 3 PRA project. Further, a compilation of results and insights derived from the collective UAs is under development and should provide a useful reference for risk-informed regulatory activities. It should be noted that the MELCOR modeling best practices are continually being updated and communicated to code users at workshops as information becomes available from modeling improvements and code assessment.

The NRC has used insights, models, methodologies, and results from the SOARCA studies to support potential rulemaking activities, licensing activities, oversight, and regulatory decisionmaking, particularly in response to the accident at the Fukushima Daiichi nuclear reactors. The SOARCA project has also helped to risk-inform current NRC programs and projects such as the Site Level 3 PRA project, which further enhances the regulatory process.

The staff at the NRC and Sandia National Laboratories (SNL) involved with the SOARCA project have trained other NRC staff and international partners for knowledge management purposes and to help support the correct application of SOARCA tools and information. Lastly, other regulatory and research institutions, both domestic and international, have used or cited aspects of the SOARCA project in diverse research areas such as accident-tolerant fuel and dynamic PRA. The appendix of this RIL lists more than 325 citations of the SOARCA project in 5

these and other areas. This section summarizes several ways in which the NRC, reactor licensees and applicants, regulatory and research organizations, academia, and other external stakeholders have used the SOARCA studies.

2.1 Regulatory Support for NRC Post-Fukushima Activities The SOARCA studies for Peach Bottom and Surry were still ongoing when the Great Thoku earthquake and subsequent tsunami in Japan caused extensive damage to the nuclear reactors at the Fukushima Daiichi nuclear power plant, a BWR with a similar containment design (Mark I) as Peach Bottom. In the U.S. response to this accident, the NRC established the Near-Term Task Force (NTTF) to determine whether the NRC should make any near- or long-term improvements to its regulatory system and make recommendations to the Commission for its policy direction. The NTTF issued its Recommendations for Enhancing Reactor Safety in the 21st Century: The Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated July 12, 2011, providing recommendations to the Commission that were intended to clarify and strengthen the regulatory framework for protecting against natural disasters, for mitigation and emergency preparedness (EP), and for improving the effectiveness of existing NRC programs. The NTTF recommendations and additional issues identified by the staff included a Tier 1 recommendation 5 related to reliable hardened vents for BWR Mark I and Mark II containments and Tier 3 recommendations 6 related to hydrogen control, reliable hardened vents for containment designs other than BWR Mark I and Mark II, and expedited transfer of spent fuel to dry cask storage. The Commission later directed the staff to evaluate these recommendations to inform separate potential rulemaking efforts. The NRC used SOARCA insights and models as inputs for evaluating these recommendations, as described below.

Evaluation of Filtered Containment Venting and Containment Protection and Release Reduction Strategies for Boiling-Water Reactors with Mark I and Mark II Containments Following the Fukushima Daiichi accident, one of the major issues identified by NRC staff was whether to require licensees of BWRs with Mark I and Mark II containments to add capabilities for containment protection and release reduction following a potential loss of power accident.

These capabilities included installing filtered containment venting systems, as many other countries had done, but also included severe accident water addition and water management strategies. In SECY-12-0157, Consideration of Additional Requirements for Containment Venting Systems for Boiling Water Reactors with Mark I and Mark II Containments, dated November 26, 2012, the staff recommended adding filtration to reliable hardened vents. In SRM-SECY-12-0157, Staff Requirements - SECY-12-0157 - Consideration of Additional Requirements for Containment Venting Systems for Boiling Water Reactors with Mark I and Mark II Containments, dated March 19, 2013, the Commission directed the staff to develop the technical bases and rulemaking for filtering strategies with drywell filtration and severe accident management of BWR Mark I and II containments in support of NTTF Tier 1 5 Tier 1 recommendations include NTTF recommendations and additional issues that should be addressed without delay.

6 Tier 3 recommendations include NTTF recommendations and additional issues identified by the staff that require further staff study to support a regulatory action, have an associated shorter term action that needs to be completed to inform the longer-term action, or depend on the resolution of a NTTF Tier 1 recommendation.

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Recommendation 5.1. This effort was known as the containment protection and release reduction (CPRR) rulemaking.

The NRC staff were able to conduct timely technical analyses to inform the CPRR rulemaking effort by leveraging insights from the SOARCA project and using updated versions of plant-specific MELCOR and MACCS models from SOARCA in its analyses. If the SOARCA MELCOR and MACCS models were not available, the staff would have required significantly more time to conduct sufficiently detailed analyses to inform the CPRR rulemaking effort. The CPRR technical analyses included accident sequence analyses, accident progression and source term analyses, and offsite consequence analyses. The BWR Mark I accident progression and source term analysis used an updated version of the Peach Bottom SOARCA MELCOR model, and the offsite consequence analysis used an updated version of the Peach Bottom SOARCA MACCS site model. The calculated offsite consequences were weighted by accident frequency to assess the relative public health risk reduction associated with the various CPRR alternatives. The CPRR analyses showed the risk reduction benefits of having severe-accident-capable hardened vents, adding water to cool core debris, and adding filtered containment vents to Mark I and II containments.

Based in part on the CPRR technical analyses, the staff modified its risk-informed recommendations in SECY-15-0085, Evaluation of the Containment Protection and Release Reduction for Mark I and Mark II Boiling Water Reactors Rulemaking Activities, dated June 18, 2015, to no longer require external filters (as was done in SECY-12-0157) but instead make the requirements of Order EA-13-109, Issuance of Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions, dated June 6, 2013, generically applicable. In SRM-SECY-15-0085, Staff Requirements - SECY-15-0085 - Evaluation of the Containment Protection and Release Reduction for Mark I and Mark II Boiling Water Reactors Rulemaking Activities (10 CFR PART

50) (RIN-3150-AJ26), dated August 2015, the Commission terminated the CPRR rulemaking and directed the staff to continue implementation of Order EA-13-109 for severe-accident-capable vents, with no additional regulatory actions. It also directed the staff to leverage the draft regulatory basis to support the resolution of Tier 3 issues related to containments of other designs (i.e., NTTF Recommendation 5.2). The staff issued JLD-ISG-2015-01, Compliance with Phase 2 of Order EA-13-109, Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation under Severe Accident Conditions, Revision 0, dated April 2015, to assist nuclear power reactor licensees with the identification of methods that could be used to comply with the requirements of Order EA-13-109.

The CPRR technical analyses were published in detail in NUREG-2206, Technical Basis for the Containment Protection and Release Reduction Rulemaking for Boiling Water Reactors with Mark I and Mark II Containments, issued March 2018. As illustrated above, these analyses, which leveraged elements of the ongoing SOARCA project, supported risk-informed decisions that led to the continued safe operation of nuclear power plants, enhanced operational flexibilities, and offset costs associated with expensive plant modifications. These analyses serve as an example of evidence-based policymaking as discussed in the Foundations for Evidence-Based Policymaking Act of 2018. NUREG-2206 discusses the data, methods, and analytical approaches that were used to ensure NRCs actions on containment filtered venting, containment protection, and release reduction were based on evidence and sound methods and approaches.

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Evaluation of Tier 3 Issues Related to Containment Venting for Other than Mark I and Mark II Containments and to Hydrogen Control and Mitigation NTTF Tier 3 Recommendation 5.2 suggested that the staff assess whether to require installation of reliable, hardened venting systems for containments other than Mark I and II designs (i.e., BWR Mark III containments and PWR ice condenser and large dry containments).

In addition, NTTF Tier 3 Recommendation 6 suggested that the staff assess the need to strengthen requirements associated with hydrogen control and mitigation inside and outside reactor containment buildings. In SECY-15-0137, Enclosure 4, Proposed Resolution Plans for Tier 3 Recommendations 5.2 and 6: Reliable Hardened Vents for Other Containment Designs and Hydrogen Control and Mitigation Inside Containment and Other Buildings, dated October 29, 2015, the staff relied, in part, on results and insights from the Surry and Sequoyah SOARCA analyses to conclude that additional study of these topics would be unlikely to identify any regulatory actions beyond those already taken that would provide a substantial safety improvement for large dry and ice condenser containments. The abovementioned draft regulatory basis for the CPRR rulemaking (NUREG-2206) provided additional insights in support of this conclusion. In SECY-16-0041, Closure of Fukushima Tier 3 Recommendations Related to Containment Vents, Hydrogen Control, and Enhanced Instrumentation, dated March 31, 2016, the staff informed the Commission of its plan to formally close NTTF Recommendation 5.2 and 6 activities.

Evaluation of Expedited Transfer of Spent Fuel to Dry Cask Storage The Fukushima Daiichi nuclear accident renewed international interest in the safety of spent nuclear fuel stored in spent fuel pools (SFPs) under prolonged loss-of-cooling conditions.

Although the SFPs and spent fuel assemblies stored in the pools remained safe after the accident, it led the NRC to consider whether it should require the expedited transfer of spent fuel from spent fuel pools to dry cask storage at U.S. nuclear power plants. Shortly after the NTTFs report was released, the staff initiated a project to evaluate SFP accident consequences from a beyond-design-basis earthquake for a reference U.S. BWR with a Mark I containment.

As part of NRCs post-September 11 security assessments, SFP models that used detailed thermal-hydraulic and severe accident progression models integrated into the MELCOR code were developed and used to assess the realistic heatup of spent fuel under various pool draining conditions. The Peach Bottom MACCS model developed for the SOARCA studies was adapted for use in the SFP offsite consequence analysis. The results of the study, documented in NUREG-2161, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, issued September 2014, and in SECY-13-0112, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, dated October 9, 2013, indicated that expedited transfer of spent fuel to dry cask storage did not provide a substantial safety enhancement for the reference plant. This study was later expanded to inform a broader regulatory analysis (enclosed with COMSECY-13-0030,Staff Evaluation and Recommendation for Japan Lessons-Learned Tier 3 Issue on Expedited Transfer of Spent Fuel, dated November 12, 2013) for an NTTF Tier 3 issue of whether expedited transfer of spent fuel to dry cask storage at all U.S. nuclear power plants substantially enhances public health and safety.

The staff concluded that the expedited transfer of spent fuel to dry cask storage would provide only a minor or limited safety benefit and that its expected implementation costs were not warranted. In SRM-COMSECY-13-0030, Staff Requirements - Evaluation and Recommendation for Japan Lessons-Learned Tier 3 Issue on Expedited Transfer of Spent Fuel, dated May 23, 2014, the Commission approved the staff's recommendation to close the 8

Tier 3 Japan lessons-learned activities for expedited transfer and perform no further generic assessments.

2.2 Licensing and Design Certification Application Reviews One of the primary regulatory functions of the NRC is licensing (and renewing) operating reactors and certifying new reactor designs. As the current fleet ages, operating reactors are required to provide severe accident evaluations as part of the renewal process, while reactor designers who seek NRC approval of their new designs provide similar evaluations. 7 Operating reactor licensees and advanced reactor applicants have employed SOARCA computational tools and insights to support assumptions and methodologies in their respective renewal and design certification applications, and NRC technical reviewers have used these insights to review these applications. This section summarizes how the NRC has used SOARCA insights and computational tools for operating license renewals and for domestic and international light-water reactor design certification applications and reviews.

NuScale Application and the NRCs Review In December 2016, NuScale Power, LLC (NuScale), submitted to the NRC a design certification application that used SOARCA model improvements and insights for NuScales small modular reactor design. The NRC is using SOARCA model improvements and insights to perform its technical reviews. The following documents NuScales use of SOARCA insights:

• NuScale Standard Plant Design Certification Application Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, Revision 2. NuScale used the MELCOR and MACCS codes, methods, and assumptions containing SOARCA insights to evaluate accident progression and source term for the PRA and severe accident mitigation.

• NuScale Standard Plant Design Certification Application Environmental Report, Revision

2. NuScale used the MACCS codes, methods, and assumptions to calculate severe accident offsite consequences to evaluate the averted offsite costs and exposures to identify whether any SAMDA candidates are potentially cost beneficial.

• NuScale Licensing Topical Report No. TR-0915-17772, Revision 0, Methodology for Establishing the Technical Basis for Plume Exposure Emergency Planning Zones at NuScale Small Modular Reactor Plant Sites, issued December 2015. NuScale used the MELCOR and MACCS codes, methods, and assumptions to evaluate accident progression and source terms and to inform the selection of the required distance over which an emergency planning zone may be needed.

Because the SOARCA studies represent the state of the art in severe accident analysis, the accident progression and offsite consequence analysis methodologies used in the NuScale 7 The environmental report that is submitted as part of a license renewal application includes a severe accident mitigation alternative (SAMA) analysis. It identifies potentially cost-beneficial enhancements that could further reduce nuclear power plant risk. A design certification application for a new reactor includes analyses of severe accident mitigation design alternatives (SAMDAs), which are similar in scope to SAMA analyses.

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application and NRC license review process may also benefit other new reactor application processes and contribute to knowledge management.

APR1400 Design Certification Review and SOARCA-Like Analysis Korea Electric Power Corporation and Korea Hydro and Nuclear Power (KHNP) submitted their APR1400 design to the NRC for review in December 2014. The NRC staff used SOARCA insights in its review of the APR1400 design certification application, particularly when reviewing the applicants models. In Chapter 19 of the NRCs safety evaluation report, dated September 28, 2018, the agency referenced these insights when it approved the APR1400 design certification application.

In 2016, KHNP, with SNL support, also initiated SOARCA-like analyses of the APR1400 and Canada Deuterium Uranium (CANDU) reactors. The analysis leverages SOARCA project methodologies and MACCS best practices to perform realistic accident progression and consequence analyses for two pilot plants, Shin-Kori (APR1400) and Wolsung (CANDU).

KHNP gave an update on its SOARCA project at the NRCs 2018 Regulatory Information Conference. The NRC staff is following the research and collaborative efforts to identify any areas of interest and consideration of model updates based on these designs.

Contentions on Severe Accident Mitigation Alternatives during Indian Point Operating License Renewal Proceedings Part of the NRCs and Entergys expert testimonies in Atomic Safety and Licensing Board Panel proceedings for renewal of the operating license for Indian Point Nuclear Generating used insights from the 2012 SOARCA studies. For example, as part of NRC staff testimony on SAMA contentions submitted by the State of New York, experts referred to the peer-reviewed SOARCA study as a reference for realistic modeling of aerosol particle sizes and corresponding deposition velocities in offsite consequence modeling.

AREVA European Pressurized Reactor Application Prior to the SOARCA project, it was thought that a severe-accident-induced SGTR would lead to a large release because airborne fission products would flow from the reactor coolant system to the environment through the ruptured tube with little or no deposition in the reactor coolant system or steam generator. However, the SOARCA project showed that hot leg rupture would occur after a SGTR. Hot leg rupture results in most airborne fission products depositing in the containment, implying that the release of airborne fission products to the environment would be less than originally assumed.

AREVA NP submitted a design certification application for its European Pressurized Reactor to the NRC in 2007. During the staffs review of the application, AREVA NP developed an update to its application to reflect the SOARCA insight regarding hot leg failure subsequent to a SGTR.

This insight showed an additional margin in the European Pressurized Reactors ability to mitigate the severe accident source term. Nevertheless, AREVA NP requested that the NRC suspend the EPR application in 2015.

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2.3 Domestic and International Cooperation in Response to the Fukushima Daiichi Accident The Great Thoku earthquake and subsequent tsunami led to core melting and radioactive material release from Units 1, 2, and 3 at the six-unit Fukushima Daiichi nuclear power plant. In addition, hydrogen production from zirconium-cladding reactions with water led to hydrogen explosions in Units 1, 3, and 4. The NRC began immediate activities to support the Federal Government response to the event. In the aftermath of the earthquake and tsunami, domestic and international efforts were initiated to reconstruct the accident, assess severe accident modeling capabilities, and enhance offsite protective action decisionmaking. This section explains how SOARCA models, tools, and insights supported these activities.

Support for Federal Government Response to Fukushima The SNL staff, under contract to the U.S. Department of Energy (DOE), used the SOARCA thermal-hydraulic analysis of the Peach Bottom LTSBO scenario to estimate the accident progression at each affected Fukushima reactor. SNL used Peach Bottom SOARCA source terms adjusted for thermal power differences at Fukushima to perform MACCS offsite consequence analyses for DOE and the NRC. DOEs consequence management group also received SOARCA source terms and provided them to Lawrence Livermore National Laboratorys National Atmospheric Release Advisory Center. The Center used the modified SOARCA source terms for consequence analyses of the Fukushima accident to support the overall U.S. Federal Governments response to the accident.

Fukushima Incident Response One of the key objectives of the SOARCA studies was to enhance communication of severe-accident-related aspects of nuclear safety, including severe accident consequences, to the public and other stakeholders. Because of the similarities between the Fukushima Daiichi accident and Peach Bottom SOARCA calculations (both SBOs at BWRs with Mark I containments), Peach Bottom SOARCA insights were used in the immediate response to Fukushima. Specifically, the NRC Operations Center used Peach Bottom SOARCA MELCOR and MACCS models as starting points to generate source terms for the reactor and estimate doses around Fukushima. The staff also used preliminary SOARCA insights to prepare for the congressional hearings immediately after the accident at Fukushima, specifically noting that Peach Bottom had hardened vents and some pre-staged portable equipment, as required by Section B.5.b of Order EA-02-026, Order for Interim Safeguards and Security and Compensatory Measures, issued February 2002, as well as the resulting severe accident mitigation those plant features provide. A few months after the accident, the New York Times referenced the SOARCA study in its article, N.R.C. Lowers Estimate of How Many Would Die in Meltdown.

Fukushima Forensic Analysis Peach Bottom SOARCA calculations were still being analyzed when the Fukushima accident occurred. After the incident response phase of the accident, the NRC and DOE jointly sponsored an accident reconstruction study as a means of (1) better understanding the Fukushima accident progression and (2) assessing the severe-accident modeling capability of MELCOR. Using knowledge gained from the Peach Bottom SOARCA studies, SNL used MELCOR and existing information to predict how the accident may have progressed. SNL 11

published the results from this study in SAND2012-6173, Fukushima Daiichi Accident Study (Status as of April 2012), issued July 2012, which is publicly available on SNLs website.

Fukushima Uncertainty Analyses SNL conducted UAs of the Fukushima Unit 1 reactor using the Peach Bottom SOARCA SBO UA model as a starting point. The identification of important uncertain parameters was also informed by the Peach Bottom SOARCA UA of the LTSBO scenario, which had many similarities to the Fukushima accident progression. The goals of the Fukushima UAs were to evaluate uncertainty in core damage progression behavior and its effect on key figures of merit such as hydrogen production, fraction of intact fuel, and vessel lower head failure; to characterize the range of predicted damage states in the reactor considering uncertainties associated with MELCOR modeling; and to help inform the decommissioning activities available to Japanese decisionmakers to defuel the damaged reactors. SAND2016-0232, Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima ReactorsVolume I, issued January 2016, summarizes this work.

Benchmark Study of the Accident at Fukushima In 2012, the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) established a joint research project known as the Benchmark Study of the Accident at Fukushima (BSAF) to improve severe accident codes and analyze the accident progression of Fukushima Daiichi Units 1, 2, and 3. Sixteen organizations, including the NRC and the Electric Power Research Institute (EPRI), from eight countries joined the project, which was conducted in two phases. The first phase of the BSAF project focused on the first 6 days of the accident progression and estimated the current status inside the reactor pressure vessels and primary containment vessels for Units 1-3. Phase 1 of the BSAF project was completed in 2015. Phase 2 of the project expanded the work scope to include fission product behavior outside primary containment vessels, which lengthened the timespan for analyses of accident events to about 3 weeks after the earthquake. The Peach Bottom SOARCA MELCOR models were used as the starting point for the BSAF analyses to compare MELCOR against other accident progression computer codes such as the Modular Accident Analysis Program (MAAP),

Accident Source Term Evaluation Code (ASTEC), and SAMPSON codes. In addition, SOARCA-initiated developments to the MACCS code have been used to help analyze atmospheric dispersion and land contamination to help benchmark efforts to simulate the Fukushima accident progression. Phase 2 of the project was completed in 2018.

Benchmarking of Fast-Running Software Tools to Inform Offsite Decisionmaking After the accident at Fukushima Daiichi, it was observed that protective measures recommended to citizens occasionally differed by country, especially during the initial stages of the accident. Such differences could impact the projected radiological dose to members of the public. Because of this observation, a group within the OECD/NEA recommended that the agency analyse the comparison of source-term methodologies utilised by countries and determine if or why the dose prediction differed for Fukushima. To that end, the OECD/NEA sponsored a research project to benchmark fast-running software tools that are used to model offsite releases during a nuclear accident to support protective action decisionmaking. Twenty organizations, including the NRC, participated in the study. The study used Peach Bottom and Surry SOARCA source terms for two of the five scenarios that were selected to benchmark different fast-running software tools. The results of this study were used to explain why calculated source terms using the various codes and methodologies differed.

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NEA/CSNI/R(2015)19, Benchmarking of Fast-running Software Tools Used to Model Releases During Nuclear Accidents, issued January 2016, summarizes this work and provides information to participants to inform their own severe accident analysis codes.

2.4 Risk-Informing NRC Programs and Projects The NRC currently has several activities and initiatives to integrate risk information and performance measures into its regulations, guidance, and oversight processes in the regulatory arenas of reactor safety, material safety, and waste management. The NRC used risk insights from the SOARCA studies to improve or develop tools for risk-informed oversight activities and to support other risk-informing projects such as the Site Level 3 PRA project. The Nuclear Energy Institute (NEI) has also used SOARCA results in discussions on quantitative health objective (QHO) margins. This section highlights the ways in which the NRC has used SOARCA studies to risk-inform its current oversight programs and projects and summarizes NEIs use of SOARCA to discuss margins.

Probabilistic Risk Assessment Technical Development The NRC used the Peach Bottom SOARCA MELCOR model, in conjunction with other analytical tools, to investigate a longstanding PRA technical issue associated with the degree of credit warranted for emergency core cooling system injection following either containment venting or failure in BWR Mark I and II designs, specifically for PRA sequences in which adequate reactor pressure vessel injection is available but suppression pool heat removal is unavailable. Additionally, the NRC employed the Peach Bottom and Surry SOARCA MELCOR models for a second project to investigate specific thermal-hydraulic success criteria and sequence timing issues as documented in NUREG-1953, Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models Surry and Peach Bottom, issued September 2011. The agency used insights gained from MELCOR calculations to establish the technical bases for a subset of the success criteria used in the Standardized Plant Analysis Risk (SPAR) models for these reactor designs, which increased the realism of the models. The NRC uses SPAR models in its significance determination process for findings in the Reactor Oversight Process and other risk-informed oversight activities.

Site Level 3 PRA Project The Site Level 3 PRA project has leveraged SOARCA knowledge and insights to enhance staff capability and extract new risk insights that may be used to enhance regulatory decisionmaking. For example, the NRC used insights from the draft Surry SOARCA UA for a STSBO to help identify and determine the appropriate alternative treatment of many model uncertainties, and some parameter uncertainties, in the Level 2 analysis portion of its Level 3 PRA. In some cases, the MELCOR model parameter distributions that were developed for the Surry SOARCA UA helped define alternative treatments (MELCOR sensitivity runs). Having MELCOR sensitivity cases informed by the Surry SOARCA UA made the resulting sensitivities more informative. The documentation for the Site Level 3 PRA project refers the reader to the Surry SOARCA UA results for insights into the potential effects of model uncertainties for which alternative MELCOR sensitivity runs were not completed for the Site Level 3 PRA project.

The Site Level 3 PRA project also benefitted more broadly from the MELCOR and MACCS analytical improvements initiated by the SOARCA project. Examples of MACCS enhancements 13

prompted by the SOARCA project include the ability to more realistically model early phase protective actions by including multiple population cohorts and the ability to capture the effect of shifts in wind directions by implementing a 64-sector spatial grid. Finally, the work carried out on documenting the technical bases for MACCS input parameter values in SOARCA was valuable for developing the documentation for the Level 3 PRA offsite consequence analyses, which assisted in fulfilling the knowledge management objectives of the Site Level 3 PRA project.

Emergency Preparedness Significance Quantification Process The NRC staff initiated a project to develop a decision process for use in regulatory oversight that would help determine the risk significance of EP planning elements. This process, known as the DedUctive Quantification Index (DUQI) method, quantifies the value of EP program elements in terms of dose avoided, a value obtained from performing consequence analyses.

The NRC used the DUQI method in a proof-of-concept application to two representative nuclear power plant sites: a PWR at a site with high population density and a BWR at a site with medium population density. As part of this application, detailed consequence analyses were performed for two accident scenarios at each site using MELCOR and MACCS models from the Peach Bottom and Surry SOARCA studies. This work successfully illustrated one approach to risk-informing EP oversight at nuclear power plants and demonstrated that the DUQI method may be adapted to determine the risk significance of mitigating actions. Such risk information could help prioritize resources while enhancing overall safety, increasing public confidence, and reducing unnecessary regulatory burden. The study is documented in NUREG/CR-7160, Emergency Preparedness Significance Quantification Process: Proof of Concept, issued June 2013.

Seismic Probabilistic Risk Assessment Relief Request The NRC NTTFs Recommendation 2.1 identified the need for nuclear power plant licensees to reevaluate the seismic hazards at their sites against current NRC requirements and guidance and, if necessary, update the design basis and structures, systems, and components important to safety to protect against the updated hazards. In response, the NRC issued a 10 CFR 50.54(f) letter asking licensees to reevaluate the seismic hazards at their respective sites by March 31, 2014. The staff also developed a screening process to identify which plants needed to conduct a seismic PRA. Two plants that screened in were Duke Energys Catawba Nuclear Station and McGuire Nuclear Station, both of which have ice condenser containments.

Duke Energy submitted a letter requesting seismic PRA relief and referenced the Sequoyah SOARCA study, which was done for a PWR with an ice condenser containment, to support its request. Sequoyah SOARCA conclusions generally confirmed conclusions of previous ice condenser studies and helped the NRC make a timely risk-informed decision when responding to Dukes seismic PRA relief request. Ultimately, the NRC approved Dukes seismic PRA relief request.

Identification of Missing Emergency Action Level The SOARCA project led staff to identify an inconsistency in the emergency action level 8 (EAL) schemes for BWRs and PWRs. The EAL scheme for Peach Bottom would trigger a declaration of a general emergency for an event that led to an immediate loss of alternating current and 8 An EAL is a predetermined, site-specific, observable threshold for a plant condition that places the plant in an emergency class.

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direct current power, which was not the case for Surry. The EAL scheme was updated to address this, which enhanced the emergency preparedness of this plant.

NEI Use of SOARCA in Discussions on Margins In October of 2018, NEI sent a letter and white paper (which relied on an earlier EPRI white paper) to the NRCs Executive Director of Operations with the subject, Facilitating Regulatory Transformation through an Understanding of the Current Levels of Safety. The letter used results from recent NRC analyses (including SOARCA) to examine the perceived margin between the core damage frequency and large early release frequency surrogate objectives and the QHOs. In comments at a public meeting, the NRC staff cautioned about extrapolating results from studies such as SOARCA, which were intended for other purposes, to draw conclusions on QHO margins. At this public meeting, NRC staff also emphasized that focusing solely on margins between QHOs and surrogate objectives overlooks the part of the Commissions 1986 safety goal statement to continue to pursue a regulatory program that has as its objective providing reasonable assurance, while giving appropriate consideration to the uncertainties involved, that a severe core damage accident will not occur.

2.5 Enhancement of Existing Computer Models and Methodologies for Technical Analyses The Peach Bottom, Surry, and Sequoyah SOARCA studies provided detailed MELCOR and MACCS models of these sites that could be used as a starting point for other studies. The NRC has also used these models to enhance its incident response capabilities. This section summarizes the applications of SOARCA MELCOR and MACCS models and insights by the NRC, SNL, DOE, EPRI, and Argentinas Nuclear Regulatory Authority for other projects.

Analysis of Cyber-Attack Vectors on Nuclear Power Plants As a laboratory-directed research and development project, SNL developed a model for evaluating how cyber-attack vectors could cause core damage at U.S. nuclear power plants.

SNL developed MELCOR models to simulate accident progression using Peach Bottom and Surry SOARCA MELCOR models as starting points. This project provided important insights about how cyber-attack vectors could cause core damage. However, the insights, results, and reports generated from this work are proprietary and are, therefore, not publicly available.

Atucha II Level 2 and 3 Probabilistic Risk Assessment Technical Review Argentinas Nuclear Regulatory Authority contracted SNL to conduct an independent technical review of the Level 2 and 3 PRAs for the Atucha Unit 2 pressurized heavy-water nuclear power plant to support the safety of these plants. SNLs technical review used accident progression and offsite consequence insights from the SOARCA project, which improved the quality of the review. The insights, results, and reports generated from this work are proprietary and are, therefore, not publicly available.

Boiling-Water Reactor Owners Group Severe Accident Guidance DOE and EPRI are jointly preparing Revision 4 of the BWR Owners Groups severe accident guidelines. They are using the Peach Bottom SOARCA UA model and performing a dynamic 15

analysis, with respect to time and equipment failures, as they review the generic severe accident management guidelines applicable to Peach Bottom. The guidance generated from this work is proprietary and not publicly available.

Integrated System Response Modeling The Light Water Reactor Sustainability Program, sponsored by DOEs Office of Nuclear Energy, has begun a physical security initiative to link force-on-force modeling with reactor system response modeling. The goal of the effort is to develop modeling and simulations for existing nuclear power plant security regimes by using identified target sets to link dynamic assessment methodologies. This is being accomplished by leveraging nuclear power plant system-level modeling (i.e., SOARCA best practices MELCOR models) with force-on-force modeling and three-dimensional (3D) visualization to develop security tabletop scenarios. The impact of this effort is to create an integrated force-on-force and nuclear power plant system response framework that enables a holistic approach to determining security-related events as they relate to the potential onset of significant core damage.

To date, the MELCOR model (based on Three Mile Island Nuclear Station, Unit 2) has been adapted for physical security scenarios and is based on the open source Lone Pine Nuclear Power Plant notional facility. The MELCOR model uses SOARCA insights and SOARCA best practices and has been linked with a dynamic event/fault tree scheduler, ADAPT. The force-on-force scenarios have been developed within the SCRIBE 3D software, and this software is currently being linked with ADAPT. The report SAND2019-12015, issued in October 2019, summarizes the proof-of-concept integrated system response model.

MELCOR Integration with the Emergency Response Data System The MELCOR Accident Simulation Trainer (MAST), a MELCOR postprocessing tool, was developed to feed MELCOR outputs to the Emergency Response Data System. The NRC used the Peach Bottom MELCOR SOARCA model as input to train its emergency responders about what to expect during station blackouts at a Mark I BWR.

Enhancement of the RASCAL Code for Incident Response At the time of the accident at Fukushima, the Radiological Assessment System for Consequence AnaLysis (RASCAL) code, the NRCs primary code for incident response, could not calculate source terms for LTSBOs. Using information obtained from SOARCA studies, RASCAL was modified to include a LTSBO option that incorporates core release fractions and accident progression timings for both PWRs and BWRs. This addition enabled LTSBO scenarios to be modelled in real time during an event. NUREG-1940, Supplement 1, RASCAL 4.3: Description of Models and Methods, issued May 2015, provides more details about specific modifications to RASCAL as well as comparisons between RASCAL and SOARCA results.

Modern Radiological Source Terms for MACCS Model Testing and Benchmarking Prior to SOARCA, radiological source terms used for consequence modeling and testing were highly simplified representations of complex patterns of radiological releases under hypothetical severe accident conditions. Radiological source terms developed in SOARCA provide more 16

realistic examples of the potentially complex time-dependent radiological release patterns from hypothetical severe accidents. These source terms are being used to test MACCS model performance using more realistic release patterns. Current examples include the use of SOARCA source terms to examine the potential differences in consequence estimates obtained using the MACCS simplified, fast-running Gaussian plume segment model and those using the more modern HYSPLIT atmospheric transport and dispersion code developed and maintained by the National Oceanic and Atmospheric Administration. To assist NRC staff in preparing to evaluate potential requests for reduced emergency planning zone sizes for new and advanced reactor technologies, SOARCA source terms are also being used by NRC staff to examine the sensitivity of radiological dose estimates to different consequence modeling choices.

2.6 Knowledge Management One of the primary objectives of the SOARCA project was to enhance the communication of severe accident aspects of nuclear safety to internal (NRC) and external stakeholders. In addition to publishing SOARCA reports and delivering presentations at conferences, staff members at NRC and SNL have also provided training to domestic and international stakeholders for knowledge management purposes. This section describes the training sessions provided to diverse stakeholders about the SOARCA studies.

Cooperative Severe Accident Research Program The Cooperative Severe Accident Research Program (CSARP) is an international program on severe accident phenomenological research and code development activities organized by the NRC since 1988. The objective of CSARP is the exchange of data and analyses on experimental and analytical research on severe accidents. Examples of information exchanged include the SOARCA studies, Fukushima lessons learned, the Phebus Fission Product Program, NEAs Behavior of Iodine Project, mixed-oxide and high-burnup fuel fission product releases, QUENCH experiments, and the OECD/NEA Melt Coolability and Concrete Interaction project. Several other entities operate within CSARP, including the European MELCOR and MACCS Users Group, the Asian MELCOR and MACCS Users Group, and IMUG.

NUREG/BR-0524, Cooperative Severe Accident Research Program (CSARP), issued November 2015, provides more information about CSARP and lists SOARCA as an example of information exchanged through the program.

NRC-Sponsored Training The NRC and DOE National Laboratories teach SOARCA insights through their Accident Progression Analysis, Accident Consequence Analysis, and Perspectives on Reactor Safety training courses. These courses are part of several NRC formal qualification programs, such as those for reliability and risk analysts, senior reactor analysts, and MACCS analysts. Several training classes and workshops offered during CSARP and IMUG meetings used material from SOARCA analyses.

Boiling-Water Reactor Owners Group and Nuclear Energy Expert Group Training The BWR Owners Group has a technical support guidance workshop, and SNL has been invited to discuss severe accident phenomenology insights, some of which come from 17

SOARCA. The Center for Strategic and International Studies hosts Nuclear Energy Expert Group meetings once or twice a year in Southeast Asia. These meetings include attendees from Southeast Asian countries who are interested in building nuclear reactors in their countries.

The SNL staff gave a presentation that included SOARCA project information at a Center for Strategic and International Studies meeting in 2016 on nuclear accidents and incident response.

Sandia National Laboratories Training to Korea Hydro and Nuclear Power KHNP contracted with SNL to provide MELCOR and MACCS training to support its SOARCA-like consequence analysis of the APR1400. The training materials used SOARCA modeling and insights.

3 CONCLUSIONS The SOARCA project leveraged decades of research on severe accident phenomenology and radiation effects and used site- and plant-specific data to evaluate the consequences arising from realistic severe accident scenarios at Peach Bottom, Surry, and Sequoyah. In addition, the studies incorporated significant plant improvements and changes not reflected in earlier reactor assessments and evaluated the potential benefits of security-related enhancements aimed at preventing core damage and reducing or delaying any potential offsite releases.

To achieve these objectives, the NRC used a five-step process to identify selected scenarios and model accident progression and mitigative measures, offsite release of radiological material, emergency response, and health effects. The results of the analyses consistently predict essentially zero individual early fatality risk for the modeled scenarios, very low long-term cancer fatality risks, and smaller radiological releases than those predicted from previous studies. The results also highlighted the importance of successful implementation of mitigative measures for preventing core damage. Additional UAs were conducted for these plants to help ensure the robustness of the models used and to identify key uncertain parameters that strongly influenced results. Importantly, the results from these studies support the results obtained from the deterministic evaluations of the three plants.

In addition to providing a more realistic estimate of severe accident consequences, the SOARCA project helped develop staff expertise in UA, accident progression and source term analysis, and consequence analysis. The project also led to improvements in NRC severe accident computer codes and accident analysis methodologies and provided detailed site- and plant-specific models ready for use by the NRC and other stakeholders for additional analyses.

In accordance with the objectives of the project, the NRC has communicated safety aspects of severe accidents to its staff and external stakeholders by holding public meetings, presenting the project at internal seminars and external conferences, and publishing nine SOARCA NUREG-series publications, including a plain-language color brochure summarizing the project.

The NRC and external stakeholders have used SOARCA computer models, methodologies, and insights in several ways. For example, they were used to support risk-informed rulemaking efforts resulting from NTTF Report recommendations, which have enhanced safety and security, while also supporting operational flexibilities, of operating nuclear power plants in the U.S. Operating reactor licensees and advanced reactor designers have used SOARCA insights to support assumptions and methodologies in the respective renewal and design certification applications, and the NRC staff has used them in reviewing these applications. During and after the accident at Fukushima, SOARCA was used for emergency response and for domestic and international efforts to reconstruct the accident, assess severe accident modeling capabilities, 18

and enhance offsite protective action decisionmaking. Other countries have leveraged SOARCA models and insights to improve their severe accident progression and consequence analysis capabilities. Ongoing risk-informed projects such as the Site Level 3 PRA project and NRC oversight programs have also benefited from risk insights leveraged by the SOARCA studies. Lastly, the up-to-date, detailed SOARCA computer models for Peach Bottom and Surry have been used as starting points for additional model development and to enhance real-time emergency response capabilities.

At least 325 conference papers, journal articles, presentations, and technical reports have used or cited the SOARCA project for scholarly research in diverse areas such as accident-tolerant fuels, advanced reactors, and dynamic PRA. The SOARCA project has proven to be useful in many ways beyond its original objectives and has been instrumental in ensuring the NRC staff has the technical capabilities to analyze the nuclear power safety issues of the future.

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25 APPENDIX A LIST OF TECHNICAL LITERATURE CITING THE SOARCA PROJECT This appendix lists the results of an extensive literature search for citations of the SOARCA project. Analysts identified more than 325 citations by using the following search terms in Google, Primo, and other search engines: SOARCA, NUREG-1935, NUREG/BR-0359, NUREG/CR-7110, NUREG/CR-7008, NUREG/CR-7009, and NUREG/CR-7155. While the list is not exhaustive, it serves to illustrate the volume of citations and variety of SOARCA applications to research that were identified as of May 2019. The list groups the citations by topic to help the reader identify areas of interest, and links to each citation are provided where possible. To move to a subject of interest, click on the topic heading from the list below.

Research Topic Severe Accident Progression Analysis Offsite Consequence Analysis Accident-Tolerant Fuel Advanced Reactor Analysis Emergency Preparedness and Response Dynamic PRA Multi-Unit PRA Fukushima Forensic Analysis Fukushima Lessons Learned Severe Accident Management Level 1 PRA Nuclear Safety/Societal Risk Sensitivity and Uncertainty Analysis Storage and Transportation Safety A-1

I. Severe Accident Progression Analysis Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Heat Up and Failure of BWR Upper Nuclear Engineering and 1 Robb, K. R. April 2017 Internals During a Severe Accident Design, vol. 314, p.293-306 Terry Turbopump Analytical Modeling Ross, K.; Cardoni, 2 Efforts in Fiscal Year 2016 - Progress April 2018 SAND2018-4337 J. N.; Osborn, D.

Report Terry Turbopump Expanded Operating 3 Osborn, D. July 2017 SAND2017-6715C Band Consequences of Degraded Jankovsky, Z.; Transactions of the American 4 Containment in a Severe Nuclear Jones, C.; November 2014 Nuclear Society, vol. 111, no.

Power Plant Accident Kalinich, D. 1 (2014 ANS Winter Meeting)

MELCOR Code Source Term Korean Nuclear Society 5 Characteristics for Fast SBO Scenario Han, S. J. et al. October 2012 Autumn Meeting, Gyeongju, of OPR1000 Plant Korea, October 25-26, 2012 Korean Nuclear Society Effects of Source Term Characteristics Han, S. J.; Ahn, K.

6 October 2012 Autumn Meeting, Gyeongju, on Off-Site Consequence I.

Korea, October 25-26, 2012 Development of the SharkFin Transactions of the American 7 Distribution for Fuel Lifetime Estimates Denman, M. R. November 2016 Nuclear Society, vol. 115, no.

in Severe Accident Codes 1 (2016 ANS Winter Meeting)

ANS PSA 2013 International Topical Meeting on Simplified Method for Assessing the Probabilistic Safety 8 Risk Associated with Consequential Azarm, M. A. et al. September 2013 Assessment and Analysis, Steam Generator Tube Rupture Events Columbia, SC, September 22-26, 2013 Models and Methods Related to Severe 9 Gauntt, R. O. April 2015 SAND2015-3101PE Accidents and Source Term Evaluation Comparative Analysis on the Influence of the MAAP4 Phenomenological Kang, S. W. and Annals of Nuclear Energy, 10 Model Parameters for the Severe July 2018 Yim, M.-S. vol. 117, p.98-108 Accident Source Term for Different Plant Designs and Accident Scenarios MAAP-MELCOR Crosswalk Phase 1 Nuclear Technology, vol. 196, 11 Luxat, D. L. et al. December 2016 Study p. 684-697 A-2

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number International Atomic Energy Agency Workshop on Severe 12 Accident Consequences and Analysis Gauntt, R. O. December 2017 Accident Management Guidelines, December 11-15, 2017 Reactor Safety Gap Evaluation of Farmer, M. T. et 13 Accident Tolerant Components and January 2015 ANL/NE-15/4 al.

Severe Accident Analysis A Simple Assessment Scheme for Silva, K. and Nuclear Engineering and 14 Severe Accident Consequences using August 2016 Okamoto, K. Design, vol. 305, p. 688-696 Release Parameters Station Blackout Mitigation Strategies Annals of Nuclear Energy, 15 Analysis for Maanshan PWR Plant Lin, H.-T. et al. March 2016 vol. 89, p. 1-18 using TRACE Fernandez-External Flooding Event Analysis in a Annals of Nuclear Energy, 16 Cosials, M. K. et February 2015 PWR-W with MAAP5 vol. 76, p.226-236 al.

A Reassessment of Low Probability Brunett, A.;

Containment Failure Modes and Nuclear Technology, vol.186, 17 Denning, R.; May 2014 Phenomena in a Long-Term Station p. 198-215 Aldemir, T.

Blackout The Ultimate Response Guideline Simulation and Analysis using TRACE, Annals of Nuclear Energy, 18 Wang, T.-C. et al. May 2017 MAAP5, and FRAPTRAN for the vol. 103, p. 402-411 Chinshan Nuclear Power Plant U.S. Nuclear Regulatory Commission's 19 State-of-the-Art Reactor Consequence Osborn, D. M. October 2017 SAND2017-10631PE Analyses (SOARCA) Project Conceptual Design Enhancement for Nuclear Technology, vol. 188, 20 Prevention and Mitigation of Severe Song, J. H. November 2014

p. 113-122 Accidents Ongoing SNL International Severe 21 Andrews, N.C. August 2018 SAND2017-9327R Accident Activities Advanced Seismic Probabilistic Risk Assessment Methodology:

22 Bolisetti, C. et al. August 2017 INL/EXT-17-43148 Development of Beta 1.0 MASTODON Toolset A-3

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Thermal-Hydraulic Study of Air-Cooled Passive Decay Heat Removal System Nuclear Engineering and 23 Kim, D. Y. et al. February 2019 for APR+ under Extended Station Technology, vol. 51, p. 60-72 Blackout Integral Analyses of Fission Product Retention at Mitigated Thermally- Rýdl, A.; Lind, T.; Nuclear Engineering and 24 February 2016 Induced SGTR using ARTIST Birchley, J. Design, vol. 297, p. 175-187 Experimental Data An Analysis of Radiological Releases Vo, T. H.; Kim, D. Nuclear Engineering and 25 during a Station Black Out Accident for June 2018 H.; Song, J. H. Design, vol. 332, p.22-30 the APR1400 Estimating Safety Valve Stochastic Failure-to-Close Probabilities for the ASME/NRC 2017 13th Pump 26 Ghosh, S. T. et al. July 2017 Purpose of Nuclear Reactor Severe and Valve Symposium Accident Analysis Organisation for Economic Co-Safety Research Opportunities Post-operation and 27 Fukushima: Initial Report of the Senior February 2017 NEA/CSNI/R(2016)19 Development/

Expert Group Nuclear Energy Agency 28 SRV Modeling Phillips, J. August 2017 SAND2017-8365PE Refined Boiling Water Reactor Station 29 Zhao, H. et al. September 2014 INL/EXT-14-33162 Blackout Simulation with RELAP-7 Developing Fully Coupled Dynamical Reactor Core Isolation System Models 30 Zhao, H. et al. April 2014 INL/CON-13-29971 in RELAP-7 for Extended Station Black-Out Analysis RCIC Governing Equation Scoping Cardoni, J. N. and 31 August 2015 SAND2015-6996C Studies for Severe Accidents Ross, K.

Demonstration of Fully Coupled 32 Simplified Extended Station Black-out Zhao, H. et al. October 2014 INL/CON-13-30940 Accident Simulation with RELAP-7 Post-Fukushima Critical Safety 33 Osborn, D. November 2015 SAND2015-10295C Equipment Evaluation A-4

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number 16th International Topical A Strongly Coupled Reactor Core Meeting on Nuclear Reactor 34 Isolation Cooling System Model for Zhao, H. et al. September 2015 Thermal Hydraulics Extended Station Black-Out Analyses (NURETH-16), Chicago, USA Modeling of the Reactor Core Isolation SAND-2015-10662 35 Cooling Response to Beyond Design Ross, K. et al. December 2015 Basis Operations - Interim Report 36 Critical Equipment Performance Osborn, D. September 2015 SAND2015-8178PE Cardoni, J. N.;

RCIC Governing Equation Scoping 37 Ross, K.; Gauntt, August 2015 SAND2015-6995C Studies R. O.

Probabilistic Economic Valuation of 38 Riley, T. et al. July 2018 DOE contract NE0008295 Safety Margin Management SARNET Benchmark on Phébus FPT3 Integral Experiment on Core Annals of Nuclear Energy, 39 Di Giuli, M. et al. July 2016 Degradation and Fission Product vol. 93, p. 65-82 Behaviour State-of-the-Art Report on Molten 40 Corium Concrete Interaction and Ex- OCED/NEA January 2017 NEA/CSNI/R(2016)15 Vessel Molten Core Coolability Modeling of Water Ingression Nuclear Technology, vol. 197, 41 Mechanism for Corium Cooling with Sevón, T. February 2017 no. 2, p. 171-179 MELCOR Exercises in Severe Accident Analysis 42 Gauntt, R. O. April 2015 SAND2015-3100PE using MELCOR: Accident Walkthrough Sensitivity Analysis of Debris Properties Galushin, S. and Nuclear Engineering and 43 June 2018 in Lower Plenum of a Nordic BWR Kudinov, P. Design, vol. 332, p. 374-382 44 SOARCA Modeling Phillips, J. August 2017 SAND2017-8638PE Nuclear Severe Accident Modeling and 45 Humphries, L. L. November 2015 SAND2015-10296PE Analysis Overview of Reactor Safety and Severe 46 Accident Analysis Technologies at Gauntt, R. O. June 2007 SAND2007-4143C Sandia National Laboratories Rýdyl, A.;

Modeling of Aerosol Fission Product Fernandez- Nuclear Technology, vol. 205, 47 Scrubbing in Experiments and in May 2019 Moguel, L.; Lind, no. 5, p. 655-670 Integral Severe Accident Scenarios T.

A-5

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Investigation of the Recriticality Darnowski, P.;

Annals of Nuclear Energy, 48 Potential during Reflooding Phase of Potapczyk, K.; January 2017 vol. 99, p. 495-509 Fukushima Daiichi Unit-3 Accident wirski, K.

Cardoni, J.; Transactions of the American Severe Accident Modeling for Cyber 49 Denman, M. November 2016 Nuclear Society, vol. 115, no.

Scenarios Wheeler, T. 1, p. 837-840 Updated Peach Bottom Model for 50 MELCOR 1.8.6: Description and Robb, K. R. September 2014 ORNL/TM-2014/207 Comparisons Analisi di un Incidente Non Mitigato di Tipo LOCA in un Reattore PWR 51 Pescarini, M. June 2016 LM-DM270 Mediante il Codice MELCOR 2.1 (Dissertation)

MELCOR 2.1 Analysis of Melt Behavior Annals of Nuclear Energy, 52 in a BWR Lower Head during LOCA Li, G. et al. April 2016 vol. 90, p. 195-204 and SBO Accident Heat Up and Potential Failure of BWR 53 Upper Internals during a Severe Robb, K. R. January 2015 OSTI ID 1213335 Accident Enhanced MELCOR 2.1 Models for Standard PWR-Westinghouse Ruiz Zapatero, M. Nuclear Espana, vol. 382, p.

54 Design/Modelo Detallado para un March 2017 et al. 49-54 Diseno Estándar PWR-Westinghouse con el Código MELCOR 2.1 Phillips, J. and 55 MELCOR Workshop OJP August 2017 SAND2017-8243PE Humphries, L.

Quicklook Overview of Model Changes 56 Humphries, L. L. May 2017 SAND2017-5599 in Melcor 2.2: Rev 6342 to Rev 9496 Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase 57 Andrews, N. et al. October 2017 SAND2017-11975 II Analyzing a Partially Recovered Accident Scenario Evaluation of JRC Source Term Methodology using MAAP5 as a Fast- Vela-García, M. Annals of Nuclear Energy, 58 October 2016 Running Crisis Tool for a BWR4 Mark I and Simola, K. vol. 96, p. 446-454 Reactor A-6

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number 2014 22nd International Long Term SBO with Selected Conference on Nuclear Mitigative Measures: MELCOR 59 Rydl, A. et al. July 2014 Engineering (ICONE22),

Parametric Calculations for a 2-Loop paper no. ICONE22-30351, p.

PWR V006T15A012 Proceedings of the Korean Benchmarking Simulation of Long Term Kim, S. K.; Lee, J.

60 May 2013 Nuclear Society 2013 spring Station Blackout Events C.; Fynan, D. A.

meeting OECD/NEA Workshop on The Role of Severe Accident Implementation of Severe 61 Management in the Advancement of Helton, D. et al. October 2009 Accident Management Level 2 PRA Modeling Techniques Measures - Oct 2009 -

Switzerland LEVEL 2 PRA: Explicit and Seamless 62 Approach (Example of a PWR with Azarm, M. A. March 2016 n/a Large Dry Containment)

Level-2/Level-3 Interface in a PSA 63 Bixler, N. E. April 2015 SAND2015-3245PE Analysis Development of a Fully-Coupled, All Zvoncek, P.; Nuclear Engineering and 64 States, All Hazards Level 2 PSA at Nusbaumer, O.; March 2017 Technology, vol. 49, no. 2, p.

Leibstadt Nuclear Power Plant Torri, A. 426-433 Nakamura, K.;

Overview of Revised Level 2 PRA 65 Narumiya, Y.; June 2016 ICONE24-61070 Standard in Japan Abe, Y.

Helton, D. M.; Proceedings of the 2014 Focus Areas for a Level 2 PSA That 66 Zavisca, M.; September 2014 European Safety and Supports a Site NPP Risk Analysis Khatib-Rahbar, M. Reliability Conference Some International Efforts to Progress OECD/NEA Workshop on in the Harmonization of L2 PSA Implementation of Severe 67 Development and Their Applications Raimond, E. et al. October 2009 Accident Management (European (ASAMPSA2), U.S. NRC, Measures - Oct 2009 -

OECD-NEA and IAEA Activities) Switzerland BWR MARK I Pressure Suppression Pool Mixing and Stratification Analysis Ozdemir, O. E. Annals of Nuclear Energy, 68 November 2015 using GOTHIC Lumped Parameter and George, T. L. vol. 85, p. 532-543 Modeling Methodology Proceedings of the 2015 In-Plant Fission Product Behavior in Kim, T. W.; Han, 69 May 2015 Korean National Society SGTR Accident with Long-Term SBO S. J.; Ahn, K. I.

Spring Meeting A-7

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number OECD/NEA/CSNI Status Report on Jacquemain, D. et 70 July 2014 NEA-CSNI-R--2014-7 Filtered Containment Venting al.

Assessment of Primary and Secondary Bleed and Feed Procedures during a Gómez-García- Annals of Nuclear Energy, 71 March 2018 Station Blackout in a German Konvoi Torano, I. et al. vol. 113, p. 476-492 PWR using ASTECV2.0 Analysis of Flammability in the de la Rosa, J. C. Nuclear Engineering and 72 Attached Buildings to Containment November 2016 and Fornós, J. Design, vol. 308, p. 154-169 under Severe Accident Conditions Risk-Informed External Hazards 73 Analysis for Seismic and Flooding Parisi, C. et al. July 2017 INL/EXT-17-42666 Phenomena for a Generic PWR Total Loss of AC Power Analysis for Darnowski, P. et Nuclear Engineering and 74 August 2015 EPR Reactor al. Design, vol. 289, p. 8-18 NUTHOS-11: The 11th International Topical Meeting Development of Core Relocation on Nuclear Reactor Thermal Surrogate Model for Prediction of 75 Galusin, S. et al. October 2016 Hydraulics, Operation and Debris Properties in Lower Plenum of a Safety, Gyeongju, Korea, Nordic BWR October 9-13, 2016, Paper N11P1234 Ex-Vessel Core Melt Modeling Robb, K. R.;

76 Comparison between MELTSPREAD- Farmer, M.; March 2014 ORNL/TM--2014/1 CORQUENCH and MELCOR 2.1 Francis, M. W.

HCVS-WP-02: Sequences for HCVS Design and Method for Determining Nuclear Energy ADAMS Accession No.

77 October 2014 Radiological Dose from HCVS Piping Institute ML14309A588 Revision 0 Effects of Degradation on the Severe NUREG/CR-7149 Accident Consequences for a PWR 78 Petti, J. P. et al. June 2013 Plant with a Reinforced Concrete ADAMS Accession No.

Containment Vessel ML13172A089 Korean Nuclear Society Effect of Cesium-Molybdate on Cs Han, S. J. and 79 October 2013 Autumn Meeting, Gyeongju, Behavior for Source Term Estimation Ahn, K. I.

Korea, October 23-25, 2013 BWR Station Blackout: A RISMC 80 Analysis using RAVEN and RELAP5- Mandelli, D. et al. January 2016 INL/JOU-14-33448 3D A-8

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Influence of the Wet-Well Nodalization of a BWR3 Mark I on the Containment Herranz, L. E. et Nuclear Engineering and 81 December 2015 Thermal-Hydraulic Response during an al. Design, vol. 295, p. 138-147 SBO Accident External Cooling of the BWR Mark I and II Drywell Head as a Potential 82 Robb, K. R. August 2017 ORNL/TM-2017/457 Accident Mitigation Measure - Scoping Assessment Industry Guidance for Compliance with Order EA-13-109: BWR Mark I & II NEI 13-02 [Rev. 0F4]

Nuclear Energy 83 Reliable Hardened Containment Vents March 2015 ADAMS Accession No.

Institute Capable of Operation Under Severe ML15084A454 Accident Conditions Benchmarking of Fast-running Software Nuclear Energy 84 Tools Used to Model Releases During January 2016 NEA/CSNI/R(2015)19 Agency Nuclear Accidents ATLAS Program for Advanced Song, C.-H.; Choi, Nuclear Engineering and 85 December 2015 Thermal-Hydraulic Safety Research K.-Y.; Kang, K.-H. Design, vol. 294, p. 242-261 Analysis of Primary Bleed and Feed Strategies for Selected SBLOCA Gómez-García- Annals of Nuclear Energy, 86 December 2017 Sequences in a German Konvoi PWR Torano, I. et al. vol. 110, p. 818-832 using ASTEC V2.0 A Study on Fission Product Behavior Proceedings of the 2015 Fall Kim, H.-C. and 87 during a Severe Accident at APR1400 October 2015 meeting of the Korean Cho, S.-W.

Nuclear Power Plants Nuclear Society A Study on Scenario Selection for Proceedings of the 2015 Evaluation of Fission Product Behavior Yoon, E. S.; Kim, 88 May 2015 Spring meeting of the Korean during a Severe Accident at APR 1400 H.-C.; Cho, S.-W.

Nuclear Society Nuclear Power Plants Updated analysis of Fukushima Unit 3 Fernandez-Annals of Nuclear Energy, 89 with MELCOR 2.1. Part 2: Fission Moguel, L.; Rydl, August 2019 vol. 130, p.93-106 product release and transport analysis A.; Lind, T.

An analysis on the consequences of a Nuclear Engineering and 90 severe accident initiated steam Song, J.-H. et al. July 2019 Design, vol. 348, p. 14-23 generator tube rupture Development of MELCOR thermal Annals of Nuclear Energy, 91 hydraulic model of AP1000 and its Malickia, M. et al. June 2019 vol. 128, p. 44-52 verification for a DECL break A-9

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Analysis of the Effect of Severe Science and Technology of Accident Scenario on Debris Properties Galushin, S. and Nuclear Installations, vol.

92 April 2019 in Lower Plenum of Nordic BWR Using Kudinov, P. 2019, Article ID 5310808, 18 Different Versions of MELCOR Code pages Science and Technology of Effect of Molten Corium Behavior Nuclear Installations, vol.

93 Uncertainty on the Severe Accident Choi, W. et al. August 2018 2018, Article ID 5706409, 9 Progress pages Journal of Nuclear Science Analysis of PWR SBO sequences with Mena-Rosell, L. et 94 June 2018 and Technology, vol. 55, no.

RCP passive thermal shutdown seals al.

6, p. 649-662 Consequential SGTR Analysis for Westinghouse and Combustion Sancaktar, S. et ADAMS Accession No.

95 Engineering Plants with Thermally May 2018 al. ML18122A012 Treated Alloy 600 and 690 Steam Generator Tubes European MELCOR User MELCOR Code Development Status -

96 Humphries, L. L. April 2018 Group Meeting, Zagreb, EMUG 2018 Croatia, April 18-20, 2018 Severe Accident Context Quantification ASME Journal of Nuclear for Long-Term Station Blackout in Petkov, G. I. and 97 March 2018 Engineering and Radiation Boiling Water Reactor Nuclear Power Vela-Garcia, M.

Science, vol. 4, no. 2, 020913 Plants Tools and Methods for Assessing the Azarm, M. A. and ADAMS Accession No.

98 March 2018 Risk Associated with CSGTR Sancaktar, S. ML18074A025 PWR Owners Group SAMG 99 Severe Accident Phenomena Wagner, K. C. March 2018 Meeting, Denver, CO, United States, March 27-28, 2018 MELCOR Best Practices in SOARCA 100 Humphries, L. November 2017 SAND2017-12028C AMUG 2017 IVR Phenomena Modeling of Lower Asian MELCOR User Group 101 Plenum and Modifications of RN Humphries, L. L. November 2017 Meeting, Daejeon, South Package (1.8.6 to 2.2) Korea, November 5-8, 2017 MELCOR Code Assessment MELCOR Code Development Status Program Meeting, Bethesda, 102 Humphries, L. L. October 2017 MCAP 2017 MD, USA, September 14-15, 2017 A-10

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Terry Turbopump Expanded Operating 103 Band Program Experimental and Osborn, D. September 2017 SAND2017-10566PE Modeling Efforts An Improvement of Estimation Method Journal of Radiation of Source Term to the Environment for Han, S.-J.; Kim, 104 June 2017 Protection and Research, vol.

Interfacing System LOCA for Typical T.-W.; Ahn, K.-Il.

42, no. 2, p. 106-113 PWR Using MELCOR code Analysis of unmitigated large break loss Journal of Physics:

105 of coolant accidents using MELCOR Pescarini, M. et al. June 2017 Conference Series, vol. 923, code Conference 1 A-11

II. Offsite Consequence Analysis Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Article, "NRC Keeps Its Focus on Nuclear News, vol. 51, no. 6 1 n/a May 2008 Safety," in "Meetings" section p. 53 A Sensitivity Study on Population Transactions of the American 2 Segmentation Effects in the MACCS Song, W. et al. June 2018 Nuclear Society, vol. 118, no.

Code 1, p.643-646 Article, Assessing SOARCA, in Nuclear News, vol. 56, no. 1, 3 n/a January 2013 Meetings section p. 59 Nuclear Energy Related Capabilities at 4 Pickering, S. Y. February 2014 SAND2014-1345 Sandia National Laboratories Technical Basis Development for Transactions of the American 5 Filtered Containment Venting System Basu, S. et al. November 2013 Nuclear Society, vol. 109, no.

Requirements 1, p. 971-974 Transactions of the American MELCOR/MACCS2 Analysis for BWR Osborn, D. M. et 6 November 2013 Nuclear Society, vol. 109, no.

Mark I Filtered Containment Venting al.

1, 2103-2106 MACCS-HYSPLIT Atmospheric 7 Transport and Dispersion Model Clayton, D. J. September 2017 SAND2017-9714PE Benchmarking MACCS-HYSPLIT Atmospheric Clayton, D. J.;

8 Transport and Dispersion Model Bixler, N. E.; June 2017 SAND2017-6184C Benchmarking Compton, K.

Software Regression Quality Eubanks, L. L. et 9 Assurance for MACCS2: Version July 2012 SAND2012-6333 al.

2.5.0.0 through Version 2.5.0.9 Effect on the Offsite Consequence of Transactions of the Korean the MACCS Non-Site-Specific Best Jin, D.-S.; Han, Nuclear Society Spring 10 May 2017 Modeling Practices Used in the US S.-K.; Lim, C.-K. Meeting, May 2017, Jeju, SOARCA Project Korea Atmospheric Transport and Dose 11 Calculations: Concepts and Bixler, N. E. July 2015 SAND2015-5545PE Implementation in the MACCS Code Level-3 Consequence Analysis Part 1 12 Bixler, N. E. April 2015 SAND2015-3244PE Atmospheric Transport and Dispersion A-12

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Emergency Planning Zones Estimation Science and Technology of for Karachi-2 and Karachi-3 Nuclear ahin, S. and Ali, 13 October 2016 Nuclear Installations, vol.

Power Plants using Gaussian Puff M.

2016 Model Avoidable Dose and Total Dose 14 Radiological Assessments in Support of Kraus, T. D. June 2013 SAND2013-5158C Public Protection Decisions Importance of Accounting for the 15 Partitioning of Iodine Released During Kraus, T. D. June 2013 SAND2013-4833C Nuclear Power Plant Accidents MACCS2 Consequence Analysis for Osborn, D. M. et 16 BWR Mark I and Mark II Filtered October 2012 SAND2012-9533 al.

Containment Venting Error Analysis of CM Data Products 17 Hunt, B. D. et al. March 2017 SAND2017-3298R Input Distributions Probabilistic Safety Effects of Source Term on Off-site Han, S.-J.; Kim, Assessment and 18 Consequence in LOCA Sequence in a June 2014 T.-W.; Ahn, K.-I. Management PSAM 12, June Typical PWR 2014, Honolulu, Hawaii Article, The NRC Staff is Calling for Nuclear News, vol. 55, no. 9, 19 Additional SOARCA Studies," in Late n/a August 2012

p. 175 News" section Article, Reactor accident study finds Nuclear News, vol. 55, no. 3, 20 essentially zero fatalities, in Power n/a March 2012

p. 25-26 section OECD/NEA Workshop on Best-Estimate Calculations of Implementation of Severe Unmitigated Severe Accidents in State- Schaperow, J. H.

21 October 2009 Accident Management of-the-Art Reactor Consequence et al.

Measures - Oct 2009 -

Analyses Switzerland Journal of Nuclear Science Consequence Analysis for Nuclear Kang, T. and Jae, 22 February 2017 and Technology, vol. 54, no.

Reactors, Yongbyon M.

2, p. 223-232 Applicability of 100 TBq Cesium 137 Journal of Nuclear Science Release into Environment as a Safety Silva, K. and 23 December 2015 and Technology, vol. 52, no.

Criterion for Consequence Assessment Okamoto, K.

12, p. 1530-1539 at Reactor Design Approval Stage A-13

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number OECD/NEA Workshop on Treatment of Accident Mitigation Implementation of Severe Schaperow, J. H.

24 Measures in State-of-the-Art Reactor October 2009 Accident Management et al.

Consequence Analyses Measures - Oct 2009 -

Switzerland Risk and regulatory considerations for Carless, T. S.;

small modular reactor emergency 25 Talabi, S. M.; January 2019 Energy, vol. 167, p. 740-756 planning zones based on passive Fischbeck, P. S.

decontamination potential Status of Practice for Level 3 26 OECD/NEA November 2018 NEA/CSNI/R(2018)1 Probabilistic Safety Assessments Andrews, N.; Fu, 27 Response Technical Tools User Guide C.; Kaberlein, A. August 2018 SAND2018-9593 M.

Technical Basis for the Containment NUREG-2206 Protection and Release Reduction 28 Barr, J. A. et al. March 2018 Rulemaking for Boiling Water Reactors ADAMS Accession No.

with Mark I and Mark II Containments ML18065A048 4th Arab Forum on the U.S. Nuclear Regulatory Commission's Prospects of Nuclear Power 29 State-of-the-Art Reactor Consequence Osborn, D. M. October 2017 Meeting, Amman, Jordan, Analyses (SOARCA) Project October 9-12, 2017 International Conference on Recommendations for Future Research Kampanart, S. Risk Analysis, Decision 30 on Nuclear Accident Consequence and Vechgama, July 2017 Analysis, and Security, Analysis W. Tsinghua University, Beijing, China, July 21-23, 2017 NUREG/BR-0527 MELCOR Accident Code System 31 Sharp, A. E. April 2017 (MACCS) ADAMS Accession No. ML17102A912 Dose assessment in level 3 PRA - a 32 Karanta, I. February 2017 VTT-R-00738-17 rewiew of recently used methods A Study of Base Technology of Korean 33 Specific Level 3 PSA Code Han, S. et al. February 2017 KAERI/RR--4232/2016 Development A-14

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Probabilistic risk assessment from Dvorzhak, A.; Reliability Engineering and potential exposures to the public 34 Mora, J. C.; August 2016 System Safety, vol.152, applied for innovative nuclear Robles, B. p.176-186 installations A-15

III. Accident-Tolerant Fuel Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Accident Tolerant Fuels (ATF) Coating and Cladding Thermal Hydraulic Wang, J.; Transactions of the American 1 Properties Evaluation by MELCOR YU Corradini, M. L.; June 2018 Nuclear Society, vol. 118, no.

1.8.6: Benchmark for SURRY Short Jo, H. 1, p. 1259-1262 Term Station Black Out Potential Recovery Actions from a Wang, J.; Jo, H.J.; Nuclear Technology, vol. 204, 2 Severe Accident in a PWR: MELCOR October 2018 Corradini, M. L. p. 1-14 Analysis of a Station Blackout Scenario Transactions of the American Estimation of Core Recovery Time with Gurgen, A. and 3 July 2017 Nuclear Society, vol. 116, TRACE Shirvan, K.

no.1, p. 422-425 State-of-the-Art Report on Light Water 4 OECD/NEA October 2018 NEA No. 7317 Reactor Accident-Tolerant Fuels Estimation of Coping Time in Gurgen, A. and Nuclear Engineering and 5 Pressurized Water Reactors for Near October 2018 Shirvan, K. Design, vol.337, p. 38-50 Term Accident Tolerant Fuel Claddings Development of Accident Tolerant Fuel Options for Near Term Applications:

Fuel Performance Modeling under Wang, J.; Transactions of the American 6 Transient/Severe Accidents by Corradini, M. L.; November 2016 Nuclear Society, vol. 115, no.

MELCOR: PART I: Benchmark for Haskin, T. 1, 1799-1802 SURRY: Short Term Station Black Out (STSBO)

Severe Accident Scoping Simulations 7 of Accident Tolerant Fuel Concepts for Robb, K. R. August 2015 ORNL/SPR-2015/347 BWRs Accident Tolerant Clad Material Nuclear Engineering and 8 Modeling by MELCOR: Benchmark for Wang, J. et al. March 2017 Design, vol. 313, p. 458-469 SURRY Short Term Station Black Out Accident-Tolerant Fuel Valuation: EPRI TR-3002015091 9 Hess, S. et al. March 2019 Safety and Economic Benefits (not publicly available) 16th International Topical Analysis of the FeCrAl Accident Meeting on Nuclear Reactor 10 Tolerant Fuel Concept Benefits during Robb, K. R. January 2015 Thermalhydraulics, Chicago, BWR Station Blackout Accidents IL, USA A-16

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number System Code Evaluation of Accident Transactions of the American Wu, X. and 11 Tolerant Claddings During BWR Station November 2018 Nuclear Society, vol. 119, no.

Shirvan, K.

Blackout Accident 1, p. 444-447 Plant-Level Scenario-Based Risk 12 Analysis for Enhanced Resilient PWR - Ma, Z. et al. September 2018 INL/EXT-18-51436-Rev000 SBO and LBLOCA State-of-the-Art Report on Light Water 13 OECD/NEA January 2018 NEA No. 7317 Reactor Accident-Tolerant Fuels Status Report on Activities of the Systems Assessment Task Force, Bragg-Sitton, S.

14 September 2017 INL/EXT-17-43389-Rev000 OECD-NEA Expert Group on Accident M.

Tolerant Fuels for LWRs Accident tolerant clad material Nuclear Engineering and 15 modeling by MELCOR: Benchmark for Wang, J. et al. March 2017 Design, vol. 313, p. 458-469 SURRY Short Term Station Black Out A-17

IV. Advanced Reactor Analysis Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Risk Management for Sodium Fast Denman, M. R. et 1 January 2015 SAND2015-0542 Reactors al.

Regulatory Technology Development Plan - Sodium Fast Reactor: Grabaskas, D. et 2 October 2016 ANL-ART-49-Vol1-Vol2 Mechanistic Source Term - Trial al.

Calculation Korean Nuclear Society Overview of Key Computer Codes for Chang, W.-P.; Lee 3 October 2016 Autumn Meeting, Gyeongju, the PGSFR Safety Analysis K.-L.; Yoo, J.

Korea, October 26-28, 2016 Advanced Small Modular Reactor 4 (SMR) Probabilistic Risk Assessment Smith, C. September 2013 INL/EXT-13-30170 (PRA) Technical Exchange Meeting Methodology for Establishing the NuScale TR-0915-17772-NP Technical Basis for Plume Exposure 5 NuScale Power December 2015 Emergency Planning Zones at NuScale ADAMS Accession No.

Small Modular Reactor Plant Sites ML15356A842 Proposed Methodology and Criteria for Establishing the Technical Basis for Nuclear Energy ADAMS Accession No.

6 December 2013 Small Modular Reactor Emergency Institute ML13364A345 Planning Zone (White Paper)

Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Denman, M. R. et 7 Analysis and Incident Management April 2015 SAND2015--2484 al.

Guidelines (Risk Management for Sodium Fast Reactors)

Feasibility and Safety Assessment for DOE/NEUP-11-3097 8 Advanced Reactor Concepts Using Klein, A. et al. January 2015 OSU-NE-VF-1204 Vented Fuel A Methodology for Accident Analysis of IEEE Transactions on Plasma Fusion Breeder Blankets and Its Panayotov, D. et 9 October 2016 Science, vol. 44, no. 10, p.

Application to Helium-Cooled Lead- al.

2511-2522 Lithium Blanket Design and Licensing Strategies for the Fluoride-Salt-Cooled, High- Scarlat, R. O. et Progress in Nuclear Energy, 10 November 2014 Temperature Reactor (FHR) al. vol. 77, p. 406-420 Technology A-18

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Fluoride-Salt-Cooled, High-Temperature Reactor (FHR)

Subsystems Definition, Functional 11 Allen, T.R. et al. August 2013 n/a Requirement Definition, and Licensing Basis Event (LBE) Identification White Paper Regulatory cross-cutting topics for fuel Denman, M. R. et 12 October 2013 SAND2013-9367 cycle facilities al.

Advanced Reactor Safety Program - Szilard, R. H. and 13 August 2014 INL/EXT-14-32928 Stakeholder Interaction and Feedback Smith, C. L.

Final White Paper on Small Advanced ADAMS Accession No.

14 Reactor Design Reviews Rev 7 Holahan, G. M. May 2017 ML17213A849 (not publicly 2017/05/08 available)

Clark, A. J.;

Mechanistic Source Term Modeling for 15 Denman, M. R.; January 2017 SAND2017-0454C Sodium Fast Reactors Grabaskas, D.

A-19

V. Emergency Preparedness and Response Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number ADAMS Accession No.

1 RASCAL 4.3 Kowalczik, J. April 2014 ML14112A349 Ramsdell, Jr, J.V.;

RASCAL 4.3: Description of Models 2 Athey, G. F.; May 2015 NUREG-1940, Supplement 1 and Methods Rishel, J. P.

International Actions to Protect the Public in an Atomic Energy Emergency due to Severe Conditions Agency, Incident 3 May 2013 EPR-NPP-PPA-2013 at a Light Water Reactor. Date and Emergency Effective: May 2013 Centre, Vienna (Austria)

International Operational Intervention Levels for Atomic Energy 4 Reactor Emergencies and Methodology Agency, Incident March 2017 EPR-NPP-OILS-2017 for Their Derivation. March 2017 and Emergency Centre Reliability Engineering &

Alternative evacuation strategies for Hammond, G. D.

5 March 2015 System Safety, vol. 135, p. 9-nuclear power accidents and Bier, V. M.

14 Significance Quantification Process for 6 Jones, J. A. January 2014 SAND2014-0677C Emergency Preparedness Oversight Emergency Preparedness Significance 7 Quantification Process: Proof of Sullivan, R. et al. June 2013 INL/LTD-12-27648, Rev. 1 Concept 8 Emergency Preparedness Osborn, D. June 2015 SAND2015-4657PE OECD/NEA Committee on 9 Inspection of Emergency Arrangements Nuclear December 2013 NEA-CNRA-R-2013-2 Regulatory Activities WGIP Emergency Preparedness. NRC Needs Government to Better Understand Likely Public 10 Accountability March 2013 GAO-13-243 Response to Radiological Incidents at Office Nuclear Power Plants Transactions of the American Rethinking Nuclear Emergency Goble, R. and 11 November 2013 Nuclear Society vol. 109 no.

Planning, Preparations, and Response Bier, V.

1, p. 1959-1961 A-20

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Nuclear Engineering and A new approach to quantify safety Kim, I. S.; Choi, 12 October 2017 Technology, vol. 49, no. 7 p.

benefits of disaster robots Y.; Jeong, K. M.

1414-1422 The Risk of Extended Power Loss and ASME Journal of Nuclear the Probability of Emergency 13 Duffey, R. B. May 2019 Engineering and Radiation Restoration for Severe Events and Science, vol. 5, no. 3, 031601 Nuclear Accidents About the development of a national Mikailova, R. A. IOP Conference Series:

14 system of response to the nuclear and Shubina, O. March 2019 Material Science and emergency for agriculture A. Engineering 487 012013 GISBased Integration of Social Vulnerability and Level 3 Probabilistic Risk Assessment to Advance Risk Analysis, vol. 39, no. 6, 15 Pence, J. et al. November 2018 Emergency Preparedness, Planning, p. 1262-1280 and Response for Severe Nuclear Power Plant Accidents A-21

VI. Dynamic PRA Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Risk-Informed Safety Margin 1 Characterization Methods Development Smith, C. L. et al. September 2014 INL/EXT-14-33191 Work System Theoretic Frameworks for Williams, A. D. et 2 Mitigating Risk Complexity in the September 2017 SAND2017-10243 al.

Nuclear Fuel Cycle Safety Relief Valve Cyclic Failure 3 Analysis for Use in Discrete Dynamic Denman, M. R. May 2013 SAND2013-3684C Event Trees Seamless Level 2/Level 3 Probabilistic 4 Risk Assessment Using Dynamic Event Osborn, D. M. January 2013 n/a Tree Analysis (Dissertation)

Development and Application of a LaChance, J. L. et 5 Dynamic Level 1 and 2 Probabilistic January 2012 SAND2012-0651C al.

Safety Assessment Tool Nuclear Power Plant Cyber Security Wheeler, T. A. et 6 Discrete Dynamic Event Tree Analysis September 2017 SAND2017-10307 al.

(LDRD 17-0958) FY17 Report Dynamical systems probabilistic risk Denman, M. R.

7 March 2014 SAND2014-4037 assessment and Ames, A. L.

Safety Relief Valve Cyclic Failure 8 Analysis for use in Discrete Dynamic Denman, M. R. September 2013 SAND2013-8217C Event Trees (slidedeck)

Discrete Dynamic Event Tree Analysis Denman, M. R. et 9 of Small Modular Reactor Severe September 2013 SAND2013-8096C al.

Accident Management (slidedeck)

Discrete Dynamic Event Tree Analysis Denman, M. R. et 10 of Small Modular Reactor Severe May 2013 SAND2013-3680C al.

Accident Management The Assessment of Low Probability 11 Containment Failure Modes using Brunett, A. J. January 2013 n/a Dynamic PRA (Dissertation)

A-22

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number SAND2012-9346 Discrete Dynamic Probabilistic Risk LaChance, J. et 12 Assessment Model Development and October 2012 al. ADAMS Accession No.

Application ML12305A351 An Assessment of Low Probability Brunett, A.; Transactions of the American 13 Containment Failure in a Long-Term Denning, R.; November 2013 Nuclear Society, vol. 109, no.

Station Blackout Using Dynamic PRA Aldemir, T. 1, 954-957 Preliminary Cyber-Informed Dynamic Transactions of the American Branch Conditions for Analysis with the Denman, M. R. et 14 November 2016 Nuclear Society, vol. 115, no.

Dynamic Simplified Cyber MELCOR al.

1, 787-790 Model Jankovsky, Z. K.;

15 How to ADAPT Haskin, T. C.; June 2018 SAND2018-6660 Denman, M. R.

Uncertainty Quantification for External Transactions of the American 16 Events Analysis of LWRS/RISMC Parisi, C. et. al. June 2017 Nuclear Society, vol. 116, no.

Project 1, 795-797 Reliability Engineering &

Quantitative risk reduction by means of 17 París, C. et al. February 2019 System Safety Meeting, vol.

recovery strategies 182, p. 13-32 Chapter 5 of Advanced Level 2 Probabilistic Risk Assessment Concepts in Nuclear Energy 18 Osborn, D. et al. June 2018 Using Dynamic Event Tree Analysis Risk Assessment and Management Development of Computational and Data Processing Tools for ADAPT to 19 Jankovsky, Z. K. January 2018 Dissertation Assist Dynamic Probabilistic Risk Assessment A Dynamic Assessment of an Jankovsky, Z. K. ;

20 Interfacing System Loss of Coolant Denman, M. R. ; September 2017 SAND2017-10141C Accident Aldemir, T.

International Conference on Topical Issues in Nuclear A Dynamic Assessment of Auxiliary Jankovsky, Z. ; Installation Safety:

Building Contamination and Failure 21 Denman, M. R. ; June 2017 Safety Demonstration of Due to a Cyber-Induced Interfacing Aldemir, T. Advanced Water Cooled System Loss of Coolant Accident Nuclear Power Plants, VIC, Vienna, 6-9 June 2017 A-23

VII. Multi-Unit PRA Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Transactions of the American Construction of Multi-Path Event 1 Kim, S. et al. October 2017 Nuclear Society, vol. 117, no.

Tree for Station Blackout Events 1, p. 951-953 Multi-Unit Level 3 Probabilistic Nuclear Engineering and Safety Assessment: Approaches 2 Kim, S.-Y. et al. December 2018 Technology, vol. 50, no. 8, p.

and Their Application to a Six-Unit 1246-1254 Nuclear Power Plant Site Multi-Unit Accident Contributions to Hudson, D. W. and Nuclear Technology, vol. 197, 3 Quantitative Health Objectives: A March 2017 Modarres, M. no. 3, p. 227-247 Safety Goal Policy Analysis Nuclear Engineering and Holistic Approach to Multi-Unit Site Kim, I. S.; Jang, M.;

4 March 2017 Technology, vol. 49, no. 2, p.

Risk Assessment: Status and Issues Kim, S. R.

286-294 Muhlheim, M. D.;

Initiating Events for Multi-Reactor 5 Flanagan, G. F.; September 2014 ORNL/TM-2014/533 Plant Sites Poore, III, W. P.

Probabilistic Safety Scoping Estimates of Multiunit Assessment and 6 Stutzke, M. A. June 2014 Accident Risk Management PSAM 12, Honolulu, Hawaii, June 2014 A Methodology for Performing 7 Consequence Analysis for Multi- Bixler, N. E. November 2014 SAND2014-20038C Unit/Spent Fuel Pool Source Terms Multi-Unit Accident Contributions to US Nuclear Regulatory Commission Quantitative Health Objectives: A 8 Safety Goal Policy Analysis using Hudson, D. W. January 2016 n/a Models from State-of-the-Art Reactor Consequence Analyses (Dissertation)

Reliability Engineering &

9 Multi-Unit Dynamic PRA Mandelli, D. et al. May 2019 System Safety, vol. 185, p.

303-317 Multi-Unit Level 3 Probabilistic Nuclear Engineering and Safety Assessment: Approaches 10 Kim, S.-Y. et al. December 2018 Technology, vol. 50, no. 8, p.

and Their Application to a Six-Unit 1246-1254 Nuclear Power Plant Site A-24

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number The Current Research Status and 25th International Conference Technical Development Framework on Nuclear Engineering 11 Ding, H. et al. July 2017 of Multi-Reactor Probabilistic (ICONE25), Shanghai, China, Consequence Assessment July 2-6, 2017 A-25

VIII. Fukushima Forensic Analysis Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number MELCOR Applications to SOARCA and 1 Gauntt, R. O. March 2014 SAND2014-1879C Fukushima A MELCOR Model of Fukushima Annals of Nuclear Energy, 2 Sevón, T. November 2015 Daiichi Unit 1 Accident vol. 85, p. 1-11 A MELCOR Model of Fukushima Nuclear Engineering and 3 Sevón, T. April 2015 Daiichi Unit 3 Accident Design, vol. 284, p. 80-90 A Review of Recent SNL MELCOR 4 Kalinich, D. A. March 2015 SAND2015-2172PE Fukushima Accident Analyses Overview of Sandia National 5 Laboratories MELCOR Fukushima Kalinich, D. A. July 2015 SAND2015-6035PE Analyses Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Gauntt, R. O. and 6 January 2016 SAND2016-0232 Implications for Decommissioning of Mattie, P. D.

Fukushima Reactors - Volume I Robb, K. R.;

Insight from Fukushima Daiichi Unit 3 Nuclear Technology, vol. 186, 7 Francis, M. W.; May 2014 Investigations Using MELCOR no. 2, p. 145-160 Ott, L. J.

MELCOR Simulations of the Severe Nuclear Technology, vol. 186, 8 Accident at the Fukushima Daiichi Unit Gauntt, R. et al. May 2014 no. 2, p. 161-178 1 Reactor MELCOR Simulations of the Severe Nuclear Technology, vol. 186, 9 Cardoni, J. et al. May 2014 Accident at Fukushima Daiichi Unit 3 no. 2, p. 179-197 Transactions of the American Sensitivity Study of 1F1 Type Accident Saito, K. and 10 October 2015 Nuclear Society, vol. 113, no.

by MELCOR Code Yamaji, A.

1, p. 1411-1414 Fukushima Daiichi unit 1 Uncertainty Analysis--Preliminary Selection of Cardoni, J. N. and 11 February 2014 SAND2014-1170 Uncertain Parameters and Analysis Kalinich, D. A.

Methodology Seismically-Induced Reactor Coolant Leakage as an Allegedly-Possible Kukita, Y. and JAEA-TECHNOLOGY--2014-12 Cause of Accident at Unit 1 of November 2014 Watanabe, N. 036 Fukushima Daiichi Nuclear Power Station MELCOR 2.1 Simulations of 13 Cardoni, J. N. November 2011 SAND2012-9742C Fukushima Unit 3 A-26

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Severe Accident Progression Analyses 14 in SOARCA and Comparisons to Gauntt, R. O. July 2013 SAND2013-6058C Fukushima Fukushima Daiichi - A Case Study for BWR Instrumentation and Control Clayton, D. A. and 15 April 2013 ORNL/TM-2013/154 Systems Performance during a Severe Poore, III, W. P.

Accident Rev 0 Fukushima Daiichi Nuclear Plant Journal of Environmental 16 Accident: Atmospheric and Oceanic Hirose, K. June 2016 Radioactivity, vol. 157, p.

Impacts over the Five Years 113-130 Analysis of the Accident in the Fernandez-Annals of Nuclear Energy, 17 Fukushima Daiichi Nuclear Power Moguel, L. and September 2015 vol. 83, p. 193-215 Station Unit 3 with MELCOR_2.1 Birchley, J.

Fukushima Daiichi - A Case Study for BWR Instrumentation and Control Clayton, D. A. and 18 June 2014 ORNL/TM-2013/154 R1 Systems Performance during a Severe Poore, III, W. P.

Accident Rev 1 ATHLET-CD/COCOSYS Analyses of Severe Accidents in Fukushima Daiichi Sonnenkalb, M. Nuclear Technology, vol. 196, 19 November 2016 Units 2 and 3: German Contribution to and Band, S. no. 2, p. 211-222 the OECD/NEA BSAF Project, Phase 1 Fukushima Daiichi Unit 1 Uncertainty Analysis-Exploration of Core Melt Denman, M. R.

20 August 2015 SAND2015-6612 Progression Uncertain Parameters- and Brooks, D. M.

Volume II Presentation of Fukushima Analyses to Osborn, D.;

21 U.S. Nuclear Power Plant Simulator Kalinich, D. A.; February 2015 SAND2015-1177R Operators and Vendors Cardoni, J. N.

Insights Gained from Forensic Analysis Andrews, N. C.;

22 with MELCOR of the Fukushima-Daiichi October 2017 SAND2017-10811R Gauntt, R. O.

Accidents Fukushima Daiichi Accident Study: Gauntt, R. O. et 23 July 2012 SAND2012-6173 Status as of April 2012 al.

Historical Overview of Fukushima ADAMS Accession No.

24 Kelly, J. E. March 2018 Forensics Work ML18090A006 A-27

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number PSAM Topical Conference on Team Performance Comparison in Human Reliability, Core-Melt Units of Fukushima Daiichi Petkov, G. and 25 June 2017 Quantitative Human Factors, NPS Based on Dynamic Context Petkov, I.

and Risk Management, Quantification of Accident Munich, Germany, June 2017 The 8th European Review Studies on the Recriticality Potential Darnowskia, P.; Meeting on Severe Accident 26 during Fukushima Unit-3 Core Potpaczyka, K.; May 2017 Research - ERMSAR-2017 Reflooding wirski, K. Warsaw, Poland, 16-18 May 2017 4th Meeting of the OECD/NEA BSAF Project 27 Fukushima Reactor Building Model Andrews, N. et al. January 2017 Phase 2, Paris, France, January 9-13, 2017 A-28

IX. Fukushima Lessons Learned Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number 1 Nuclear Power Plant Severe Accidents Osborn, D. March 2015 SAND2015-1591PE Health Physics: Radiation-Generating 2 Devices, Characteristics, and Hazards Bevelacqua, J. J. April 2016 n/a (Ch. 7: Regulatory Considerations)

Nuclear Regulation in the United States International Nuclear Safety 3 and a Possible Framework for an Bevelacqua, J. J. February 2013 Journal, vol. 2, no. 1, p. 52-57 International Regulatory Approach Lessons Learned from the Fukushima National Research Contract No. NRC-HQ-12-G-4 Nuclear Accident for Improving Safety January 2014 Council 03-0002 of U.S. Nuclear Plants Lessons Learned from the Fukushima Accident for Improving Safety and National Research Contract No. NRC-HQ-12-G-5 January 2016 Security of U.S. Nuclear Plants. Phase Council 03-0002 2

Environmental Quality The Nuclear Regulatory Commission 6 Eccleston, C. H. September 2012 Management, vol. 22, no.1, p.

and NEPA Review 43-58 7 Fukushima Lessons Learned Gauntt, R. O. February 2012 SAND2012-0893C The Canary, the Ostrich, and the Black Swan: A Historical Perspective on our Nuclear Technology, vol. 186, 8 Greene, S. R. May 2014 Understanding of BWR Severe no. 2, p. 115-138 Accidents and Their Mitigation 9 SNL BSAF Phase 2 Activities Andrews, N. et al. December 2015 SAND2015-10448PE A-29

X. Severe Accident Management Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Scoping Study Investigating PWR Rempe, J. L.;

1 Instrumentation during a Severe Knudson, D. L.; September 2015 INL/EXT-15-35940 Accident Scenario Lutz, R. J.

Technical Basis for Severe Accident 2 Duvall, A. et al. April 2015 EPRI 3002003301 Mitigating Strategies: Volume 1 Verification of SAMGs in SBO Annals of Nuclear Energy, 3 Sequences with Seal LOCA. Multiple Queral, C. et al. December 2016 vol. 98, p.90-111 Damage Domains Preliminary Assessment for the Proceedings of the 2013 fall Park, S. Y. and 4 Mitigative Effectiveness of External October 2013 Korean Nuclear Society Ahn, K. I.

Injection during Extended SBO meeting The Relative Importance of Mitigation, Transactions of the American 5 Early Phase, Intermediate Phase, and Denning, R. et al. November 2013 Nuclear Society, vol. 109, no.

Late Phase Response 1 External Cooling of the BWR Mark I and II Drywell Head as a Potential 6 Robb, K. R. July 2018 ORNL/TM-2018/901 Accident Mitigation Measure -

Expanded Scoping Assessment Key Parameters for Operator Diagnosis Clayton, D. A. and 7 of BWR Plant Condition during a January 2015 ORNL/LTR-2014/320 Poore, III, W. P.

Severe Accident Evaluation of an Accident Management Nuclear Engineering and Strategy of Emergency Water Injection Park, S.-Y. and 8 October 2015 Technology, vol. 47, no. 6, p.

using Fire Engines in a Typical Ahn, K.-I.

719-728 Pressurized Water Reactor Post-Severe Accident Environmental Conditions for Essential Clayton, D. and 9 September 2015 ORNL/TM-2015/278 Instrumentation for Boiling Water Poore, M.

Reactors ADAMS Accession No. ML12243A196 Regulatory Perspective and Accident Management Procedure Influences on Rahn, D. L. and 10 September 2012 IAEA/JNESO Workshop on Accident Monitoring Instrumentation Cowdrey, C. B.

Accident Monitoring Design Criteria Instrumentation, Tokyo, Japan, September 4, 2012 A-30

Probabilistic Safety Dealing with Beyond-Design-Basis Assessment and 11 Nourbakhsh, H. P. June 2014 Accidents in Nuclear Safety Decisions Management PSAM 12, Honolulu, Hawaii, June 2014 iROCS: Integrated Accident Management Framework for Coping Nuclear Engineering and 12 Kim, J. et al. March 2016 with Beyond-Design-Basis External Design, vol. 298, p. 1-13 Events APR1400 Design Certification Severe Accident Mitigation Design Alternatives ADAMS Accession No.

13 Palmrose, D. E. September 2018 Technical Analysis in Support of the ML18096A697 Environmental Assessment 26th International Conference FLEX Strategy Implementation for Ruiz-Zapatero, Nuclear Energy for New 14 LOCA Sequences in PWR- M.; Bocanegra, September 2017 Europe, Bled, Slovenia, Westinghouse R.; Queral, C.

September 11-14, 2017 Dynamic Human Performance Context 2017 European Safety and Comparison for Severe Accident Petkov, G. and Reliability Conference 15 June 2017 Management during Long Term Station Petkov, I. (ESREL2017), Portoroz, Blackout in Light Water Reactors Slovenia, June 18-22, 2017 Development of Site Risk Assessment 16 and Management Technology including Yang, J. E. et al. March 2017 KAERI/RR--4225/2016 Extreme External Events A-31

XI. Level 1 PRA Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Use and Development of Probabilistic Nuclear Energy 1 December 2012 NEA/CSNI/R(2012)11 Safety Assessment Agency The NRC's SPAR Models: Current 2 Status, Future Development, and Buell, R. F. September 2008 INL/CON-08-14484 Modeling Issues SPAR Integrated Capabilities Model INL/LTD-12-27648, Rev. 1 3 Ma, Z. et al. July 2013 (SPAR-ICM) Project (not publicly available)

Helton, D.;

U.S. NRC Confirmatory Level 1 PRA 4 Esmaili, H.; Buell, March 2011 INL/CON-11-21026 Success Criteria Activities R.

Development Strategy of Optimized Level 1&2 PSA Model for APR1400 Hwang, S.-W. et Annals of Nuclear Energy, 5 December 2018 NPPs Reflecting New Severe Accident al. vol. 122, p. 256-269 Mitigating System and Its Application Reliability Engineering &

Lessons Learned from Applying a New Taylor, C.; Øie, S.;

6 Oct 2018 System Safety, available HRA Method for the Petroleum Industry Gould, K.

online 5 October 2018 A-32

XII. Nuclear Safety/Societal Risk Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Insights on Risk Margins at Nuclear Power Plants: A Technical Evaluation of Margins in Relation to Electric Power 1 May 2018 EPRI Report No. 3002012967 Quantitative Health Objectives and Research Institute Subsidiary Risk Goals in the United States The Three Mile Island, Chernobyl, International Nuclear Safety 2 and Fukushima Daiichi Accidents Bevelacqua, J. J. June 2016 Journal, vol. 5, no.1, p. 21-79 and Their Radiological Impacts ADAMS Accession No. ML15253A634 Software Verification and Validation:

3 Siu, N. September 2015 Examples from the Safety Arena INMM Workshop on VA Tools, Boston, MA; September 14-16, 2015 Development of an Updated 4 Societal-Risk Goal for Nuclear Bier, B. et al. July 2014 INL/CON-13-30391 Power Safety Transactions of the American The Societal Risk of Severe Denning, R. and 5 June 2013 Nuclear Society, vol. 108, no.

Accidents in Nuclear Power Plants McGhee, S.

1, p. 521-525 Insights into the Societal Risk of Denning, R. and Risk Analysis, vol. 37, no. 1, 6 January 2017 Nuclear Power Plant Accidents Mubayi, V. p. 160-172 ICAPP '12: 2012 International Congress on Advances in Understanding the Nature of 7 Denning, R. S. June 2012 Nuclear Power Plants, Nuclear Power Plant Risk Chicago, IL (United States),

June 24-28, 2012 Critical Reflections on Nuclear and Renewable Energy: Environmental Protection and Safety in the Wake 8 Kuo, W. March 2014 n/a (Book) of the Fukushima Nuclear Accident.

(Chapter 1: Reliability and Nuclear Power and Postscript)

Preliminary Study for Development Transactions of the American of Technical Basis of Quantitative 9 Shehhi, O. A. A. et al. June 2017 Nuclear Society, vol. 116, no.

Safety Goals for NPPs Operations 1, p. 801-803 in UAE A-33

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Impact of Probabilistic Risk Denning, R. S. and Progress in Nuclear Energy, 10 Assessment and Severe Accident January 2018 Budnitz, R. J. v. 102 p.90-102 Research in Reducing Reactor Risk Nuclear Weapon Pit Production:

11 Options to Help Meet A Medalia, J. E. May 2015 CRS Report No. R44033 Congressional Requirement Nuclear Renaissance, Public Goodfellow, M. J.;

Energy Policy, vol. 39, no. 10, 12 Perception and Design Criteria: An Williams, H. R.; October 2011

p. 6199-6210 Exploratory Review Azapagic, A.

IEEE Transactions on 13 Reliability and Nuclear Power Kuo, W. June 2011 Reliability, vol. 60 no. 2, p.

365-367 International Conference on Topical Issues in Nuclear Installation Safety:

International Atomic Defence in Depth Advances and Energy Agency, Challenges for Nuclear Installation 14 Safety Assessment October 2014 IAEA-TECDOC-CD--1749 Safety. Proceedings of an Section, Vienna International Conference held in (Austria)

Vienna, Austria, 21-24 October 2013 Handbook of Advanced Nuclear 15 Hino, R. et al. January 2017 JAEA-REVIEW-2016-038 Hydrogen Safety. 1st edition Nuclear Facility Accident (NFAC)

Lee, R. W. and 16 Unit Test Report for HPAC Version March 2017 ORNL/TM-2017/94 Sulfredge, C. D.

6.4 Nuclear Energy: Overview of 17 Congressional Issues. Updated Holt, M. November 2018 CRS Report No. R42853 November 16, 2018 Environmental Impact Statement for NUREG-2105 v.3 the Combined License (COL) for NRC and U.S. Army 18 January 2013 Enrico Fermi Unit 3. Final Report. Corps of Engineers ADAMS Accession No.

Appendix E ML12307A177 Comparative Radioecological Assessment of Serious-Accident Spiridonov, S. and Atomic Energy, vol.125, no.

19 January 2019 Scenarios in NPP on the Basis of Mikailova, R. 3, p. 198-203 the Risk for Natural Communities A-34

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number System Studies for Global Nuclear Assurance & Security: 3S Risk 20 Analysis for Small Modular Reactors Williams, A. D. et al. October 2018 SAND2018-12447 (Volume I) - Technical Evaluation of Safety Safeguards & Security Integrating Thermal Energy Storage 21 and Nuclear Reactors: A Technical Abel, C. R. May 2018 Dissertation and Policy Study Irradiation Dose of the Woody Tier Mikailova, R. A. and Atomic Energy, vol. 123, no.

22 of a Coniferous Forest Due to January 2018 Spiridonov, S. I. 3, p. 202-208 Accidental Emissions from NPP Uncertainties in Estimating Health Journal of Radiological 23 Risks Associated with Exposure to Preston, R. J. et al. June 2013 Protection, vol. 33, no. 3, p.

Ionising Radiation 573-588 A-35

XIII. Sensitivity and Uncertainty Analyses Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Proposed Analyses for xLPR 2.0 Sallaberry, C. J.-

1 June 2015 SAND2015-5118PE Outputs M. et al.

Bootstrapped-Ensemble-Based Sensitivity Analysis of a Trace Thermal- Reliability Engineering &

2 Hydraulic Model Based on a Limited Di Maio, F. et al. September 2016 System Safety, vol. 153, p.

Number of PWR Large Break LOCA 122-134 Simulations Probabilistic Safety SOARCA Surry Power Station Assessment and 3 Uncertainty Analysis: Parameter Jones, J. et al. June 2014 Management PSAM 12, Methodology and Insights Honolulu, Hawaii, June 2014 SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout 4 Osborn, D. et al. April 2013 SAND2013-3107C Uncertainty Analysis MELCOR Parameters and Probabilistic Results SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout 5 Osborn, D. et al. April 2013 SAND2013-3102C Uncertainty Analysis MACCS2 Aleatory Weather Effects SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout 6 Osborn, D. et al. April 2013 SAND2013-3105C Uncertainty Analysis MACCS2 Dose Truncation Sensitivity SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout 7 Osborn, D. et al. April 2013 SAND2013-3106C Uncertainty Analysis MACCS2 Parameters and Probabilistic Results SOARCA Peach Bottom Atomic Power Probabilistic Safety Station Long-Term Station Blackout Assessment and 8 Bixler, N. E. et al. June 2014 Uncertainty Analysis: Contributions to Management PSAM 12, Overall Uncertainty Honolulu, Hawaii, June 2014 SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Gauntt, R. O. et 9 February 2014 SAND2014-1344C Uncertainty Analysis: Knowledge al.

Advancement A-36

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout 10 Uncertainty Analysis Probabilistic Osborn, D. et al. April 2013 SAND2013-3104C Methodology and Regression Technique SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout 11 Bixler, N. E. et al. February 2014 SAND2014-1346C Uncertainty Analysis: Convergence of the Uncertainty Results Uncertainty Analysis of Consequence 12 Hunt, B. D. et al. January 2018 SAND2018-0329 Management (CM) Data Products Phenomenological Uncertainties in 13 Gauntt, R. O. March 2011 SAND2011-2024C SOARCA Study The SOARCA Surry Power Station Transactions of the American Short-Term Station Blackout 14 Bixler, N. E. et al. June 2015 Nuclear Society, vol. 112, no.

Uncertainty Analysis: MACCS 1, p. 515-516 Parameter Development A Systematic Framework for Effective Uncertainty Assessment of Severe Reliability Engineering &

Hoseyni, S. M. et 15 Accident Calculations; Hybrid May 2014 System Safety, vol. 125, p.

al.

Qualitative and Quantitative 22-35 Methodology Sensitivity Analysis of Successful Transactions of the American Kim, S.-Y. and 16 Evacuee Proportion in Hypothetical June 2017 Nuclear Society, vol. 116, no.

Lim, H.-G.

NPP Accident by Earthquake 1, p. 808-811 Development of a Technical Basis for ADAMS Accession No.

Raynaud, P. and 17 Component Integrity Assessment Using August 2018 ML18235A101 (*not publicly Kirk, M.

Probabilistic Methods available)

Probabilistic Safety Assessment and Sequoyah SOARCA Uncertainty 18 Bixler, N. E. et al. March 2018 Management Conference Analysis of a STSBO Accident (PSAM 14), Los Angeles, CA, September 16-21, 2018 Using Deterministic and Probabilistic ADAMS Accession No.

19 Methods for MELCOR Severe Accident Mattie, P. D. et al. October 2017 ML17278A919 Uncertainty Analysis A-37

Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Findings from Uncertainty Studies Evaluating Severe Accident ADAMS Accession No.

20 Ghosh, S. T. et al. October 2017 Phenomena and Off-Site ML17278A920 Consequences ASME 2017 Pressure Vessels and Piping Conference, Sensitivity Analysis for XLPR Sallaberry, C. J. et 21 July 2017 Volume 6B: Materials and Acceptance Testing al.

Fabrication, Waikoloa, Hawaii, USA, July 16-20, 2017 A-38

XIV. Storage and Transportation Safety Journal/Technical

1. Title Author Date Report/Conference Citation or Report Number Example of Integration of Safety Security and Safeguard Using Dynamic Kalinina, E. A. et 1 April 2017 SAND2017-3572C Probabilistic Risk Assessment Under a al.

System-Theoretic Framework Advancing US Public Acceptance of Packaging, Transportation, Spent Fuel Storage and Transport: Storage & Security of 2 Pennington, C. W. 2013 Proposed Outreach Services for Radioactive Material, vol. 24, Ionising Radiation Education Support No. 3, p.95-107 COMSECY 13-0030, Regulatory Analysis for Japan Lessons- Enclosure 1 3 Learned Tier 3 Issue on Expedited U.S. NRC November 2013 Transfer of Spent Fuel ADAMS Accession No. ML13273A628 Analysis of Dose Consequences Durbin, S. G. and 4 Arising from the Release of Spent January 2013 SAND2013-0533 Morrow, C.

Nuclear Fuel from Dry Storage Casks NUREG-2161 Consequence Study of a Beyond-Design-Basis Earthquake Affecting the 5 Barto, A. et al. June 2013 ADAMS Accession No.

Spent Fuel Pool for a US Mark I Boiling ML14255A365 Water Reactor A-39