ML20100J883

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

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

PROJECT Research Information Letter Office of Nuclear Regulatory Research Date Published: March 2020 Prepared by:

C. Leggett Reactor Systems Engineer Division of Systems Analysis 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.

iii 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.

iv 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.

v 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

vi 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

1 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.

2 (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 case2, 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.

3 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 deterministic3 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 report4, 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.

4 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 NUREG-1935 (2012)

State-of-the-Art Reactor Consequence Analyses (SOARCA) (Parts 1 and 2)

BWR-4 (Mark I containment), 3-loop PWR (large dry containment)

STSBO, LTSBO, thermally induced SGTR, interfacing systems loss of coolant accident*

NUREG/BR-

0359, Revision 2 (2016)

Modeling Potential Reactor Accident Consequences BWR-4 (Mark I containment), 3-loop PWR (large dry containment),

4-loop PWR (ice condenser containment)

STSBO, LTSBO, thermally induced SGTR, interfacing systems loss of coolant accident NUREG/CR-7110, vol. 1, Revision 1 (2013)

State-of-the-Art Reactor Consequence Analyses Project:

Volume 1: Peach Bottom Integrated Analysis BWR-4 (Mark I containment)

STSBO, LTSBO*

NUREG/CR-7110, vol. 2, Revision 1 (2013)

State-of-the-Art Reactor Consequence Analyses Project:

Volume 2: Surry Integrated Analysis 3-loop PWR (large dry containment)

STSBO, LTSBO, thermally induced SGTR, interfacing systems loss of coolant accident NUREG/CR-7155 (2016)

State-of-the-Art Reactor Consequence Analyses Project:

Uncertainty Analysis of the Unmitigated Long-Term Station Blackout of the Peach Bottom Atomic Power Station BWR-4 (Mark I Containment)

STSBO NUREG/CR-7245 (2019)

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Sequoyah Integrated Deterministic and Uncertainty Analyses 4-loop PWR (ice condenser containment)

STSBO, LTSBO NUREG/CR-7262 (Draft)

State-of-the-Art Reactor Consequence Analyses Project:

Uncertainty Analysis of the Unmitigated Short-Term Station Blackout of the Surry Power Station 3-loop PWR (large dry containment)

STSBO

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

5 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

6 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 recommendation5 related to reliable hardened vents for BWR Mark I and Mark II containments and Tier 3 recommendations6 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.

7 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.

8 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

9 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.

10 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.

11 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

12 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.

13 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

14 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.

15 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

16 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

17 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

18 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,

19 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.

20 4

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25

A-1 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-2 I.

Severe Accident Progression Analysis Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Heat Up and Failure of BWR Upper Internals During a Severe Accident Robb, K. R.

April 2017 Nuclear Engineering and Design, vol. 314, p.293-306 2

Terry Turbopump Analytical Modeling Efforts in Fiscal Year 2016 - Progress Report Ross, K.; Cardoni, J. N.; Osborn, D.

April 2018 SAND2018-4337 3

Terry Turbopump Expanded Operating Band Osborn, D.

July 2017 SAND2017-6715C 4

Consequences of Degraded Containment in a Severe Nuclear Power Plant Accident Jankovsky, Z.;

Jones, C.;

Kalinich, D.

November 2014 Transactions of the American Nuclear Society, vol. 111, no.

1 (2014 ANS Winter Meeting) 5 MELCOR Code Source Term Characteristics for Fast SBO Scenario of OPR1000 Plant Han, S. J. et al.

October 2012 Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, October 25-26, 2012 6

Effects of Source Term Characteristics on Off-Site Consequence Han, S. J.; Ahn, K.

I.

October 2012 Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, October 25-26, 2012 7

Development of the SharkFin Distribution for Fuel Lifetime Estimates in Severe Accident Codes Denman, M. R.

November 2016 Transactions of the American Nuclear Society, vol. 115, no.

1 (2016 ANS Winter Meeting) 8 Simplified Method for Assessing the Risk Associated with Consequential Steam Generator Tube Rupture Events Azarm, M. A. et al.

September 2013 ANS PSA 2013 International Topical Meeting on Probabilistic Safety Assessment and Analysis, Columbia, SC, September 22-26, 2013 9

Models and Methods Related to Severe Accidents and Source Term Evaluation Gauntt, R. O.

April 2015 SAND2015-3101PE 10 Comparative Analysis on the Influence of the MAAP4 Phenomenological Model Parameters for the Severe Accident Source Term for Different Plant Designs and Accident Scenarios Kang, S. W. and Yim, M.-S.

July 2018 Annals of Nuclear Energy, vol. 117, p.98-108 11 MAAP-MELCOR Crosswalk Phase 1 Study Luxat, D. L. et al.

December 2016 Nuclear Technology, vol. 196,

p. 684-697

A-3 Title Author Date Journal/Technical Report/Conference Citation or Report Number 12 Accident Consequences and Analysis Gauntt, R. O.

December 2017 International Atomic Energy Agency Workshop on Severe Accident Management Guidelines, December 11-15, 2017 13 Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis Farmer, M. T. et al.

January 2015 ANL/NE-15/4 14 A Simple Assessment Scheme for Severe Accident Consequences using Release Parameters Silva, K. and Okamoto, K.

August 2016 Nuclear Engineering and Design, vol. 305, p. 688-696 15 Station Blackout Mitigation Strategies Analysis for Maanshan PWR Plant using TRACE Lin, H.-T. et al.

March 2016 Annals of Nuclear Energy, vol. 89, p. 1-18 16 External Flooding Event Analysis in a PWR-W with MAAP5 Fernandez-Cosials, M. K. et al.

February 2015 Annals of Nuclear Energy, vol. 76, p.226-236 17 A Reassessment of Low Probability Containment Failure Modes and Phenomena in a Long-Term Station Blackout Brunett, A.;

Denning, R.;

Aldemir, T.

May 2014 Nuclear Technology, vol.186,

p. 198-215 18 The Ultimate Response Guideline Simulation and Analysis using TRACE, MAAP5, and FRAPTRAN for the Chinshan Nuclear Power Plant Wang, T.-C. et al.

May 2017 Annals of Nuclear Energy, vol. 103, p. 402-411 19 U.S. Nuclear Regulatory Commission's State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Osborn, D. M.

October 2017 SAND2017-10631PE 20 Conceptual Design Enhancement for Prevention and Mitigation of Severe Accidents Song, J. H.

November 2014 Nuclear Technology, vol. 188,

p. 113-122 21 Ongoing SNL International Severe Accident Activities Andrews, N.C.

August 2018 SAND2017-9327R 22 Advanced Seismic Probabilistic Risk Assessment Methodology:

Development of Beta 1.0 MASTODON Toolset Bolisetti, C. et al.

August 2017 INL/EXT-17-43148

A-4 Title Author Date Journal/Technical Report/Conference Citation or Report Number 23 Thermal-Hydraulic Study of Air-Cooled Passive Decay Heat Removal System for APR+ under Extended Station Blackout Kim, D. Y. et al.

February 2019 Nuclear Engineering and Technology, vol. 51, p. 60-72 24 Integral Analyses of Fission Product Retention at Mitigated Thermally-Induced SGTR using ARTIST Experimental Data Rýdl, A.; Lind, T.;

Birchley, J.

February 2016 Nuclear Engineering and Design, vol. 297, p. 175-187 25 An Analysis of Radiological Releases during a Station Black Out Accident for the APR1400 Vo, T. H.; Kim, D.

H.; Song, J. H.

June 2018 Nuclear Engineering and Design, vol. 332, p.22-30 26 Estimating Safety Valve Stochastic Failure-to-Close Probabilities for the Purpose of Nuclear Reactor Severe Accident Analysis Ghosh, S. T. et al.

July 2017 ASME/NRC 2017 13th Pump and Valve Symposium 27 Safety Research Opportunities Post-Fukushima: Initial Report of the Senior Expert Group Organisation for Economic Co-operation and Development/

Nuclear Energy Agency February 2017 NEA/CSNI/R(2016)19 28 SRV Modeling Phillips, J.

August 2017 SAND2017-8365PE 29 Refined Boiling Water Reactor Station Blackout Simulation with RELAP-7 Zhao, H. et al.

September 2014 INL/EXT-14-33162 30 Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis Zhao, H. et al.

April 2014 INL/CON-13-29971 31 RCIC Governing Equation Scoping Studies for Severe Accidents Cardoni, J. N. and Ross, K.

August 2015 SAND2015-6996C 32 Demonstration of Fully Coupled Simplified Extended Station Black-out Accident Simulation with RELAP-7 Zhao, H. et al.

October 2014 INL/CON-13-30940 33 Post-Fukushima Critical Safety Equipment Evaluation Osborn, D.

November 2015 SAND2015-10295C

A-5 Title Author Date Journal/Technical Report/Conference Citation or Report Number 34 A Strongly Coupled Reactor Core Isolation Cooling System Model for Extended Station Black-Out Analyses Zhao, H. et al.

September 2015 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16), Chicago, USA 35 Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report Ross, K. et al.

December 2015 SAND-2015-10662 36 Critical Equipment Performance Osborn, D.

September 2015 SAND2015-8178PE 37 RCIC Governing Equation Scoping Studies Cardoni, J. N.;

Ross, K.; Gauntt, R. O.

August 2015 SAND2015-6995C 38 Probabilistic Economic Valuation of Safety Margin Management Riley, T. et al.

July 2018 DOE contract NE0008295 39 SARNET Benchmark on Phébus FPT3 Integral Experiment on Core Degradation and Fission Product Behaviour Di Giuli, M. et al.

July 2016 Annals of Nuclear Energy, vol. 93, p. 65-82 40 State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability OCED/NEA January 2017 NEA/CSNI/R(2016)15 41 Modeling of Water Ingression Mechanism for Corium Cooling with MELCOR Sevón, T.

February 2017 Nuclear Technology, vol. 197, no. 2, p. 171-179 42 Exercises in Severe Accident Analysis using MELCOR: Accident Walkthrough Gauntt, R. O.

April 2015 SAND2015-3100PE 43 Sensitivity Analysis of Debris Properties in Lower Plenum of a Nordic BWR Galushin, S. and Kudinov, P.

June 2018 Nuclear Engineering and Design, vol. 332, p. 374-382 44 SOARCA Modeling Phillips, J.

August 2017 SAND2017-8638PE 45 Nuclear Severe Accident Modeling and Analysis Humphries, L. L.

November 2015 SAND2015-10296PE 46 Overview of Reactor Safety and Severe Accident Analysis Technologies at Sandia National Laboratories Gauntt, R. O.

June 2007 SAND2007-4143C 47 Modeling of Aerosol Fission Product Scrubbing in Experiments and in Integral Severe Accident Scenarios Rýdyl, A.;

Fernandez-Moguel, L.; Lind, T.

May 2019 Nuclear Technology, vol. 205, no. 5, p. 655-670

A-6 Title Author Date Journal/Technical Report/Conference Citation or Report Number 48 Investigation of the Recriticality Potential during Reflooding Phase of Fukushima Daiichi Unit-3 Accident Darnowski, P.;

Potapczyk, K.;

wirski, K.

January 2017 Annals of Nuclear Energy, vol. 99, p. 495-509 49 Severe Accident Modeling for Cyber Scenarios Cardoni, J.;

Denman, M.

Wheeler, T.

November 2016 Transactions of the American Nuclear Society, vol. 115, no.

1, p. 837-840 50 Updated Peach Bottom Model for MELCOR 1.8.6: Description and Comparisons Robb, K. R.

September 2014 ORNL/TM-2014/207 51 Analisi di un Incidente Non Mitigato di Tipo LOCA in un Reattore PWR Mediante il Codice MELCOR 2.1 (Dissertation)

Pescarini, M.

June 2016 LM-DM270 52 MELCOR 2.1 Analysis of Melt Behavior in a BWR Lower Head during LOCA and SBO Accident Li, G. et al.

April 2016 Annals of Nuclear Energy, vol. 90, p. 195-204 53 Heat Up and Potential Failure of BWR Upper Internals during a Severe Accident Robb, K. R.

January 2015 OSTI ID 1213335 54 Enhanced MELCOR 2.1 Models for Standard PWR-Westinghouse Design/Modelo Detallado para un Diseno Estándar PWR-Westinghouse con el Código MELCOR 2.1 Ruiz Zapatero, M.

et al.

March 2017 Nuclear Espana, vol. 382, p.

49-54 55 MELCOR Workshop OJP Phillips, J. and Humphries, L.

August 2017 SAND2017-8243PE 56 Quicklook Overview of Model Changes in Melcor 2.2: Rev 6342 to Rev 9496 Humphries, L. L.

May 2017 SAND2017-5599 57 Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario Andrews, N. et al.

October 2017 SAND2017-11975 58 Evaluation of JRC Source Term Methodology using MAAP5 as a Fast-Running Crisis Tool for a BWR4 Mark I Reactor Vela-García, M.

and Simola, K.

October 2016 Annals of Nuclear Energy, vol. 96, p. 446-454

A-7 Title Author Date Journal/Technical Report/Conference Citation or Report Number 59 Long Term SBO with Selected Mitigative Measures: MELCOR Parametric Calculations for a 2-Loop PWR Rydl, A. et al.

July 2014 2014 22nd International Conference on Nuclear Engineering (ICONE22),

paper no. ICONE22-30351, p.

V006T15A012 60 Benchmarking Simulation of Long Term Station Blackout Events Kim, S. K.; Lee, J.

C.; Fynan, D. A.

May 2013 Proceedings of the Korean Nuclear Society 2013 spring meeting 61 The Role of Severe Accident Management in the Advancement of Level 2 PRA Modeling Techniques Helton, D. et al.

October 2009 OECD/NEA Workshop on Implementation of Severe Accident Management Measures - Oct 2009 -

Switzerland 62 LEVEL 2 PRA: Explicit and Seamless Approach (Example of a PWR with Large Dry Containment)

Azarm, M. A.

March 2016 n/a 63 Level-2/Level-3 Interface in a PSA Analysis Bixler, N. E.

April 2015 SAND2015-3245PE 64 Development of a Fully-Coupled, All States, All Hazards Level 2 PSA at Leibstadt Nuclear Power Plant Zvoncek, P.;

Nusbaumer, O.;

Torri, A.

March 2017 Nuclear Engineering and Technology, vol. 49, no. 2, p.

426-433 65 Overview of Revised Level 2 PRA Standard in Japan Nakamura, K.;

Narumiya, Y.;

Abe, Y.

June 2016 ICONE24-61070 66 Focus Areas for a Level 2 PSA That Supports a Site NPP Risk Analysis Helton, D. M.;

Zavisca, M.;

Khatib-Rahbar, M.

September 2014 Proceedings of the 2014 European Safety and Reliability Conference 67 Some International Efforts to Progress in the Harmonization of L2 PSA Development and Their Applications (European (ASAMPSA2), U.S. NRC, OECD-NEA and IAEA Activities)

Raimond, E. et al.

October 2009 OECD/NEA Workshop on Implementation of Severe Accident Management Measures - Oct 2009 -

Switzerland 68 BWR MARK I Pressure Suppression Pool Mixing and Stratification Analysis using GOTHIC Lumped Parameter Modeling Methodology Ozdemir, O. E.

and George, T. L.

November 2015 Annals of Nuclear Energy, vol. 85, p. 532-543 69 In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO Kim, T. W.; Han, S. J.; Ahn, K. I.

May 2015 Proceedings of the 2015 Korean National Society Spring Meeting

A-8 Title Author Date Journal/Technical Report/Conference Citation or Report Number 70 OECD/NEA/CSNI Status Report on Filtered Containment Venting Jacquemain, D. et al.

July 2014 NEA-CSNI-R--2014-7 71 Assessment of Primary and Secondary Bleed and Feed Procedures during a Station Blackout in a German Konvoi PWR using ASTECV2.0 Gómez-García-Torano, I. et al.

March 2018 Annals of Nuclear Energy, vol. 113, p. 476-492 72 Analysis of Flammability in the Attached Buildings to Containment under Severe Accident Conditions de la Rosa, J. C.

and Fornós, J.

November 2016 Nuclear Engineering and Design, vol. 308, p. 154-169 73 Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR Parisi, C. et al.

July 2017 INL/EXT-17-42666 74 Total Loss of AC Power Analysis for EPR Reactor Darnowski, P. et al.

August 2015 Nuclear Engineering and Design, vol. 289, p. 8-18 75 Development of Core Relocation Surrogate Model for Prediction of Debris Properties in Lower Plenum of a Nordic BWR Galusin, S. et al.

October 2016 NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Gyeongju, Korea, October 9-13, 2016, Paper N11P1234 76 Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1 Robb, K. R.;

Farmer, M.;

Francis, M. W.

March 2014 ORNL/TM--2014/1 77 HCVS-WP-02: Sequences for HCVS Design and Method for Determining Radiological Dose from HCVS Piping Revision 0 Nuclear Energy Institute October 2014 ADAMS Accession No. ML14309A588 78 Effects of Degradation on the Severe Accident Consequences for a PWR Plant with a Reinforced Concrete Containment Vessel Petti, J. P. et al.

June 2013 NUREG/CR-7149 ADAMS Accession No. ML13172A089 79 Effect of Cesium-Molybdate on Cs Behavior for Source Term Estimation Han, S. J. and Ahn, K. I.

October 2013 Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, October 23-25, 2013 80 BWR Station Blackout: A RISMC Analysis using RAVEN and RELAP5-3D Mandelli, D. et al.

January 2016 INL/JOU-14-33448

A-9 Title Author Date Journal/Technical Report/Conference Citation or Report Number 81 Influence of the Wet-Well Nodalization of a BWR3 Mark I on the Containment Thermal-Hydraulic Response during an SBO Accident Herranz, L. E. et al.

December 2015 Nuclear Engineering and Design, vol. 295, p. 138-147 82 External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure - Scoping Assessment Robb, K. R.

August 2017 ORNL/TM-2017/457 83 Industry Guidance for Compliance with Order EA-13-109: BWR Mark I & II Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions Nuclear Energy Institute March 2015 NEI 13-02 [Rev. 0F4]

ADAMS Accession No. ML15084A454 84 Benchmarking of Fast-running Software Tools Used to Model Releases During Nuclear Accidents Nuclear Energy Agency January 2016 NEA/CSNI/R(2015)19 85 ATLAS Program for Advanced Thermal-Hydraulic Safety Research Song, C.-H.; Choi, K.-Y.; Kang, K.-H.

December 2015 Nuclear Engineering and Design, vol. 294, p. 242-261 86 Analysis of Primary Bleed and Feed Strategies for Selected SBLOCA Sequences in a German Konvoi PWR using ASTEC V2.0 Gómez-García-Torano, I. et al.

December 2017 Annals of Nuclear Energy, vol. 110, p. 818-832 87 A Study on Fission Product Behavior during a Severe Accident at APR1400 Nuclear Power Plants Kim, H.-C. and Cho, S.-W.

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

May 2015 Proceedings of the 2015 Spring meeting of the Korean Nuclear Society 89 Updated analysis of Fukushima Unit 3 with MELCOR 2.1. Part 2: Fission product release and transport analysis Fernandez-Moguel, L.; Rydl, A.; Lind, T.

August 2019 Annals of Nuclear Energy, vol. 130, p.93-106 90 An analysis on the consequences of a severe accident initiated steam generator tube rupture Song, J.-H. et al.

July 2019 Nuclear Engineering and Design, vol. 348, p. 14-23 91 Development of MELCOR thermal hydraulic model of AP1000 and its verification for a DECL break Malickia, M. et al.

June 2019 Annals of Nuclear Energy, vol. 128, p. 44-52

A-10 Title Author Date Journal/Technical Report/Conference Citation or Report Number 92 Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code Galushin, S. and Kudinov, P.

April 2019 Science and Technology of Nuclear Installations, vol.

2019, Article ID 5310808, 18 pages 93 Effect of Molten Corium Behavior Uncertainty on the Severe Accident Progress Choi, W. et al.

August 2018 Science and Technology of Nuclear Installations, vol.

2018, Article ID 5706409, 9 pages 94 Analysis of PWR SBO sequences with RCP passive thermal shutdown seals Mena-Rosell, L. et al.

June 2018 Journal of Nuclear Science and Technology, vol. 55, no.

6, p. 649-662 95 Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes Sancaktar, S. et al.

May 2018 ADAMS Accession No. ML18122A012 96 MELCOR Code Development Status -

EMUG 2018 Humphries, L. L.

April 2018 European MELCOR User Group Meeting, Zagreb, Croatia, April 18-20, 2018 97 Severe Accident Context Quantification for Long-Term Station Blackout in Boiling Water Reactor Nuclear Power Plants Petkov, G. I. and Vela-Garcia, M.

March 2018 ASME Journal of Nuclear Engineering and Radiation Science, vol. 4, no. 2, 020913 98 Tools and Methods for Assessing the Risk Associated with CSGTR Azarm, M. A. and Sancaktar, S.

March 2018 ADAMS Accession No. ML18074A025 99 Severe Accident Phenomena Wagner, K. C.

March 2018 PWR Owners Group SAMG Meeting, Denver, CO, United States, March 27-28, 2018 100 MELCOR Best Practices in SOARCA AMUG 2017 Humphries, L.

November 2017 SAND2017-12028C 101 IVR Phenomena Modeling of Lower Plenum and Modifications of RN Package (1.8.6 to 2.2)

Humphries, L. L.

November 2017 Asian MELCOR User Group Meeting, Daejeon, South Korea, November 5-8, 2017 102 MELCOR Code Development Status MCAP 2017 Humphries, L. L.

October 2017 MELCOR Code Assessment Program Meeting, Bethesda, MD, USA, September 14-15, 2017

A-11 Title Author Date Journal/Technical Report/Conference Citation or Report Number 103 Terry Turbopump Expanded Operating Band Program Experimental and Modeling Efforts Osborn, D.

September 2017 SAND2017-10566PE 104 An Improvement of Estimation Method of Source Term to the Environment for Interfacing System LOCA for Typical PWR Using MELCOR code Han, S.-J.; Kim, T.-W.; Ahn, K.-Il.

June 2017 Journal of Radiation Protection and Research, vol.

42, no. 2, p. 106-113 105 Analysis of unmitigated large break loss of coolant accidents using MELCOR code Pescarini, M. et al.

June 2017 Journal of Physics:

Conference Series, vol. 923, Conference 1

A-12 II.

Offsite Consequence Analysis Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Article, "NRC Keeps Its Focus on Safety," in "Meetings" section n/a May 2008 Nuclear News, vol. 51, no. 6

p. 53 2

A Sensitivity Study on Population Segmentation Effects in the MACCS Code Song, W. et al.

June 2018 Transactions of the American Nuclear Society, vol. 118, no.

1, p.643-646 3

Article, Assessing SOARCA, in Meetings section n/a January 2013 Nuclear News, vol. 56, no. 1,

p. 59 4

Nuclear Energy Related Capabilities at Sandia National Laboratories Pickering, S. Y.

February 2014 SAND2014-1345 5

Technical Basis Development for Filtered Containment Venting System Requirements Basu, S. et al.

November 2013 Transactions of the American Nuclear Society, vol. 109, no.

1, p. 971-974 6

MELCOR/MACCS2 Analysis for BWR Mark I Filtered Containment Venting Osborn, D. M. et al.

November 2013 Transactions of the American Nuclear Society, vol. 109, no.

1, 2103-2106 7

MACCS-HYSPLIT Atmospheric Transport and Dispersion Model Benchmarking Clayton, D. J.

September 2017 SAND2017-9714PE 8

MACCS-HYSPLIT Atmospheric Transport and Dispersion Model Benchmarking Clayton, D. J.;

Bixler, N. E.;

Compton, K.

June 2017 SAND2017-6184C 9

Software Regression Quality Assurance for MACCS2: Version 2.5.0.0 through Version 2.5.0.9 Eubanks, L. L. et al.

July 2012 SAND2012-6333 10 Effect on the Offsite Consequence of the MACCS Non-Site-Specific Best Modeling Practices Used in the US SOARCA Project Jin, D.-S.; Han, S.-K.; Lim, C.-K.

May 2017 Transactions of the Korean Nuclear Society Spring Meeting, May 2017, Jeju, Korea 11 Atmospheric Transport and Dose Calculations: Concepts and Implementation in the MACCS Code Bixler, N. E.

July 2015 SAND2015-5545PE 12 Level-3 Consequence Analysis Part 1 Atmospheric Transport and Dispersion Bixler, N. E.

April 2015 SAND2015-3244PE

A-13 Title Author Date Journal/Technical Report/Conference Citation or Report Number 13 Emergency Planning Zones Estimation for Karachi-2 and Karachi-3 Nuclear Power Plants using Gaussian Puff Model ahin, S. and Ali, M.

October 2016 Science and Technology of Nuclear Installations, vol.

2016 14 Avoidable Dose and Total Dose Radiological Assessments in Support of Public Protection Decisions Kraus, T. D.

June 2013 SAND2013-5158C 15 Importance of Accounting for the Partitioning of Iodine Released During Nuclear Power Plant Accidents Kraus, T. D.

June 2013 SAND2013-4833C 16 MACCS2 Consequence Analysis for BWR Mark I and Mark II Filtered Containment Venting Osborn, D. M. et al.

October 2012 SAND2012-9533 17 Error Analysis of CM Data Products Input Distributions Hunt, B. D. et al.

March 2017 SAND2017-3298R 18 Effects of Source Term on Off-site Consequence in LOCA Sequence in a Typical PWR Han, S.-J.; Kim, T.-W.; Ahn, K.-I.

June 2014 Probabilistic Safety Assessment and Management PSAM 12, June 2014, Honolulu, Hawaii 19 Article, The NRC Staff is Calling for Additional SOARCA Studies," in Late News" section n/a August 2012 Nuclear News, vol. 55, no. 9,

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

p. 25-26 21 Best-Estimate Calculations of Unmitigated Severe Accidents in State-of-the-Art Reactor Consequence Analyses Schaperow, J. H.

et al.

October 2009 OECD/NEA Workshop on Implementation of Severe Accident Management Measures - Oct 2009 -

Switzerland 22 Consequence Analysis for Nuclear Reactors, Yongbyon Kang, T. and Jae, M.

February 2017 Journal of Nuclear Science and Technology, vol. 54, no.

2, p. 223-232 23 Applicability of 100 TBq Cesium 137 Release into Environment as a Safety Criterion for Consequence Assessment at Reactor Design Approval Stage Silva, K. and Okamoto, K.

December 2015 Journal of Nuclear Science and Technology, vol. 52, no.

12, p. 1530-1539

A-14 Title Author Date Journal/Technical Report/Conference Citation or Report Number 24 Treatment of Accident Mitigation Measures in State-of-the-Art Reactor Consequence Analyses Schaperow, J. H.

et al.

October 2009 OECD/NEA Workshop on Implementation of Severe Accident Management Measures - Oct 2009 -

Switzerland 25 Risk and regulatory considerations for small modular reactor emergency planning zones based on passive decontamination potential Carless, T. S.;

Talabi, S. M.;

Fischbeck, P. S.

January 2019 Energy, vol. 167, p. 740-756 26 Status of Practice for Level 3 Probabilistic Safety Assessments OECD/NEA November 2018 NEA/CSNI/R(2018)1 27 Response Technical Tools User Guide Andrews, N.; Fu, C.; Kaberlein, A.

M.

August 2018 SAND2018-9593 28 Technical Basis for the Containment Protection and Release Reduction Rulemaking for Boiling Water Reactors with Mark I and Mark II Containments Barr, J. A. et al.

March 2018 NUREG-2206 ADAMS Accession No. ML18065A048 29 U.S. Nuclear Regulatory Commission's State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Osborn, D. M.

October 2017 4th Arab Forum on the Prospects of Nuclear Power Meeting, Amman, Jordan, October 9-12, 2017 30 Recommendations for Future Research on Nuclear Accident Consequence Analysis Kampanart, S.

and Vechgama, W.

July 2017 International Conference on Risk Analysis, Decision Analysis, and Security, Tsinghua University, Beijing, China, July 21-23, 2017 31 MELCOR Accident Code System (MACCS)

Sharp, A. E.

April 2017 NUREG/BR-0527 ADAMS Accession No. ML17102A912 32 Dose assessment in level 3 PRA - a rewiew of recently used methods Karanta, I.

February 2017 VTT-R-00738-17 33 A Study of Base Technology of Korean Specific Level 3 PSA Code Development Han, S. et al.

February 2017 KAERI/RR--4232/2016

A-15 Title Author Date Journal/Technical Report/Conference Citation or Report Number 34 Probabilistic risk assessment from potential exposures to the public applied for innovative nuclear installations Dvorzhak, A.;

Mora, J. C.;

Robles, B.

August 2016 Reliability Engineering and System Safety, vol.152, p.176-186

A-16 III.

Accident-Tolerant Fuel Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Accident Tolerant Fuels (ATF) Coating and Cladding Thermal Hydraulic Properties Evaluation by MELCOR YU 1.8.6: Benchmark for SURRY Short Term Station Black Out Wang, J.;

Corradini, M. L.;

Jo, H.

June 2018 Transactions of the American Nuclear Society, vol. 118, no.

1, p. 1259-1262 2

Potential Recovery Actions from a Severe Accident in a PWR: MELCOR Analysis of a Station Blackout Scenario Wang, J.; Jo, H.J.;

Corradini, M. L.

October 2018 Nuclear Technology, vol. 204,

p. 1-14 3

Estimation of Core Recovery Time with TRACE Gurgen, A. and Shirvan, K.

July 2017 Transactions of the American Nuclear Society, vol. 116, no.1, p. 422-425 4

State-of-the-Art Report on Light Water Reactor Accident-Tolerant Fuels OECD/NEA October 2018 NEA No. 7317 5

Estimation of Coping Time in Pressurized Water Reactors for Near Term Accident Tolerant Fuel Claddings Gurgen, A. and Shirvan, K.

October 2018 Nuclear Engineering and Design, vol.337, p. 38-50 6

Development of Accident Tolerant Fuel Options for Near Term Applications:

Fuel Performance Modeling under Transient/Severe Accidents by MELCOR: PART I: Benchmark for SURRY: Short Term Station Black Out (STSBO)

Wang, J.;

Corradini, M. L.;

Haskin, T.

November 2016 Transactions of the American Nuclear Society, vol. 115, no.

1, 1799-1802 7

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs Robb, K. R.

August 2015 ORNL/SPR-2015/347 8

Accident Tolerant Clad Material Modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out Wang, J. et al.

March 2017 Nuclear Engineering and Design, vol. 313, p. 458-469 9

Accident-Tolerant Fuel Valuation:

Safety and Economic Benefits Hess, S. et al.

March 2019 EPRI TR-3002015091 (not publicly available) 10 Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents Robb, K. R.

January 2015 16th International Topical Meeting on Nuclear Reactor Thermalhydraulics, Chicago, IL, USA

A-17 Title Author Date Journal/Technical Report/Conference Citation or Report Number 11 System Code Evaluation of Accident Tolerant Claddings During BWR Station Blackout Accident Wu, X. and Shirvan, K.

November 2018 Transactions of the American Nuclear Society, vol. 119, no.

1, p. 444-447 12 Plant-Level Scenario-Based Risk Analysis for Enhanced Resilient PWR -

SBO and LBLOCA Ma, Z. et al.

September 2018 INL/EXT-18-51436-Rev000 13 State-of-the-Art Report on Light Water Reactor Accident-Tolerant Fuels OECD/NEA January 2018 NEA No. 7317 14 Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs Bragg-Sitton, S.

M.

September 2017 INL/EXT-17-43389-Rev000 15 Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out Wang, J. et al.

March 2017 Nuclear Engineering and Design, vol. 313, p. 458-469

A-18 IV.

Advanced Reactor Analysis Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Risk Management for Sodium Fast Reactors Denman, M. R. et al.

January 2015 SAND2015-0542 2

Regulatory Technology Development Plan - Sodium Fast Reactor:

Mechanistic Source Term - Trial Calculation Grabaskas, D. et al.

October 2016 ANL-ART-49-Vol1-Vol2 3

Overview of Key Computer Codes for the PGSFR Safety Analysis Chang, W.-P.; Lee K.-L.; Yoo, J.

October 2016 Korean Nuclear Society Autumn Meeting, Gyeongju, Korea, October 26-28, 2016 4

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting Smith, C.

September 2013 INL/EXT-13-30170 5

Methodology for Establishing the Technical Basis for Plume Exposure Emergency Planning Zones at NuScale Small Modular Reactor Plant Sites NuScale Power December 2015 NuScale TR-0915-17772-NP ADAMS Accession No. ML15356A842 6

Proposed Methodology and Criteria for Establishing the Technical Basis for Small Modular Reactor Emergency Planning Zone (White Paper)

Nuclear Energy Institute December 2013 ADAMS Accession No. ML13364A345 7

Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

Denman, M. R. et al.

April 2015 SAND2015--2484 8

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel Klein, A. et al.

January 2015 DOE/NEUP-11-3097 OSU-NE-VF-1204 9

A Methodology for Accident Analysis of Fusion Breeder Blankets and Its Application to Helium-Cooled Lead-Lithium Blanket Panayotov, D. et al.

October 2016 IEEE Transactions on Plasma Science, vol. 44, no. 10, p.

2511-2522 10 Design and Licensing Strategies for the Fluoride-Salt-Cooled, High-Temperature Reactor (FHR)

Technology Scarlat, R. O. et al.

November 2014 Progress in Nuclear Energy, vol. 77, p. 406-420

A-19 Title Author Date Journal/Technical Report/Conference Citation or Report Number 11 Fluoride-Salt-Cooled, High-Temperature Reactor (FHR)

Subsystems Definition, Functional Requirement Definition, and Licensing Basis Event (LBE) Identification White Paper Allen, T.R. et al.

August 2013 n/a 12 Regulatory cross-cutting topics for fuel cycle facilities Denman, M. R. et al.

October 2013 SAND2013-9367 13 Advanced Reactor Safety Program -

Stakeholder Interaction and Feedback Szilard, R. H. and Smith, C. L.

August 2014 INL/EXT-14-32928 14 Final White Paper on Small Advanced Reactor Design Reviews Rev 7 2017/05/08 Holahan, G. M.

May 2017 ADAMS Accession No. ML17213A849 (not publicly available) 15 Mechanistic Source Term Modeling for Sodium Fast Reactors Clark, A. J.;

Denman, M. R.;

Grabaskas, D.

January 2017 SAND2017-0454C

A-20 V.

Emergency Preparedness and Response Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

RASCAL 4.3 Kowalczik, J.

April 2014 ADAMS Accession No. ML14112A349 2

RASCAL 4.3: Description of Models and Methods Ramsdell, Jr, J.V.;

Athey, G. F.;

Rishel, J. P.

May 2015 NUREG-1940, Supplement 1 3

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

May 2013 EPR-NPP-PPA-2013 4

Operational Intervention Levels for Reactor Emergencies and Methodology for Their Derivation. March 2017 International Atomic Energy Agency, Incident and Emergency Centre March 2017 EPR-NPP-OILS-2017 5

Alternative evacuation strategies for nuclear power accidents Hammond, G. D.

and Bier, V. M.

March 2015 Reliability Engineering &

System Safety, vol. 135, p. 9-14 6

Significance Quantification Process for Emergency Preparedness Oversight Jones, J. A.

January 2014 SAND2014-0677C 7

Emergency Preparedness Significance Quantification Process: Proof of Concept Sullivan, R. et al.

June 2013 INL/LTD-12-27648, Rev. 1 8

Emergency Preparedness Osborn, D.

June 2015 SAND2015-4657PE 9

Inspection of Emergency Arrangements OECD/NEA Committee on Nuclear Regulatory Activities WGIP December 2013 NEA-CNRA-R-2013-2 10 Emergency Preparedness. NRC Needs to Better Understand Likely Public Response to Radiological Incidents at Nuclear Power Plants Government Accountability Office March 2013 GAO-13-243 11 Rethinking Nuclear Emergency Planning, Preparations, and Response Goble, R. and Bier, V.

November 2013 Transactions of the American Nuclear Society vol. 109 no.

1, p. 1959-1961

A-21 Title Author Date Journal/Technical Report/Conference Citation or Report Number 12 A new approach to quantify safety benefits of disaster robots Kim, I. S.; Choi, Y.; Jeong, K. M.

October 2017 Nuclear Engineering and Technology, vol. 49, no. 7 p.

1414-1422 13 The Risk of Extended Power Loss and the Probability of Emergency Restoration for Severe Events and Nuclear Accidents Duffey, R. B.

May 2019 ASME Journal of Nuclear Engineering and Radiation Science, vol. 5, no. 3, 031601 14 About the development of a national system of response to the nuclear emergency for agriculture Mikailova, R. A.

and Shubina, O.

A.

March 2019 IOP Conference Series:

Material Science and Engineering 487 012013 15 GISBased Integration of Social Vulnerability and Level 3 Probabilistic Risk Assessment to Advance Emergency Preparedness, Planning, and Response for Severe Nuclear Power Plant Accidents Pence, J. et al.

November 2018 Risk Analysis, vol. 39, no. 6,

p. 1262-1280

A-22 VI.

Dynamic PRA Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Risk-Informed Safety Margin Characterization Methods Development Work Smith, C. L. et al.

September 2014 INL/EXT-14-33191 2

System Theoretic Frameworks for Mitigating Risk Complexity in the Nuclear Fuel Cycle Williams, A. D. et al.

September 2017 SAND2017-10243 3

Safety Relief Valve Cyclic Failure Analysis for Use in Discrete Dynamic Event Trees Denman, M. R.

May 2013 SAND2013-3684C 4

Seamless Level 2/Level 3 Probabilistic Risk Assessment Using Dynamic Event Tree Analysis (Dissertation)

Osborn, D. M.

January 2013 n/a 5

Development and Application of a Dynamic Level 1 and 2 Probabilistic Safety Assessment Tool LaChance, J. L. et al.

January 2012 SAND2012-0651C 6

Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis (LDRD 17-0958) FY17 Report Wheeler, T. A. et al.

September 2017 SAND2017-10307 7

Dynamical systems probabilistic risk assessment Denman, M. R.

and Ames, A. L.

March 2014 SAND2014-4037 8

Safety Relief Valve Cyclic Failure Analysis for use in Discrete Dynamic Event Trees (slidedeck)

Denman, M. R.

September 2013 SAND2013-8217C 9

Discrete Dynamic Event Tree Analysis of Small Modular Reactor Severe Accident Management (slidedeck)

Denman, M. R. et al.

September 2013 SAND2013-8096C 10 Discrete Dynamic Event Tree Analysis of Small Modular Reactor Severe Accident Management Denman, M. R. et al.

May 2013 SAND2013-3680C 11 The Assessment of Low Probability Containment Failure Modes using Dynamic PRA (Dissertation)

Brunett, A. J.

January 2013 n/a

A-23 Title Author Date Journal/Technical Report/Conference Citation or Report Number 12 Discrete Dynamic Probabilistic Risk Assessment Model Development and Application LaChance, J. et al.

October 2012 SAND2012-9346 ADAMS Accession No. ML12305A351 13 An Assessment of Low Probability Containment Failure in a Long-Term Station Blackout Using Dynamic PRA Brunett, A.;

Denning, R.;

Aldemir, T.

November 2013 Transactions of the American Nuclear Society, vol. 109, no.

1, 954-957 14 Preliminary Cyber-Informed Dynamic Branch Conditions for Analysis with the Dynamic Simplified Cyber MELCOR Model Denman, M. R. et al.

November 2016 Transactions of the American Nuclear Society, vol. 115, no.

1, 787-790 15 How to ADAPT Jankovsky, Z. K.;

Haskin, T. C.;

Denman, M. R.

June 2018 SAND2018-6660 16 Uncertainty Quantification for External Events Analysis of LWRS/RISMC Project Parisi, C. et. al.

June 2017 Transactions of the American Nuclear Society, vol. 116, no.

1, 795-797 17 Quantitative risk reduction by means of recovery strategies París, C. et al.

February 2019 Reliability Engineering &

System Safety Meeting, vol.

182, p. 13-32 18 Level 2 Probabilistic Risk Assessment Using Dynamic Event Tree Analysis Osborn, D. et al.

June 2018 Chapter 5 of Advanced Concepts in Nuclear Energy Risk Assessment and Management 19 Development of Computational and Data Processing Tools for ADAPT to Assist Dynamic Probabilistic Risk Assessment Jankovsky, Z. K.

January 2018 Dissertation 20 A Dynamic Assessment of an Interfacing System Loss of Coolant Accident Jankovsky, Z. K. ;

Denman, M. R. ;

Aldemir, T.

September 2017 SAND2017-10141C 21 A Dynamic Assessment of Auxiliary Building Contamination and Failure Due to a Cyber-Induced Interfacing System Loss of Coolant Accident Jankovsky, Z. ;

Denman, M. R. ;

Aldemir, T.

June 2017 International Conference on Topical Issues in Nuclear Installation Safety:

Safety Demonstration of Advanced Water Cooled Nuclear Power Plants, VIC, Vienna, 6-9 June 2017

A-24 VII.

Multi-Unit PRA Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Construction of Multi-Path Event Tree for Station Blackout Events Kim, S. et al.

October 2017 Transactions of the American Nuclear Society, vol. 117, no.

1, p. 951-953 2

Multi-Unit Level 3 Probabilistic Safety Assessment: Approaches and Their Application to a Six-Unit Nuclear Power Plant Site Kim, S.-Y. et al.

December 2018 Nuclear Engineering and Technology, vol. 50, no. 8, p.

1246-1254 3

Multi-Unit Accident Contributions to Quantitative Health Objectives: A Safety Goal Policy Analysis Hudson, D. W. and Modarres, M.

March 2017 Nuclear Technology, vol. 197, no. 3, p. 227-247 4

Holistic Approach to Multi-Unit Site Risk Assessment: Status and Issues Kim, I. S.; Jang, M.;

Kim, S. R.

March 2017 Nuclear Engineering and Technology, vol. 49, no. 2, p.

286-294 5

Initiating Events for Multi-Reactor Plant Sites Muhlheim, M. D.;

Flanagan, G. F.;

Poore, III, W. P.

September 2014 ORNL/TM-2014/533 6

Scoping Estimates of Multiunit Accident Risk Stutzke, M. A.

June 2014 Probabilistic Safety Assessment and Management PSAM 12, Honolulu, Hawaii, June 2014 7

A Methodology for Performing Consequence Analysis for Multi-Unit/Spent Fuel Pool Source Terms Bixler, N. E.

November 2014 SAND2014-20038C 8

Multi-Unit Accident Contributions to US Nuclear Regulatory Commission Quantitative Health Objectives: A Safety Goal Policy Analysis using Models from State-of-the-Art Reactor Consequence Analyses (Dissertation)

Hudson, D. W.

January 2016 n/a 9

Multi-Unit Dynamic PRA Mandelli, D. et al.

May 2019 Reliability Engineering &

System Safety, vol. 185, p.

303-317 10 Multi-Unit Level 3 Probabilistic Safety Assessment: Approaches and Their Application to a Six-Unit Nuclear Power Plant Site Kim, S.-Y. et al.

December 2018 Nuclear Engineering and Technology, vol. 50, no. 8, p.

1246-1254

A-25 Title Author Date Journal/Technical Report/Conference Citation or Report Number 11 The Current Research Status and Technical Development Framework of Multi-Reactor Probabilistic Consequence Assessment Ding, H. et al.

July 2017 25th International Conference on Nuclear Engineering (ICONE25), Shanghai, China, July 2-6, 2017

A-26 VIII.

Fukushima Forensic Analysis Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

MELCOR Applications to SOARCA and Fukushima Gauntt, R. O.

March 2014 SAND2014-1879C 2

A MELCOR Model of Fukushima Daiichi Unit 1 Accident Sevón, T.

November 2015 Annals of Nuclear Energy, vol. 85, p. 1-11 3

A MELCOR Model of Fukushima Daiichi Unit 3 Accident Sevón, T.

April 2015 Nuclear Engineering and Design, vol. 284, p. 80-90 4

A Review of Recent SNL MELCOR Fukushima Accident Analyses Kalinich, D. A.

March 2015 SAND2015-2172PE 5

Overview of Sandia National Laboratories MELCOR Fukushima Analyses Kalinich, D. A.

July 2015 SAND2015-6035PE 6

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I Gauntt, R. O. and Mattie, P. D.

January 2016 SAND2016-0232 7

Insight from Fukushima Daiichi Unit 3 Investigations Using MELCOR Robb, K. R.;

Francis, M. W.;

Ott, L. J.

May 2014 Nuclear Technology, vol. 186, no. 2, p. 145-160 8

MELCOR Simulations of the Severe Accident at the Fukushima Daiichi Unit 1 Reactor Gauntt, R. et al.

May 2014 Nuclear Technology, vol. 186, no. 2, p. 161-178 9

MELCOR Simulations of the Severe Accident at Fukushima Daiichi Unit 3 Cardoni, J. et al.

May 2014 Nuclear Technology, vol. 186, no. 2, p. 179-197 10 Sensitivity Study of 1F1 Type Accident by MELCOR Code Saito, K. and Yamaji, A.

October 2015 Transactions of the American Nuclear Society, vol. 113, no.

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

February 2014 SAND2014-1170 12 Seismically-Induced Reactor Coolant Leakage as an Allegedly-Possible Cause of Accident at Unit 1 of Fukushima Daiichi Nuclear Power Station Kukita, Y. and Watanabe, N.

November 2014 JAEA-TECHNOLOGY--2014-036 13 MELCOR 2.1 Simulations of Fukushima Unit 3 Cardoni, J. N.

November 2011 SAND2012-9742C

A-27 Title Author Date Journal/Technical Report/Conference Citation or Report Number 14 Severe Accident Progression Analyses in SOARCA and Comparisons to Fukushima Gauntt, R. O.

July 2013 SAND2013-6058C 15 Fukushima Daiichi - A Case Study for BWR Instrumentation and Control Systems Performance during a Severe Accident Rev 0 Clayton, D. A. and Poore, III, W. P.

April 2013 ORNL/TM-2013/154 16 Fukushima Daiichi Nuclear Plant Accident: Atmospheric and Oceanic Impacts over the Five Years Hirose, K.

June 2016 Journal of Environmental Radioactivity, vol. 157, p.

113-130 17 Analysis of the Accident in the Fukushima Daiichi Nuclear Power Station Unit 3 with MELCOR_2.1 Fernandez-Moguel, L. and Birchley, J.

September 2015 Annals of Nuclear Energy, vol. 83, p. 193-215 18 Fukushima Daiichi - A Case Study for BWR Instrumentation and Control Systems Performance during a Severe Accident Rev 1 Clayton, D. A. and Poore, III, W. P.

June 2014 ORNL/TM-2013/154 R1 19 ATHLET-CD/COCOSYS Analyses of Severe Accidents in Fukushima Daiichi Units 2 and 3: German Contribution to the OECD/NEA BSAF Project, Phase 1 Sonnenkalb, M.

and Band, S.

November 2016 Nuclear Technology, vol. 196, no. 2, p. 211-222 20 Fukushima Daiichi Unit 1 Uncertainty Analysis-Exploration of Core Melt Progression Uncertain Parameters-Volume II Denman, M. R.

and Brooks, D. M.

August 2015 SAND2015-6612 21 Presentation of Fukushima Analyses to U.S. Nuclear Power Plant Simulator Operators and Vendors Osborn, D.;

Kalinich, D. A.;

Cardoni, J. N.

February 2015 SAND2015-1177R 22 Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents Andrews, N. C.;

Gauntt, R. O.

October 2017 SAND2017-10811R 23 Fukushima Daiichi Accident Study:

Status as of April 2012 Gauntt, R. O. et al.

July 2012 SAND2012-6173 24 Historical Overview of Fukushima Forensics Work Kelly, J. E.

March 2018 ADAMS Accession No. ML18090A006

A-28 Title Author Date Journal/Technical Report/Conference Citation or Report Number 25 Team Performance Comparison in Core-Melt Units of Fukushima Daiichi NPS Based on Dynamic Context Quantification of Accident Petkov, G. and Petkov, I.

June 2017 PSAM Topical Conference on Human Reliability, Quantitative Human Factors, and Risk Management, Munich, Germany, June 2017 26 Studies on the Recriticality Potential during Fukushima Unit-3 Core Reflooding Darnowskia, P.;

Potpaczyka, K.;

wirski, K.

May 2017 The 8th European Review Meeting on Severe Accident Research - ERMSAR-2017 Warsaw, Poland, 16-18 May 2017 27 Fukushima Reactor Building Model Andrews, N. et al.

January 2017 4th Meeting of the OECD/NEA BSAF Project Phase 2, Paris, France, January 9-13, 2017

A-29 IX.

Fukushima Lessons Learned Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Nuclear Power Plant Severe Accidents Osborn, D.

March 2015 SAND2015-1591PE 2

Health Physics: Radiation-Generating Devices, Characteristics, and Hazards (Ch. 7: Regulatory Considerations)

Bevelacqua, J. J.

April 2016 n/a 3

Nuclear Regulation in the United States and a Possible Framework for an International Regulatory Approach Bevelacqua, J. J.

February 2013 International Nuclear Safety Journal, vol. 2, no. 1, p. 52-57 4

Lessons Learned from the Fukushima Nuclear Accident for Improving Safety of U.S. Nuclear Plants National Research Council January 2014 Contract No. NRC-HQ-12-G-03-0002 5

Lessons Learned from the Fukushima Accident for Improving Safety and Security of U.S. Nuclear Plants. Phase 2

National Research Council January 2016 Contract No. NRC-HQ-12-G-03-0002 6

The Nuclear Regulatory Commission and NEPA Review Eccleston, C. H.

September 2012 Environmental Quality Management, vol. 22, no.1, p.

43-58 7

Fukushima Lessons Learned Gauntt, R. O.

February 2012 SAND2012-0893C 8

The Canary, the Ostrich, and the Black Swan: A Historical Perspective on our Understanding of BWR Severe Accidents and Their Mitigation Greene, S. R.

May 2014 Nuclear Technology, vol. 186, no. 2, p. 115-138 9

SNL BSAF Phase 2 Activities Andrews, N. et al.

December 2015 SAND2015-10448PE

A-30 X.

Severe Accident Management Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario Rempe, J. L.;

Knudson, D. L.;

Lutz, R. J.

September 2015 INL/EXT-15-35940 2

Technical Basis for Severe Accident Mitigating Strategies: Volume 1 Duvall, A. et al.

April 2015 EPRI 3002003301 3

Verification of SAMGs in SBO Sequences with Seal LOCA. Multiple Damage Domains Queral, C. et al.

December 2016 Annals of Nuclear Energy, vol. 98, p.90-111 4

Preliminary Assessment for the Mitigative Effectiveness of External Injection during Extended SBO Park, S. Y. and Ahn, K. I.

October 2013 Proceedings of the 2013 fall Korean Nuclear Society meeting 5

The Relative Importance of Mitigation, Early Phase, Intermediate Phase, and Late Phase Response Denning, R. et al.

November 2013 Transactions of the American Nuclear Society, vol. 109, no.

1 6

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure -

Expanded Scoping Assessment Robb, K. R.

July 2018 ORNL/TM-2018/901 7

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident Clayton, D. A. and Poore, III, W. P.

January 2015 ORNL/LTR-2014/320 8

Evaluation of an Accident Management Strategy of Emergency Water Injection using Fire Engines in a Typical Pressurized Water Reactor Park, S.-Y. and Ahn, K.-I.

October 2015 Nuclear Engineering and Technology, vol. 47, no. 6, p.

719-728 9

Post-Severe Accident Environmental Conditions for Essential Instrumentation for Boiling Water Reactors Clayton, D. and Poore, M.

September 2015 ORNL/TM-2015/278 10 Regulatory Perspective and Accident Management Procedure Influences on Accident Monitoring Instrumentation Design Criteria Rahn, D. L. and Cowdrey, C. B.

September 2012 ADAMS Accession No. ML12243A196 IAEA/JNESO Workshop on Accident Monitoring Instrumentation, Tokyo, Japan, September 4, 2012

A-31 11 Dealing with Beyond-Design-Basis Accidents in Nuclear Safety Decisions Nourbakhsh, H. P.

June 2014 Probabilistic Safety Assessment and Management PSAM 12, Honolulu, Hawaii, June 2014 12 iROCS: Integrated Accident Management Framework for Coping with Beyond-Design-Basis External Events Kim, J. et al.

March 2016 Nuclear Engineering and Design, vol. 298, p. 1-13 13 APR1400 Design Certification Severe Accident Mitigation Design Alternatives Technical Analysis in Support of the Environmental Assessment Palmrose, D. E.

September 2018 ADAMS Accession No. ML18096A697 14 FLEX Strategy Implementation for LOCA Sequences in PWR-Westinghouse Ruiz-Zapatero, M.; Bocanegra, R.; Queral, C.

September 2017 26th International Conference Nuclear Energy for New Europe, Bled, Slovenia, September 11-14, 2017 15 Dynamic Human Performance Context Comparison for Severe Accident Management during Long Term Station Blackout in Light Water Reactors Petkov, G. and Petkov, I.

June 2017 2017 European Safety and Reliability Conference (ESREL2017), Portoroz, Slovenia, June 18-22, 2017 16 Development of Site Risk Assessment and Management Technology including Extreme External Events Yang, J. E. et al.

March 2017 KAERI/RR--4225/2016

A-32 XI.

Level 1 PRA Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Use and Development of Probabilistic Safety Assessment Nuclear Energy Agency December 2012 NEA/CSNI/R(2012)11 2

The NRC's SPAR Models: Current Status, Future Development, and Modeling Issues Buell, R. F.

September 2008 INL/CON-08-14484 3

SPAR Integrated Capabilities Model (SPAR-ICM) Project Ma, Z. et al.

July 2013 INL/LTD-12-27648, Rev. 1 (not publicly available) 4 U.S. NRC Confirmatory Level 1 PRA Success Criteria Activities Helton, D.;

Esmaili, H.; Buell, R.

March 2011 INL/CON-11-21026 5

Development Strategy of Optimized Level 1&2 PSA Model for APR1400 NPPs Reflecting New Severe Accident Mitigating System and Its Application Hwang, S.-W. et al.

December 2018 Annals of Nuclear Energy, vol. 122, p. 256-269 6

Lessons Learned from Applying a New HRA Method for the Petroleum Industry Taylor, C.; Øie, S.;

Gould, K.

Oct 2018 Reliability Engineering &

System Safety, available online 5 October 2018

A-33 XII.

Nuclear Safety/Societal Risk Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Insights on Risk Margins at Nuclear Power Plants: A Technical Evaluation of Margins in Relation to Quantitative Health Objectives and Subsidiary Risk Goals in the United States Electric Power Research Institute May 2018 EPRI Report No. 3002012967 2

The Three Mile Island, Chernobyl, and Fukushima Daiichi Accidents and Their Radiological Impacts Bevelacqua, J. J.

June 2016 International Nuclear Safety Journal, vol. 5, no.1, p. 21-79 3

Software Verification and Validation:

Examples from the Safety Arena Siu, N.

September 2015 ADAMS Accession No. ML15253A634 INMM Workshop on VA Tools, Boston, MA; September 14-16, 2015 4

Development of an Updated Societal-Risk Goal for Nuclear Power Safety Bier, B. et al.

July 2014 INL/CON-13-30391 5

The Societal Risk of Severe Accidents in Nuclear Power Plants Denning, R. and McGhee, S.

June 2013 Transactions of the American Nuclear Society, vol. 108, no.

1, p. 521-525 6

Insights into the Societal Risk of Nuclear Power Plant Accidents Denning, R. and Mubayi, V.

January 2017 Risk Analysis, vol. 37, no. 1,

p. 160-172 7

Understanding the Nature of Nuclear Power Plant Risk Denning, R. S.

June 2012 ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States),

June 24-28, 2012 8

Critical Reflections on Nuclear and Renewable Energy: Environmental Protection and Safety in the Wake of the Fukushima Nuclear Accident.

(Chapter 1: Reliability and Nuclear Power and Postscript)

Kuo, W.

March 2014 n/a (Book) 9 Preliminary Study for Development of Technical Basis of Quantitative Safety Goals for NPPs Operations in UAE Shehhi, O. A. A. et al.

June 2017 Transactions of the American Nuclear Society, vol. 116, no.

1, p. 801-803

A-34 Title Author Date Journal/Technical Report/Conference Citation or Report Number 10 Impact of Probabilistic Risk Assessment and Severe Accident Research in Reducing Reactor Risk Denning, R. S. and Budnitz, R. J.

January 2018 Progress in Nuclear Energy,

v. 102 p.90-102 11 Nuclear Weapon Pit Production:

Options to Help Meet A Congressional Requirement Medalia, J. E.

May 2015 CRS Report No. R44033 12 Nuclear Renaissance, Public Perception and Design Criteria: An Exploratory Review Goodfellow, M. J.;

Williams, H. R.;

Azapagic, A.

October 2011 Energy Policy, vol. 39, no. 10,

p. 6199-6210 13 Reliability and Nuclear Power Kuo, W.

June 2011 IEEE Transactions on Reliability, vol. 60 no. 2, p.

365-367 14 International Conference on Topical Issues in Nuclear Installation Safety:

Defence in Depth Advances and Challenges for Nuclear Installation Safety. Proceedings of an International Conference held in Vienna, Austria, 21-24 October 2013 International Atomic Energy Agency, Safety Assessment Section, Vienna (Austria)

October 2014 IAEA-TECDOC-CD--1749 15 Handbook of Advanced Nuclear Hydrogen Safety. 1st edition Hino, R. et al.

January 2017 JAEA-REVIEW-2016-038 16 Nuclear Facility Accident (NFAC)

Unit Test Report for HPAC Version 6.4 Lee, R. W. and Sulfredge, C. D.

March 2017 ORNL/TM-2017/94 17 Nuclear Energy: Overview of Congressional Issues. Updated November 16, 2018 Holt, M.

November 2018 CRS Report No. R42853 18 Environmental Impact Statement for the Combined License (COL) for Enrico Fermi Unit 3. Final Report.

Appendix E NRC and U.S. Army Corps of Engineers January 2013 NUREG-2105 v.3 ADAMS Accession No. ML12307A177 19 Comparative Radioecological Assessment of Serious-Accident Scenarios in NPP on the Basis of the Risk for Natural Communities Spiridonov, S. and Mikailova, R.

January 2019 Atomic Energy, vol.125, no.

3, p. 198-203

A-35 Title Author Date Journal/Technical Report/Conference Citation or Report Number 20 System Studies for Global Nuclear Assurance & Security: 3S Risk Analysis for Small Modular Reactors (Volume I) - Technical Evaluation of Safety Safeguards & Security Williams, A. D. et al.

October 2018 SAND2018-12447 21 Integrating Thermal Energy Storage and Nuclear Reactors: A Technical and Policy Study Abel, C. R.

May 2018 Dissertation 22 Irradiation Dose of the Woody Tier of a Coniferous Forest Due to Accidental Emissions from NPP Mikailova, R. A. and Spiridonov, S. I.

January 2018 Atomic Energy, vol. 123, no.

3, p. 202-208 23 Uncertainties in Estimating Health Risks Associated with Exposure to Ionising Radiation Preston, R. J. et al.

June 2013 Journal of Radiological Protection, vol. 33, no. 3, p.

573-588

A-36 XIII.

Sensitivity and Uncertainty Analyses Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Proposed Analyses for xLPR 2.0 Outputs Sallaberry, C. J.-

M. et al.

June 2015 SAND2015-5118PE 2

Bootstrapped-Ensemble-Based Sensitivity Analysis of a Trace Thermal-Hydraulic Model Based on a Limited Number of PWR Large Break LOCA Simulations Di Maio, F. et al.

September 2016 Reliability Engineering &

System Safety, vol. 153, p.

122-134 3

SOARCA Surry Power Station Uncertainty Analysis: Parameter Methodology and Insights Jones, J. et al.

June 2014 Probabilistic Safety Assessment and Management PSAM 12, Honolulu, Hawaii, June 2014 4

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis MELCOR Parameters and Probabilistic Results Osborn, D. et al.

April 2013 SAND2013-3107C 5

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis MACCS2 Aleatory Weather Effects Osborn, D. et al.

April 2013 SAND2013-3102C 6

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis MACCS2 Dose Truncation Sensitivity Osborn, D. et al.

April 2013 SAND2013-3105C 7

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis MACCS2 Parameters and Probabilistic Results Osborn, D. et al.

April 2013 SAND2013-3106C 8

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Contributions to Overall Uncertainty Bixler, N. E. et al.

June 2014 Probabilistic Safety Assessment and Management PSAM 12, Honolulu, Hawaii, June 2014 9

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Knowledge Advancement Gauntt, R. O. et al.

February 2014 SAND2014-1344C

A-37 Title Author Date Journal/Technical Report/Conference Citation or Report Number 10 SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis Probabilistic Methodology and Regression Technique Osborn, D. et al.

April 2013 SAND2013-3104C 11 SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Convergence of the Uncertainty Results Bixler, N. E. et al.

February 2014 SAND2014-1346C 12 Uncertainty Analysis of Consequence Management (CM) Data Products Hunt, B. D. et al.

January 2018 SAND2018-0329 13 Phenomenological Uncertainties in SOARCA Study Gauntt, R. O.

March 2011 SAND2011-2024C 14 The SOARCA Surry Power Station Short-Term Station Blackout Uncertainty Analysis: MACCS Parameter Development Bixler, N. E. et al.

June 2015 Transactions of the American Nuclear Society, vol. 112, no.

1, p. 515-516 15 A Systematic Framework for Effective Uncertainty Assessment of Severe Accident Calculations; Hybrid Qualitative and Quantitative Methodology Hoseyni, S. M. et al.

May 2014 Reliability Engineering &

System Safety, vol. 125, p.

22-35 16 Sensitivity Analysis of Successful Evacuee Proportion in Hypothetical NPP Accident by Earthquake Kim, S.-Y. and Lim, H.-G.

June 2017 Transactions of the American Nuclear Society, vol. 116, no.

1, p. 808-811 17 Development of a Technical Basis for Component Integrity Assessment Using Probabilistic Methods Raynaud, P. and Kirk, M.

August 2018 ADAMS Accession No. ML18235A101 (*not publicly available) 18 Sequoyah SOARCA Uncertainty Analysis of a STSBO Accident Bixler, N. E. et al.

March 2018 Probabilistic Safety Assessment and Management Conference (PSAM 14), Los Angeles, CA, September 16-21, 2018 19 Using Deterministic and Probabilistic Methods for MELCOR Severe Accident Uncertainty Analysis Mattie, P. D. et al.

October 2017 ADAMS Accession No. ML17278A919

A-38 Title Author Date Journal/Technical Report/Conference Citation or Report Number 20 Findings from Uncertainty Studies Evaluating Severe Accident Phenomena and Off-Site Consequences Ghosh, S. T. et al.

October 2017 ADAMS Accession No. ML17278A920 21 Sensitivity Analysis for XLPR Acceptance Testing Sallaberry, C. J. et al.

July 2017 ASME 2017 Pressure Vessels and Piping Conference, Volume 6B: Materials and Fabrication, Waikoloa, Hawaii, USA, July 16-20, 2017

A-39 XIV.

Storage and Transportation Safety Title Author Date Journal/Technical Report/Conference Citation or Report Number 1

Example of Integration of Safety Security and Safeguard Using Dynamic Probabilistic Risk Assessment Under a System-Theoretic Framework Kalinina, E. A. et al.

April 2017 SAND2017-3572C 2

Advancing US Public Acceptance of Spent Fuel Storage and Transport:

Proposed Outreach Services for Ionising Radiation Education Support Pennington, C. W.

2013 Packaging, Transportation, Storage & Security of Radioactive Material, vol. 24, No. 3, p.95-107 3

Regulatory Analysis for Japan Lessons-Learned Tier 3 Issue on Expedited Transfer of Spent Fuel U.S. NRC November 2013 COMSECY 13-0030, ADAMS Accession No. ML13273A628 4

Analysis of Dose Consequences Arising from the Release of Spent Nuclear Fuel from Dry Storage Casks Durbin, S. G. and Morrow, C.

January 2013 SAND2013-0533 5

Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a US Mark I Boiling Water Reactor Barto, A. et al.

June 2013 NUREG-2161 ADAMS Accession No. ML14255A365