Operating Reactors Sub-Arena
The Nation's fleet of operating reactors comprises one of four sub-arenas that the staff of the U.S. Nuclear Regulatory Commission (NRC) identified in considering which areas of the reactor safety arena to target for greater use of risk information. This page summarizes the following aspects of the Operating Reactors Sub-Arena:
Make continuing, incremental improvements in rulemaking, licensing, and oversight of operating reactors, while focusing on implementing existing risk-informed and performance based activities.
This objective focuses on activities that are already in progress to risk-inform the operating reactor subarena, including completed rulemaking activities, guidance documents, and implementation of some initiatives.
The NRC will revisit and update this objective (as appropriate) once the industry has implemented the currently planned activities and feedback becomes available.
The risk-informed initiatives currently in progress were originally selected using screening criteria similar to those presented in the RPP. Consequently, the five activities (listed below) that support the goals for this subarena satisfy the following screening criteria:
- The risk-informed initiatives that are currently underway help to improve the effectiveness and efficiency of the NRC's regulatory process, including improved safety and reduction of unnecessary regulatory burden.
- Information and analytical models of operating reactors, particularly for at-power operations, exist and are fairly mature.
- The cost-beneficial nature of several of the risk-informed initiatives is evidenced by their voluntary adoption by licensees.
- No factors have been identified to date that would motivate changing the regulatory approach in the areas where risk-informed activities are already underway. Stakeholder feedback substantiates that there is no immediate need to initiate any new risk-informed initiatives, and that the NRC should focus on completing currently identified activities and allowing the industry time to implement those activities.
- Goals and activities to meet the objective for this subarena will be performance-based, to the extent that they meet the following four criteria:
- measurable parameters to monitor performance
- objective criteria to assess performance
- flexibility to allow licensees to determine how to meet the performance criteria
- no immediate safety concern as a result of failure to meet the performance criteria
Risk-informed activities for operating reactors occur in five broad categories:
- applicable regulations
- licensing process
- revised oversight process
- regulatory guidance
- risk analysis tools, methods, and data
The activities in these categories are derived from the Commission's policy statements and guidance, and include revisions to technical requirements in the regulations; risk-informed technical specifications; a new framework for inspection, assessment, and enforcement actions; guidance on other risk-informed applications (e.g., in-service inspections); and improved standardized plant analysis risk models.
The following goals are derived from the Commission's policy statements and guidance, which reflect the current phase of NRC and industry development, as well as the current implementation of risk-informed activities:
- Finish the development of current risk-informed regulations (e.g., 10 CFR 50.46a rulemaking) and associated regulatory/staff guidance.
- Implement existing NRC risk-informed activities [e.g., risk-informed technical specifications and pilots for 10 CFR 50.69 and the National Fire Protection Association (NFPA) Standard 805].
- Encourage the industry to implement risk-informed rules and approved/endorsed activities.
- Continue making incremental improvements to the established licensing, rulemaking, and oversight activities.
- Modify/update established activities to account for lessons learned.
List of Risk-Informed and Performance-Based Activities
This list shows the ongoing licensing initiatives, projects, and activities that the staff of the U.S. Nuclear Regulatory Commission (NRC) has targeted for greater use of risk information in the Operating Reactors Sub-Arena within the Reactor Safety Arena:
- Implementing Lessons Learned from Fukushima
- Risk-Informing Agency Actions on Low Risk Compliance Issues
- Probabilistic Flood Hazard Assessment (PFHA)
- Risk Informed Security Workshop
- Methods, Tools and Guidance for Including Digital Systems in Nuclear Power Plant PRAs
- Risk Assessment of Operation Events
- Maintenance and Development of the Systems Analysis Programs for Hands-on Analysis Integrated Reliability Evaluations (SAPHIRE) Code
- Standardized Plant Analysis Risk Models (SPAR)
- Full-Scope Site Level 3 PRA
- Data Collection for Human Reliability Analysis (HRA)
- Human Reliability Analysis (HRA) Methods and Practices
- Development of Human Reliability Analysis
- Develop Improved PRA Methods for Consequential Steam Generator Tube Rupture
- National Fire Protection Association (NFPA) Standard 805
- Assess Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance, Generic Safety Issue (GSI)-191
- Risk Prioritization Initiatives (RPI)
- Risk Informing Oversight of Emergency Preparedness (EP) and Response Plans
- Emergency Core-Cooling System (ECCS) Requirements: Redefinition of Loss-of-Coolant Accidents (LOCA)
- Emergency Core Cooling System (ECCS) Requirements: Loss of Coolant Accident and Loss of Offsite Power (ECCS-LOCA/LOOP)
- Develop Risk-Informed Improvements to Standard Technical Specifications (STS)
- Implement 10 CFR 50.69: Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
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Implementing Lessons Learned from Fukushima
Following the accident at the Fukushima Dai-ichi Nuclear Plant in Japan, the NRC initiated actions to evaluate lessons learned and to implement appropriate changes in nuclear power plant designs and procedures. Initial recommendations were included in the Near Term Task Force (NTTF) report entitled "Recommendations for Enhancing Reactor Safety in the 21st Century." Several of the items (e.g., Recommendation 1 regarding improving the regulatory framework and recommendation 2.1 on re-evaluating seismic and flooding hazards) include incorporation of risk-informed, performance-based approaches into NRC activities. The status and program plans for items identified for longer term evaluations were reported to the Commission in SECY 12-0095. Recommendation 1 was closed by the Commission without approving staff proposed improvement activities in SRM-SECY-13-0132. For NTTF recommendation 2.1—Seismic, some licensees are using a probabilistic seismic hazard approach in their responses to NRC's request for updated seismic hazard information. More information is available from the Japan Lessons Learned Web site.
FY 2015 Status
Licensees submitted updated seismic hazard information in FY 2014 and, if required, "expedited seismic evaluation process" results in FY 2015. The updated hazard information and other factors (e.g., risk insights from the Individual Plant Examination of External Events for Severe Accident Vulnerabilities) were used to determine whether certain plants need to perform a seismic risk assessment, (on the order of 20 sites screened in for performing the risk assessment.) For those sites, NRC will use that information as part of the determination of whether additional regulatory action is warranted.
Risk-informed licensing reviews (NFPA 805 reviews, R-I tech specs) The seismic hazard reevaluations are in response to a 10 CFR 50.54(f) letter, which seeks information to decide whether a licensee's license should be suspended, modified, or revoked.
Risk-Informing Agency Actions on Low Risk Compliance Issues
The agency is developing a risk-informed approach to resolve licensee compliance issues that are determined to be of low risk/low safety significance. The goal is to provide a tool to the staff that provides a risk-informed alternative to enforcement of technical specification compliance when it can be demonstrated that the non-compliance does not pose an undue risk to public health and safety.
The staff envisions developing a risk-informed process that would ensure that the level of licensee and staff resources applied to a non-compliance issue correlate to the potential risk and safety significance of the issue. The staff envisions that this approach would focus first on evaluating the risk significance of the non-compliance. If the risk significance is determined to be low, then the staff interaction with the licensee would focus on establishing a reasonable timetable for corrective action by the licensee combined with implementing appropriate interim compensatory measures that would maintain adequate safety while the corrective action is being taken. The approach would include enforcement discretion (possibly for a long duration) to provide the licensee adequate time for implementing corrective action. This approach is envisioned to be an improvement over the current practice in that it would eliminate the need for urgent action to be taken for low risk significance compliance issues.
This approach is consistent with the NRC's Enforcement Policy (NUREG 1600, "General Statement of Policy and Procedure for NRC Enforcement Action", Section 1.5 "Adequate Protection Standard," which states:
"Adequate protection of the public health and safety and assurance of the common defense and security and protection of the environment are the NRC's fundamental regulatory objectives. Compliance with NRC requirements plays a critical role in giving the NRC confidence that safety and security are being maintained. While adequate protection is presumptively assured by compliance with NRC requirements, circumstances may arise where new information reveals that an unforeseen hazard or security issue or security event exists or that a substantially greater potential exists for a known hazard to occur. In such situations, the NRC has the statutory authority to require action by licensees, their employees and contractors, and certificate holders above and beyond existing regulations to maintain the level of protection necessary to avoid undue risk to public health and safety, and to ensure security of materials.
"The NRC also has the authority to exercise discretion to permit continued operations—despite the existence of a noncompliance—where the noncompliance is not significant from a risk perspective and does not, in the particular circumstances, pose an undue risk to public health and safety. When noncompliance with NRC requirements occurs, the NRC must evaluate the degree of risk posed by that noncompliance to determine whether immediate action is required. If the NRC determines that the noncompliance itself is of such safety significance that adequate protection is no longer provided, or that the noncompliance was caused by a failure of licensee controls so significant that it calls into question the licensee's ability to ensure adequate protection, the NRC may demand immediate action, up to and including a shutdown or suspension of licensed activities. Based on the NRC's evaluation of noncompliance, the appropriate action could include refraining from taking any action, taking specific enforcement action including the use of civil penalties, issuing Orders, or providing input to other regulatory actions or assessments, such as increased NRC oversight of a licensee's activities. Since some requirements are more important to safety than others, the NRC endeavors to use a risk-informed approach when applying NRC resources to the oversight of licensed activities, including enforcement activities."
FY 2015 Status
A working group with members from NRR, the Regions, OGC, and OE has formed, and is currently evaluating the feasibility of the proposed approach, including verifying the legality of the approach determining how the risk significance would be evaluated, and gaging the industry's interest in participating in the process once developed. The working group is also looking at the process for implementing this new approach. One implementation method under consideration is modifying the Notice of Enforcement Discretion (NOED) process for low risk compliance issues.
"Risk-Informed Oversight Activities" – The purpose of this activity is to provide the staff with a risk-informed tool for handling compliance issues that are of low risk significance.
Probabilistic Flood Hazard Assessment (PFHA)
The Commission was briefed in January, 2014, on staff (NRR, NRO, and RES) activities and plans concerning the need for and development of a systematic program to establish a probabilistic approach for flood hazard assessment. Near term aspects of the program will address information needs in the reactor oversight program for reviews of operating reactors while the long term program will develop a comprehensive approach for probabilistic flood assessment for new reactors. The offices agreed on a joint user need that endorsed a Research Plan developed jointly by RES, NRR, and NRO staff. A copy of the plan (cover sheet and final plan) was provided to the Commission. RES has begun implementation of the research plan.
FY 2015 Status
The "Probabilistic Flood Hazard Assessment Research Plan" has been prepared and endorsed by NRR and NRO. Eleven new research projects have been initiated with the US Army Corps of Engineers, the US Geological Survey, the Department of Interior Bureau of Reclamation, Idaho National Laboratory (INL), Pacific Northwest National Laboratory (PNNL), and the University of California at Davis. A twelfth research activity that was issued for bid as a commercial contract has not yet been awarded. On October 13 and 14, 2015, the first annual program review on the progress for these projects will be held at NRC headquarters. Cooperative efforts are under development with Electric Power Research Institute (EPRI) and the Institute de Sûreté Nucléaire et de Radioprotection (IRSN).
"Risk Tool, Maintenance & Development" – The purpose of this activity is to develop a comprehensive approach for probabilistic flood assessment for new reactors.
Risk Informed Security Workshop
The staff has worked with several organizations to conduct a series of workshops on risk informed security to identify potential opportunities to better risk inform security. The initial workshop was hosted by Sandia National Laboratories in September 2010. Based on the results of the 2010 workshop a follow-on workshop was hosted by the Institute of Nuclear Materials Management (INMM) in Stone Mountain, GA in February 2014.
These workshops covered a broad range of topics in order to identify areas where further research might be conducted to improve the way that risk information is incorporated into the security regulatory process at the Nuclear Regulatory Commission. These workshops have been coordinated with the Departments of Energy and Homeland Security and the National Security Council. In addition, they have been attended by representatives of a number of foreign governments.
FY 2015 Status
Three workshops were conducted in FY 2015. The first workshop was hosted by the INMM and the George Washington University Elliott School of International Affairs in Washington, DC in March 2015. Risk informed security topics included cyber security, perception of risk and insider mitigation. The second workshop was hosted by the INMM and the American Nuclear Society (ANS) in conjunction with the ANS International Topical Meeting on Probabilistic Safety Assessment and Analysis in Sun Valley, ID in April 2015. This workshop brought safety and security risk professionals together to discuss how risk information is used within the two disciplines. The final workshop of the year was hosted by the INMM in Boston in September 2015, entitled the Vulnerability Assessment Tools Workshop.
Risk-Informed licensing and oversight activities.
Methods, Tools and Guidance for Including Digital Systems in Nuclear Power Plant PRAs
The NRC has been investigating reliability modeling of digital systems, which encompasses both hardware and software. The objective of this research is to identify and develop methods, analytical tools, and regulatory guidance for (1) including models of digital systems in nuclear power plant probabilistic risk assessments (PRAs) and (2) incorporating digital systems in the NRC's risk-informed licensing and oversight activities.
FY 2015 Status
Recent accomplishments and near-term objectives include the following:
- NRC support to the development of a failure mode taxonomy for a digital instrument and control (I&C) systems performed by the OECD/NEA Working Group on Risk Assessment (WGRISK) (NEA/CSNI/R(2014)16, "Failure Modes Taxonomy for Reliability Assessment of Digital I&C Systems for PRA").
- In collaboration with the Korea Atomic Energy Research Institute, the staff developed an approach for quantifying software reliability using a Bayesian Belief Network (BBN)-based model of the software development cycle quality attributes. A report describing the BBN approach will be submitted for publication in FY2016.
- Pilot an approach for estimating the reliability of the INL Advanced Test Reactor Loop Operating Control System using PRA-based statistical testing. A report describing the statistical testing application will be submitted for publication in FY2016.
More background on this approach can be found in the transcripts from an ACRS subcommittee meeting held in November 2014.
Risk Tool, Maintenance & Development. The objective of this activity is to develop methods for incorporating digital instrumentation and control (I&C) systems into nuclear plant PRAs. As the activity proceeds, additional insights on the practicality and usefulness of including digital systems in nuclear plant PRAs will be gained.
Risk Assessment of Operation Events
Maintain an integrated handbook and analysis methods support for staff to analyze internal, external, and low-power/shutdown operation events.
Support: The support provided is in response to requests from the program offices (NRR, NRO, and the Regions) and on-call technical assistance to senior reactor analysts (SRAs) and other practitioners of risk analysis when needed with PRA models or the risk analysis software. Specifically, staff implementing programs in the following areas are frequent recipients of this technical assistance: MD 8.3 Incident Investigation Program, Reactor Oversite Process and the Significance Determination Process, Accident Sequence Precursor Program, and the Risk Assessment Standardization Project (RASP) help desk.
Handbook: A Risk Assessment Handbook and associated Web site provides methods and guidance that NRC staff use to achieve more consistent results when performing risk assessments of operational events and licensee performance issues. It is updated periodically based on user comments and insights gained from field application. The Handbook consists of four volumes, designed to address internal events analysis, external events analysis, Standardized Plant Assessment Risk (SPAR) model reviews, and shutdown event analysis. The Handbook represents best practices based on feedback and experience from the analyses of precursors in the Accident Sequence Precursor (ASP) Program and numerous Significance Determination Process (SDP) Phase 3 analyses.
FY 2015 Status
This activity continually provides support to risk analysts and routinely updates the RASP Handbook and the associated Web site to assure accuracy and provide additional references for risk analysts' use.
Risk tool maintenance & development. Maintaining analysis tools and formal guidance associated with risk analysis supports risk informed decision making by the staff.
Maintenance and Development of the Systems Analysis Programs for Hands-on Analysis Integrated Reliability Evaluations (SAPHIRE) Code
The NRC has developed and maintains the SAPHIRE (Systems Analysis Programs for Handson Integrated Reliability Evaluation) computer code for performing probabilistic risk analyses (PRAs). SAPHIRE offers state-of-the-art capability for assessing the risk associated with core damage frequency (Level 1 PRA) and the risk from containment performance and radioactive releases (Level 2 PRA). SAPHIRE supports the agency's risk-informed activities, which include the Standardized Plant Analysis Risk (SPAR) model development plan, the risk assessment standardization project, the Significance Determination Process (SDP), risk-informing 10 CFR Part 50, vulnerability assessment, advanced reactors, operational experience, generic issues, and regulatory backfit.
FY 2015 Status
An updated status of SAPHIRE computer code activities can be found in SECY 15-0124, "Status of the Accident Sequence Precursor Program and the Standardized Plant Analysis Risk Models."
Risk Tool, Maintenance & Development. The SPAHIRE computer code is used to develop and run PRA models (e.g., SPAR models) for a variety of regulatory applications.
Standardized Plant Analysis Risk Models (SPAR)
The SPAR models provide agency risk analysts with an independent risk assessment tool to support a variety of risk-informed agency programs, including the Reactor Oversight Program (ROP) and the Accident Sequence Precursor (ASP) program. SPAR models are built with a standard modeling approach, using consistent modeling conventions, that enables staff to easily use the models across a variety of U.S. NPP designs. Unlike industry PRA models, SPAR models are run on a single software platform, the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) computer code. The staff currently maintains and updates the 75 SPAR models representing 99 commercial NPPs. The scope of every SPAR model includes logic modeling covering internal initiating events at power through core damage (i.e., Level-1 PRA model). A portion of the SPAR models also include external hazard (e.g., seismic and high wind), internal fire, and shutdown models. The staff develops and maintains SPAR models for both operating reactors and new reactor designs (e.g., AP1000).
FY 2015 Status
An updated status of the SPAR model program can be found in SECY 15-0124, "Status of the Accident Sequence Precursor Program and the Standardized Plant Analysis Risk Models."
"Risk Tool, Maintenance & Development" – The purpose of this activity is to develop standardized risk analysis models and tools for staff analysts to support various regulatory activities, including the Accident Sequence Precursor (ASP) Program and Phase 3 of the Significance Determination Process (SDP as described in Inspection Manual Chapter 0609).
Full-Scope Site Level 3 PRA
As directed in SRM-SECY-11-0089, "Options for Proceeding with Future Level 3 Probabilistic Risk Assessment (PRA) Activities," the staff is conducting a full-scope multi-unit site Level 3 PRA that addresses all internal and external hazards; all plant operating modes; and all reactor units, spent fuel pools, and dry cask storage.
The full-scope site Level 3 PRA project includes the following objectives:
- Develop a Level 3 PRA, generally based on current state-of-practice methods, tools, and data, that (1) reflects technical advances since completion of the NUREG-1150 studies, and (2) addresses scope considerations that were not previously considered (e.g., low power and shutdown, multi-unit risk, and spent fuel storage).
- Extract new risk insights to enhance regulatory decision making and help focus limited agency resources on issues most directly related to the agency's mission to protect public health and safety and the environment.
- Enhance PRA staff capability and expertise and improve documentation practices to make PRA information more accessible, retrievable, and understandable.
- Obtain insight into the technical feasibility and cost of developing new Level 3 PRAs.
Consistent with the objectives of this project, the Level 3 PRA study is based on current state of-practice methods, tools, and data. However, there are several gaps in current PRA technology and other challenges that require advancement in the PRA state-of-practice. The general approach to addressing these challenges for the Level 3 PRA study is to primarily rely on existing research and the collective expertise of the NRC's senior technical advisors and contractors, and to perform limited new research only for a few specific technical areas (e.g., multi-unit risk).
Based on a set of site selection criteria and with the support of the NEI, Southern Nuclear Operating Company's Vogtle Electric Generating Plant, Units 1 and 2, was selected as the volunteer site for the Level 3 PRA study. The Level 3 PRA project team is leveraging the existing and available information on Vogtle and its licensee PRAs, in addition to related research efforts (e.g., SOARCA), to enhance efficiency in performing the study.
The Level 3 PRA project team is using the following NRC tools and models for performing the Level 3 PRA study:
- SAPHIRE, Version 8.
- MELCOR Severe Accident Analysis Code.
- MELCOR Accident Consequence Code System, Version 2 (MACCS).
In addition, the Level 3 PRA study is being developed consistent with many of the modeling conventions used for NRC's SPAR models.
FY 2015 Status
A PWR Owners Group (PWROG)-led ASME/ANS PRA Standard-based peer review of the reactor, at-power, high wind, Level 1 PRA and a screening evaluation of reactor, at-power "other" hazards (i.e., hazards other than internal events, internal floods, internal fires, high winds, and seismic events) was performed in November 2014. A PWROG-led ASME/ANS PRA Standard-based peer review of the reactor, at-power, internal event and internal flood Level 2 PRA was performed in December 2014. A PWROG-led workshop was held in January 2015 to identify peer review criteria for dry cask storage PRA. An expert elicitation was completed in June 2015 to address the frequency of interfacing systems LOCAs. The reactor, at-power, internal event and internal flood Level 3 PRA was completed in August 2015 and its peer review will be completed in October 2015. Initial versions of reactor, at-power, Level 1 PRA models for internal fires and seismic events were completed in FY 2015, but they are in the process of being significantly revised to incorporate more recent licensee-supplied information.
Infrastructure Development in Support of Risk-Informed Regulations. The Commission directed Level 3 PRA Project is not supporting a specific risk-informed regulatory application. However, as described in SECY 12-0123, "Update on Staff Plans to Apply the Full-Scope Site Level 3 PRA Project Results to the NRC's Regulatory Framework," the results and insights of the Level 3 Project are expected to benefit a variety of ongoing regulatory initiatives.
Data Collection for Human Reliability Analysis (HRA)
Consistent with the Commission's policy statements on the use of PRA and for achieving an appropriate PRA technical adequacy for NRC risk-informed regulatory decision-making, the NRC has ongoing activities to improve the technical basis for HRA. The adequacy of data available for HRA is an issue for the adequacy and consistency of human error probability estimates. To address this need, the staff has signed an agreement with the STP Nuclear Operating Company (STPNOC) to collect the STPNOC's licensed operator simulator training data for HRA; an amendment to this agreement was signed on September 2013 to extend the agreement until June 2018). The staff has developed the Scenario Authoring, Characterization, and Debriefing Application (SACADA) database application to support the agreement. The SACADA system was developed to collect licensed operator simulator data to inform HRA and to improve operator training programs. Since a data collection pilot study in May 2012, the STPNOC has used the SACADA system for its operator training program to collect licensed operator simulator training data and has shared the data with the NRC.
A database is useful only if it has sufficient data and the data are informative. The staff has worked to achieve the following two objectives: increasing the number of data providers and evaluating the data effectiveness in informing HRA. In the past year the following international organizations have collaborated with the NRC on the use of and the evaluation of the SACADA system:
- The NRC signed an agreement in September, 2013 with the Korea Atomic Power Research Institute (KAERI) on HRA data research. KAERI researchers use the SACADA system to collect and analyze Korean nuclear power plants' operator simulator exercise data to evaluate the SACADA system.
- The Halden Reactor Project (HRP) has used the SACADA system to collect the operator simulator experiment data generated at the HRP's Halden Human Machine Laboratory (HAMMLAB) since June 2014.
- A bilateral agreement between the NRC and the nuclear research institute, ÚJV Řež, a. s., of the Czech Republic, was signed in February 2015 to promote collaboration on HRA data collection.
- A work item on HRA data collection was added as a working item of the TECRO-AIT (Taiwan Economic and Culture Representative Office – American Institute in Taiwan) Joint Standing Committee on Civil Nuclear Cooperation (JSCCNC) in November, 2014. As such, an agreement between TECRO and AIT (designated to Taiwan Power Company (TPC) and the NRC) is in process to collect TPC's operator simulator training data using the SACADA system.
- The staff held an HRA data workshop in April 2015 with domestic and international participants to discuss the experience in using the SACADA system and the use of SACADA data for HRA. In this workshop, participants demonstrated several methods of using SACADA data to inform HRA, using the roughly 8,000 data points collected in the SACADA database from three sources (plant operator simulator training and two simulator experiments). The results showed positive indications on using SACADA data for HRA method improvement.
- The staff presented the acquisition of SACADA data for HRA method improvement to the Advisory Committee on Reactor Safety Subcommittee on Reliability and Probabilistic Risk Assessment. The subcommittee members provided positive feedback about the SACADA system and recommended that the staff continue its data analysis program to further enhance the use of simulator data for HRA method improvement.
- The NRC continues its outreach to domestic nuclear power stations and nuclear industry human performance stakeholders for HRA data collection.
FY 2015 Status
The key near term SACADA research activities include:
- Analyzing the collected data to inform human reliability and human performance. This includes demonstrating the use of the data to inform human error probability (HEP) calculations in HRA.
- Collaborating with more data providers to increase the size of the data pool.
Risk Tool, Maintenance and Development. Improvement of HRA methods and data collection will improve the quality of analyses.
Human Reliability Analysis (HRA) Methods and Practices
The purpose of the HRA method effort is to improve the methods for regulatory applications. This improvement involves increasing the consistency amongst HRA practitioners in the use of methods and developing guidance on the rigor needed for quantifying human reliability given the scarcity of empirical data to evaluate human performance. The ongoing activities include:
- Develop the IDHEAS General Methodology for risk analyses of all NPP HRA applications (SRM-M061020)
- NRC/EPRI collaborative effort to implement the General Methodology for internal at-power application
- Implement the General Methodology for Ex-Control Room actions
Risk Basis: Regulatory Guide (RG) 1.200 provides an acceptable approach for determining the technical adequacy of PRA results for risk-informed activities. HRA is a key element in the PRA; different HRA methods often have different assumptions and approximations and, therefore, may yield different results. Thus, improving HRA methods enhances the consistency and quality of HRA and PRA.
FY 2015 Status
The report "Cognitive Basis for HRA" is finalized and will be published in 2015. The staff has been working with the ACRS Reliability and PRA Subcommittee to construct the IDHEAS General Methodology so that it can be implemented in various NPP applications. The IDHEAS internal, at-power application is currently being tested.
Risk-Informed oversight activities. The purpose of the HRA method efforts is to improve the method to be used for regulatory applications and consistency among HRA practitioners in performing HRA. This will help improve HRA / PRA quality and provide a basis for risk-informed rule-making.
Development of Human Reliability Analysis
Regulatory Guide (RG) 1.200 provides an acceptable approach for determining the technical adequacy of PRA results for risk-informed activities. However, RG 1.200 (including the PRA standards reflected and endorsed by RG 1.200) is a high-level regulatory guide, addressing what to do but not the how to do it. Consequently, there may be several approaches for addressing certain analytical elements, which may meet the RG 1.200 and associated standards, but may do so by making different assumptions and approximations and, therefore, may yield different results. This is particularly true for human reliability analyses (HRA) for which many methods are available to model mitigative actions in PRAs. The staff is addressing this issue by developing lower level guidance documents to support the implementation of RG 1.200.
This work supports the NRC's action plan for stabilizing PRA quality expectations and requirements (described in SECY-04-0118 and SECY-00-0007). It also is responsive to the November 8, 2006, staff requirements memorandum (SRM) (SRM-M061020) in which the Commission, based on ACRS concerns, directed the staff to evaluate different human reliability models in an effort to propose a single model for the agency to use or guidance on which model(s) should be used in specific circumstances." The following activities are addressing HRA improvement needs:
(1) HRA Method Benchmarking:
- Participate in the International HRA Empirical Study in an effort to benchmark HRA methods by comparing HRA predictions to crew performance on a nuclear power plant simulator.
- The International HRA Empirical study was a multinational multi-team effort supported by the Organization for Economic Co-Operation and Development (OECD) Halden Reactor Project. The Halden Reactor Project provided facilities, crews, and expertise to collect and analyze simulator crew performance data and HRA analyst teams from multiple organizations used their preferred HRA methods to analyze and predict the performance of these crews. The objective of the study was to develop an empirically-based understanding of the performance, strengths, and weaknesses of the various HRA methods used to model human response to accident sequences in probabilistic risk assessments (PRAs).
- This study was the first of its kind; no previous HRA benchmarking studies have been performed using crew simulator data. Its pilot phase has been documented in NUREG/IA-0216, Vol.1, "International HRA Empirical Study – Phase 1 Report, Description of Overall Approach and First Pilot Results from Comparing HRA Methods to Simulator Data," November 2009 (Halden report HWR-844). Its second phase consisted of the analysis and comparison of HRA predictions for nine steam generator tube rupture (SGTR) human actions and is documented in NUREG/IA, Vol. 2, "International HRA Empirical Study – Phase 2 Report, Results from Comparing HRA Method Predictions to Simulator Data from SGTR Scenarios," (Halden report: HWR-915), August 2011. Phase 3 consisted of the comparison of four loss-of-feedwater (LOFW) human actions and will be documented in NUREG/IA Vol.3, "The International Empirical Study – Phase 3 Report – Results from Comparing HRA Method Predictions to Simulator Data from LOFW Scenarios," (Halden report HWR-951), published in 2014.
- The overall findings of the study were documented in NUREG-2127 (HWR-373), entitled "The International HRA Empirical Study – Final Report – Lessons Learned from Comparing HRA Methods Predictions to HAMMLAB Simulator Data," published in August 2014. The results of the Empirical Study will provide a technical basis for improving individual methods, improving existing guidance documents for performing and reviewing HRAs (e.g., NUREG-1792, HRA Good Practices), and developing additional guidance and training materials for implementing individual methods.
- The study has also produced many conference papers, presented at the annual Institute of Electrical and Electronics Engineers Conference on Human Factors, August 2007, at the American Nuclear Society International Probabilistic Safety Conference 2008 (PSA8), September 2008, and at the International Conference on Probabilistic Safety Assessment and Management (PSAM) conferences: PSAM9, May 2008, PSAM10, May 2010, and PSAM 11, June 2012.
(2) HRA Method Improvement Using US Simulator Runs:
- As noted above, RES participates in and supports the International HRA Empirical Study to benchmark HRA models by comparing HRA results to empirical data generated through crew simulator runs. The International HRA Empirical Study has clearly identified important strengths and weaknesses of the various methods and identified areas for improvement in HRA methods and practices. In particular, an important conclusion from the study was that improving the qualitative analysis aspects of HRA methods could increase their robustness and reduce some of the sources in the variability of results that are seen in applications of different methods. However, since there was only one case in the International study where the same HRA method was applied by different teams, it was difficult to clearly separate method specific effects from differences created by the analysts' application of a given method. Thus, in addition to examining differences across methods, a major objective of the US simulator study (performed on a US nuclear power plant simulator) is to test the consistency and accuracy of HRA predictions among different analyst teams using the same methods. A particular area of interest in these comparisons is examination of the qualitative analysis performed by different methods and teams to identify shortcomings that contribute to inconsistencies in results and to determine the extent to which the shortcomings are due to analyst differences or due to inherent shortcomings in the methods.
- Two other potential limitations of the International study are also addressed in the US study:
- First, in the International study, the HRA teams were unable to visit the Halden simulator and collect HRA related information through interviews with plant operators and trainers and through observations of actual operating crews in the simulator, as is typically done in performing an HRA for a nuclear power plant PRA. This type of information was provided to the HRA teams to the extent possible by the study team in the International study and the HRA teams were allowed to submit written questions that were answered by the study team and plant personnel as needed. Some of the HRA teams in the International study felt this significantly limited their ability to perform an adequate HRA. In the US study, the HRA teams were able to visit the reference plant and collect information relevant to performing their HRA as it would normally be done in a PRA.
- Second, there was some concern that because the international study was based on the results of simulator runs using European crews at the Halden Reactor Project, the results might not be directly generalizable to what would occur with US nuclear power plant crews. Some of the HRA teams in the international study thought that their expertise was more geared to understanding what US crews would do and that their US bias may have influenced their decision-making in applying their HRA method. Thus, the US study would serve as a check against the effects of such bias on the results.
- In SRM-M090204B, dated February 18, 2009, the Commission directed the staff "to continue to pursue possibly working with EPRI, INPO, and/or international partners to test U.S. nuclear plant operating crews' performance in a variety of situations and keep the Commission informed on the progress in developing a human reliability analysis (HRA) database and benchmarking projects." Thus, the US Empirical Study is directly responsive to this SRM.
- To perform the US Empirical Study, RES established an MOU with a US utility volunteering to participate in this study by offering simulator facilities, crews and expertise to support the design and execution of the experimental runs. Further, RES initiated work with the objective to evaluate HRA methods currently used in regulatory applications through a comparison of HRA predictions to crew performance in simulator experiments performed at the US nuclear power plant. The Halden Reactor Project, Idaho National Laboratory, Sandia National Laboratories, Paul Scherrer Institute (Switzerland) and EPRI are also participating and supporting the study.
- To accomplish the goals of the study, 4 crews from the plant performed 3 different scenarios: 1) a Loss of Feedwater followed (after recovery of feedwater) by a Steam Generator Tube Rupture, 2) a Loss of Component Cooling Water and Reactor Coolant Pump seal water, and 3) a basic Steam Generator Tube Rupture. Crew performance on several human failure events that would normally be modeled in a PRA was evaluated and compared with the predictions from 9 human reliability analysis teams using 4 different methods (ATHEANA, SPAR-H, EPRI Calculator, and ASEP/THERP). Both qualitative and quantitative predictions are being evaluated.
- A workshop was hosted by the NRC in June, 2011 to discuss the preliminary results from the study with the study participants. Based on the input from the workshop participants, the data analysis is continuing. Plans are to complete a NUREG documenting the experimental design, results, and conclusions of the US study and another NUREG discussing the overall conclusions and lessons learned from both the International and US studies in 2015.
- The results will be used to:
- Assess the impact of potential limitations in the data collected in the International Empirical Study as described above.
- Provide an improved basis for determining how to best improve HRA methodology and use this information as an input to the HRA Model Differences Project (Item 3 below).
(3) Address HRA Model Differences:
- Many models are available for HRA. There is evidence that the results associated with a particular human failure event analysis could vary depending on the HRA model/method used and/or the analyst applying the method. Because HRA results and insights are frequently used to support risk-informed regulatory decision making, the NRC continues to improve the robustness of PRA/HRA through targeted activities (e.g., supporting and endorsing PRA standards developed by professional societies). Recognizing that HRA model differences contribute to the variability of PRA/HRA results, the Commission directed the Advisory Committee on Reactor Safeguards (ACRS) (SRM-M061020) to work with the staff and external stakeholders to evaluate the different human reliability models and either propose a single model for the agency to use or guidance about which model(s) should be used for the different regulatory applications.
- The Office of Nuclear Regulatory Research (RES) has taken the lead in addressing SRM-M061020. The ACRS has kept abreast of developments and provides input through periodic meetings. This work is performed collaboratively with the Electric Power Research Institute (EPRI) under a RES/EPRI Memorandum of Understanding and its update.
- The main tasks of this work include: (1) Identification of current and emerging regulatory applications in which HRA results could have an impact on the decision; (2) identification and evaluation of currently available methods for their suitability and adequacy to treat human performance issues associated with the various regulatory applications and domains of interest (e.g., event analysis for shutdown operations); (3) development a cognitive foundation for HRA through synthesizing literature on why human makes errors (4) development of a generic HRA methodology based on the cognitive foundation for all NPP HRA applications; and 5) development of a HRA method for internal, procedural HRA analysis that that integrates the strengths of the existing HRA models into a unified HRA structure and have new components to address the key limitations in current models. The new method is referred to as "Integrated Human Event Analysis System (IDHEAS)."
- The development of the new HRA method needs to go through all stages of new model development: (1) developing a technical basis to understand human performance under accident situations from cognitive sciences and operational experience; (2) constructing a method for analyzing human performance and estimating human error probabilities supported by the technical basis; (3) developing tools for using the model; (4) reviewing and testing the work; (5) documenting the results and the development process; and (6) producing training materials and user guides. The staff is currently engaged in the development of the method, sought to be ready for pilot testing and revision in the end of 2015. The staff expects to complete the material development by December 2015.
- The staff believes that these efforts will result in producing a HRA method that is well understood and appropriately characterized for its suitability and usefulness in different regulatory applications.
- Primary Priority: High
- Secondary Priority: Medium
Project Considerations: The HRA guidance addresses many issues associated with the use of HRA in decision-making, including the suitability of an individual method to a regulatory application, consistency among HRA practitioners in implementing HRA methods, and the absence of guidance on the rigor needed for quantification of human reliability.
FY 2015 Status
Risk tools maintenance and development.
Develop Improved PRA Methods for Consequential Steam Generator Tube Rupture
Consequential steam generator tube ruptures (C-SGTRs) are potentially risk-significant events because thermally-induced steam generator tube failures caused by hot gases from a damaged reactor core can result in a containment bypass event and a large release of fission products to the environment. The main accident scenarios of interest are those that lead to core damage with high reactor pressure, a dry-steam generator, and low steam generator pressure (high-dry low) conditions. A typical example of such an accident scenario is a station blackout with loss of auxiliary feedwater. The objective of this program is to develop a simplified methodology for the quantitative assessment C-SGTR probability and large early-release frequency (LERF) for pressurized-water reactors (PWRs). A draft report was updated using the latest thermal hydraulic MELCOR results for Combustion Engineering (CE) plants.
FY 2015 Status
A draft report is being finalized to document the research results from this study. It is expected that the report will be issued for public review and comment in late calendar year 2015 and finalized in 2016. This work was presented to the ACRS Metallurgy and Reactor Fuels Subcommittee on April 7, 2015. A draft version of the report was provided to the ACRS.
Risk Tool Maintenance & Development. This work is intended to develop an enhanced risk assessment tool for assessing C-SGTR. A key focus of the work is closing technical gaps associated with thermal-hydraulic and structural analyses, assessment of SG flaw distributions, and PRA modeling. The risk insights obtained and process and tools developed can be used to support operating reactor and new reactor risk assessment.
National Fire Protection Association (NFPA) Standard 805
In 2004, the Commission approved a voluntary risk-informed and performance-based fire protection rule for existing nuclear power plants. The rule endorsed NFPA consensus standard NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants." In addition, the NEI developed NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," dated September 2005. The staff endorsed NEI 04-02 in RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," issued in May 2006. To date, nearly half of the nuclear power units operating in the United States, including those that participated in the pilot program, have committed to transition to NFPA 805 as their licensing basis. The Oconee Nuclear Station (Oconee) and Shearon Harrison Nuclear Power Plant (Shearon Harris) were the pilot plants for 10 CFR 50.48(c). In June 2010, a safety evaluation approved the Shearon Harris NFPA 805 pilot application. A safety evaluation in December 2010 approved the Oconee NFPA 805 pilot application. NEI 04-02 was revised (Revision 2) in April 2008 and the staff revised RG 1.205 (Revision 1) in December 2009 to reflect lessons learned from the pilot reviews. The staff developed NUREG-800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 9, "Auxiliary Systems," Section 18.104.22.168, "Risk-Informed, Performance-Based Fire Protection Program Review Responsibilities," issued December 2009, to provide staff guidance for the review of licensee applications to transition to NFPA 805. In addition, the NRC developed a Frequently Asked Question process to review and establish a preliminary staff position on NFPA 805 application, review, and implementation issues.
Lessons learned from the pilot applications indicated that the staff and the industry underestimated the complexity and resources necessary to complete the reviews. In SRMSECY-11-0033, "Proposed NRC Staff Approach to Address Resource Challenges Associated with Review of a Large Number of NFPA 805 License Amendment Requests," dated April 20, 2011, the Commission approved the staff's recommendation to increase resources to review NFPA 805 applications, develop a staggered review process, and modify the current enforcement policy. The NRC sent the revised enforcement policy to the Commission in SECY-11-0061, "A Request to Revise the Interim Enforcement Policy for Fire Protection Issues on 10 CFR 50.48(c) to Allow Licensees to Submit License Amendment Requests in a Staggered Approach," dated April 29, 2011 and approved in SRM-SECY-11-0061, dated June 10, 2011. To enhance the efficiency and effectiveness of the NFPA 805 application reviews, the industry developed an application template and the staff developed a safety evaluation template. The staff has received 26 applications to date and expects another two by the end of calendar year 2016. Additional information is available.
FY 2015 Status
In 2015, the NRC staff issued six non-pilot NFPA 805 license amendments with three more expected to be completed by the end of the year. Thirteen LARs are currently under review. The current status and update of work, dated June 2015, is available.
Risk-Informed Licensing Reviews. NFPA 805 is a performance-based standard, endorsed via 10 CFR 50.48(c) that critically depends on risk information in the form of Fire PRA to enable licensees to transition from existing "deterministic" fire protection programs to ones that are "risk-informed, performance-based." Fire PRA is an integral part of the new licensing basis, and includes both quantitative evaluations of base risk and changes to base risk in accordance with RG 1.174 guidelines as well as supporting qualitative considerations, such as traditional defense in depth and safety margin, also as per RG 1.174.
Assess Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance, Generic Safety Issue (GSI)-191
This generic issue concerns the possibility that following a LOCA in a PWR, debris accumulation on the containment sump strainer(s) may inhibit flow to the Emergency Core Cooling System (ECCS) and the Containment Spray System. An additional concern is that debris may penetrate or bypass the strainer and block flow to the core.
In SECY-12-0093 dated July 9, 2012, the staff identified several options for resolving GSI-191. These options included two risk-informed approaches. One approach, piloted by South Texas Project (STP), would address both strainer and in-vessel effects using risk. The other approach would use risk for in-vessel effects and would resolve strainer issues deterministically.
The Commission endorsed the staff's proposed options for resolving GSI-191 in SRM-SECY12-0093, dated December 14, 2012. As part of the resolution process, licensees seeking additional time to pursue new testing or new approaches (including risk-informed approaches) will implement compensatory measures to mitigate the potential for debris blockage of the strainer or reactor core. Industry is also performing additional testing to support risk-informed evaluations for GSI-191.
Tentatively, 14 units propose to implement a full risk-informed resolution to GSI-191. Two units plan to risk-inform the in-vessel evaluation and use a deterministic evaluation for the ECCS strainer.
SRM-SECY-12-0034, "Proposed Rulemaking – 10 CFR 50.46c: Emergency Core Cooling System Performance During Loss-of-Coolant Accidents RIN 3150-AH42," dated January 7, 2013 directed that the "the 50.46c proposed rule should contain a provision allowing NRC licensees, on a case-by-case basis, to use risk informed alternatives without an exemption request." The proposed rule containing this provision was published on March 24, 2014.
Per SRM-COMSECY-13-006, "10 CFR 50.46c Rulemaking: Request to Defer Draft Guidance and Extension Request for Final Rule and Final Guidance," dated May 9, 2013, the draft guidance related to the GSI-191 risk-informed alternative was not published concurrent with the proposed rule. Rather, that draft guidance (DG-1322, "Alternative Risk-Informed Approach for Addressing the Effects of Debris on Post Accident Long-Term Core Cooling") was developed in parallel with the staff's review of the STP pilot and was issued for public comment on April 20, 2015. The public comment period closed on July 6, 2015. The final regulatory guide (RG 1.229 (DG-1322), "Risk-Informed Approach for Addressing the Effects of Debris") will be issued with the final 10 CFR 50.46c rule. Additional information is available.
FY 2015 Status
In FY 2015, the staff has continued to review the STP pilot and has published draft guidance (DG-1322) for licensees choosing to implement the optional, risk-informed provision in 10 CFR 50.46c. the draft guide (which will ultimately be published as RG 1.229) was issued for public comment on April 20, 2015. The public comment period closed on July 6, 2015, and the staff has since resolved all public comments and updated the DG accordingly. RG 1.229 is scheduled to be issued with the new 10 CFR 50.46c rule in the second quarter of FY 2016.
Rulemaking Applications Using Risk Insights. The purpose of this activity consistent with Commission direction is to provide a provision in 10 CFR 50.46c that would allow a risk informed treatment of debris when assessing long-term core cooling.
Risk-Informed Licensing Reviews. The purpose of this activity is to perform a risk-informed review of the STP pilot License Amendment Request (LAR) using the guidance in RG 1.174 and SRP Chapter 19. The STP pilot has informed the development of 10 CFR 50.46c and DG1322.
Risk Prioritization Initiatives (RPI)
In February 2013, the Commission approved SRM-COMGEA-12-0001/COMWDM-12-0002, "Proposed Initiative to Improve Nuclear Safety and Regulatory Efficiency", to further explore the idea of enhancing nuclear safety and regulatory efficiency by applying PRA. This initiative could encourage the use and development of high quality, plant-specific PRA models by allowing licensees to use qualitative and quantitative risk insight to propose a schedule for implementing regulatory actions on a plant-specific basis.
In October 2013, NEI began to develop a draft process as a potential way to address RPI for operating power reactors. The NEI's draft process consists of three main elements: (1) generic prioritization by an industry generic assessment expert team, (2) plant-specific prioritization by an integrated decision-making panel of licensee experts, and (3) issue aggregation for plant specific scheduling. The NRC staff provided comments on NEI's guidance. The guidance described the process at various stages using insights gained from tabletop exercises and discussions with stakeholders during public meetings.
Subsequently, the NRC staff informed the Commission about its observation of tabletop exercises of the NEI draft process in COMSECY-14-0014. Afterwards, six licensees also participated in the industry-led demonstration pilots that were conducted between May and September of 2014 to exercise the draft guidance prioritizing plant-specific issues. Lastly, a public meeting in September 2014 was held to further exercise the process in the areas of security, emergency preparedness, and radiation protection.
Other information about the NRC staff's observations can be found in "Summary of the NRC Staff Observations on the Nuclear Energy Institute Demonstration Pilots for Prioritizing and Scheduling Implementation". In addition, NEI provided its summary and observations of the demonstration pilots in the "Nuclear Energy Institute, Report on Prioritization and Scheduling Pilot." The latest version of the NEI guidance was submitted to the NRC by letter dated November 14, 2014.
Based on insights and feedback obtained from the public and with experience gained during tabletop exercises and demonstration pilots, the staff presented four options to the Commission in SECY-15-0050, "Cumulative Effects of Regulation Process Enhancements and Risk Prioritization Initiative: Response to Commission Direction and Recommendations" dated April 1, 2015. In the SRM-SECY-15-0050 issued on August 25, 2015, the Commission did not approve separate RPI activities, but supported the consideration of risk insights in regulatory decision-making through existing agency processes. Additional information is available.
FY 2015 Status
In March 2015, the staff briefed ACRS with respect to a draft version of the Commission paper in which the staff presented options of RPI as a tool to reduce cumulative effects of regulation (CER). In its letter on this topic, ACRS agreed with the staff's recommendations and recommended that the staff should explicitly include risk information as an input to decisions and priorities for proposed regulatory actions regardless of the Commission's decisions about specific options or approaches in the SECY paper.
On April 1, 2015, the staff submitted SECY-15-0050, "Cumulative Effects of Regulation Process Enhancements and Risk Prioritization Initiative: Response to Commission Direction and Recommendations." This paper responds to the Commission's direction in SRM-COMSECY-14-0014, "Cumulative Effects of Regulation and Risk Prioritization Initiative: Update on Recent Activities and Recommendations for Path Forward," dated July 18, 2014. This paper provided the Commission with four options of using RPI as a tool to reduce CER for operating reactor licensees.
The first option would have maintained the status quo. Option 2 would have augmented existing regulatory processes allowing licensees to request exemptions and changes to implementation schedules for existing regulatory commitments. This option would have allowed licensees to use a risk-informed prioritization methodology as a basis for such request. Option 3 would have allowed licensees to submit a risk-informed, plant-specific implementation plan when the NRC adopts a new rule. Option 4 would have established a voluntary process that enables licensees to make plant-specific, risk-informed changes to implementation schedules for certain regulatory issues without requesting prior NRC approval.
On May 19, 2015, the staff, along with an external panel, briefed the Commission on issues related to CER and RPI. The discussions included the staff's identified lessons learned, possible approaches for implementing the RPI, as well as licensee experiences with RPI pilot projects. In the SRM-SECY-15-0050 issued on August 25, 2015, the Commission did not approve the RPI options. However, the Commission stated that it supports consideration of risk insights in regulatory decision-making through existing agency processes. The staff is exploring the development of additional guidance to enhance licensees' ability to use risk information in existing agency processes such as Title 10 of the Code of Federal Regulations (10 CFR) 50.12, "Specific Exemptions."
"Infrastructure Development in Support of Risk-Informed Regulations" – Per Commission direction in SRM-SECY-15-0050, staff work on RPI is discontinued.
Risk Informing Oversight of Emergency Preparedness (EP) and Response Plans
In coordination with the Federal Emergency Management Agency (FEMA), the staff initiated a study of performance based evaluation techniques that could be used for offsite response organization Radiological Emergency Response Plans (RERP). This effort also intends to identify how RERP program elements could be integrated with nation-wide FEMA preparedness initiatives.
State and local emergency response programs have significantly matured since the EP regulations of 1980 were implemented. FEMA has initiated several nation-wide preparedness efforts and the level of capability has greatly advanced. The effectiveness and efficiency of EP oversight may be improved by further integrating NRC radiological emergency response programs with the broader FEMA preparedness initiatives.
The staff retained a knowledgeable consultant to review FEMA evaluation techniques and NRC regulations that apply to RERP. The consultant proposed elements of evaluation that could be performance based and examined FEMA preparedness programs that may duplicate or parallel NRC EP requirements and proposed methods for integration.
The study and evaluation of a performance based regimen for offsite response organizations has been completed with the conclusion that a performance based system is feasible and could enhance the effectiveness and efficiency of EP oversight. The results are documented in NUREG/CR-7195, "Risk-Informed and Performance-Based Oversight of Radiological Emergency Response Plans." This project also informed the development of SECY-14-0038, "Performance-Based Framework for Nuclear Power Plant Emergency Preparedness Oversight," that presented the results of a staff study on the potential for a performance based EP framework. In SRM-14-038, the Commission approved the staff's recommendation to continue under the current regulatory framework while remaining vigilant to the possibility of moving to a performance-based framework in the future.
FY 2015 Status
This study is complete; no further action is planned at this time.
Risk-Informed Oversight Activities. The purpose of this activity was to conduct a study of performance based evaluation techniques that could be used for offsite response organization Radiological Emergency Response Plans (RERP). The study and evaluation of a performance based regimen for offsite response organizations has been completed with the conclusion that a performance based system is feasible and could enhance the effectiveness and efficiency of EP oversight.
Emergency Core-Cooling System (ECCS) Requirements: Redefinition of Loss-of-Coolant Accidents (LOCA)
The staff prepared a proposed rule containing ECCS evaluation requirements that could be used as an alternative to the current requirements in 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light-Water Nuclear Power Reactors." That proposed rulemaking is designed to redefine the large-break LOCA (LBLOCA) requirements to provide a risk-informed alternative maximum break size. In October 2006, the staff produced a draft final rule and briefed the Advisory Committee on Reactor Safeguards (ACRS).
In response, ACRS recommended that the Commission should not issue the rule in its present form. As a result, the staff prepared SECY-07-0082, "Rulemaking To Make Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements: 10 CFR 50.46a, 'Alternative Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors,'" dated May 16, 2007, to provide a plan (including resource and schedule estimates) for responding to the ACRS recommendation and related comments.
On April 1, 2008, the Executive Director for Operations provided the staff's schedule for completing the final rule to the Commission. Following Commission approval, the NRC published a supplemental proposed rule, "Performance-Based Emergency Core Cooling System Acceptance Criteria" (74 FR 40765, August 13, 2009), for public comment. The public comment period ended in January 2010.
After reviewing public comments and making changes to address these comments (and ACRS comments), the staff submitted a final rulemaking package to the Commission for approval on December 10, 2010, in SECY-10-0161, "Final Rule: Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements (10 CFR 50.46a) (RIN 3150-AH29)." On April 20, 2012, the staff requested withdrawal of the 10 CFR 50.46a final rule from Commission consideration so that the staff could review the rule and ensure its compatibility with the ongoing regulatory framework activities under Recommendation 1 of the Fukushima Near-Term Task Force (NTTF) report. The Commission approved the staff's request in SRM-SECY-10-0161, dated April 26, 2012. The staff does not plan to publish a notice in the Federal Register withdrawing the 10 CFR 50.46a final rule. The staff intends to resubmit the draft final rule for Commission consideration after receiving Commission direction in conjunction with NTTF Recommendation 1. In response to the Staff Requirements Memorandum (SRM) on SECY-13-0132, "Nuclear Regulatory Commission Staff Recommendation for the Disposition of Recommendation 1 of the Near-Term Task Force Report,", the staff requested an extension to this and other initiatives, across other NRC program areas, to evaluate the Risk Management Regulatory Framework (RMRF) approach recommended in NUREG-2150 as well as alternative approaches for a achieving a risk-informed regulatory framework. The staff will submit a Commission paper on RMRF by December 18, 2015 and it will provide an update on the staff's path forward on this activity. Additional information is available.
FY 2015 Status
No action in fiscal year (FY) 2015, as this item is on hold.
Rulemaking Applications Using Risk Insights. The purpose of this activity is to incorporate risk insights into the Code of Federal Regulations.
Emergency Core Cooling System (ECCS) Requirements: Loss of Coolant Accident and Loss of Offsite Power (ECCS-LOCA/LOOP)
The proposed rule would amend the Commission's regulations to eliminate, based upon appropriate risk considerations, the assumption of a coincident LOOP for postulated LBLOCAs (low frequency) in General Design Criterion (GDC) 35. The proposed rule would provide a voluntary alternative to existing requirements in situations where specified acceptance criteria are satisfied, and also would address a petition for rulemaking submitted by Bob Christie (Performance Technology) (PRM-50-77). The staff's approach was to develop the technical basis for a LOOP-LOCA rule by reviewing the Boiling Water Reactor Owners Group (BWROG) topical report (TR), NEDO-33148, "Separation of Loss of Offsite Power from Large Break LOCA," dated April 27, 2004. In the March 31, 2003, SRM directing the staff to go forward with a risk-informed rule decoupling LOOP from LOCA, the Commission stated that the rule should consider the risk impacts of a "delayed LOOP and possible double-sequencing of safety functions." During the review of the BWROG TR, the potential safety impact of a LOCA followed by a delayed LOOP became a major issue. Existing nuclear plants are designed to handle only the simultaneous LOCA and LOOP. The capability of many plants to successfully mitigate upsets causing a delayed LOOP has not been determined. In December 2007, in COMSECY-07-0041, "Status of Staff Activities on Proposed Rule for Risk-Informed Decoupling of Assumed Loss-of-Offsite Power From Loss-of-Coolant Accident Analyses," the staff indicated its plans to reassess the need for a LOOP-LOCA rule after making final decisions on the BWROG TR and on the 10 CFR 50.46a risk-informed ECCS rule. In an SRM related to SECY-07-0082 dated August 10, 2007, the Commission agreed with the staff's recommendation that completing the rulemaking should be assigned a medium priority. Prior to competing its review of the TR, the staff concluded that the approach could not be approved without evaluating an individual plant's capability to successfully cope with a delayed LOOP. By letter dated June 12, 2008, the BWROG withdrew the TR from further NRC review after concluding that continued development of the report was no longer cost effective, and if ultimately approved in the form desired by NRC staff, adoption by licensees would most likely be prohibitively expensive. In September 2009, SECY-09-0140, "Rulemaking Related to Decoupling an Assumed Loss of Offsite Power from a Loss of Coolant Accident, 10 CFR part 50, Appendix A, General Design Criterion 35," provided options for completing the rulemaking and recommended the option to discontinue the rulemaking effort. The staff's recommendation was based on the lack of a fully developed regulatory basis and expenditures of staff time to develop one would not be expected to result in a quantifiable safety improvement. In the SRM related to SECY-09-0140 dated July 12, 2010, the Commission directed the staff to defer the decision on the rulemaking effort until the 10 CFR 50.46a rule is implemented. In response to the SRM on SECY-13-0132, "Nuclear Regulatory Commission Staff Recommendation for the Disposition of Recommendation 1 of the Near-Term Task Force Report", the staff requested extension to this and other initiatives, across all NRC program areas, to evaluate the Risk Management Regulatory Framework (RMRF) approach recommended in NUREG-2150 as well as alternative approaches for a achieving a risk-informed regulatory framework. Additional information is available.
FY 2015 Status
No action in FY 2015, as this item is on hold.
Rulemaking applications using risk insights. The purpose of this activity is to incorporate risk insights into the Code of Federal Regulations.
Develop Risk-Informed Improvements to Standard Technical Specifications (STS)
The staff continues to work on the risk-informed technical specifications (RITS) initiatives to add a risk-informed component to the STS. The following summaries highlight these activities:
Initiative 1, "Modified End States," would allow licensees to repair equipment during hot shutdown rather than cold shutdown. The Topical Reports (TRs) supporting this initiative for boiling water reactor (BWR), Combustion Engineering (CE), Babcock & Wilcox (B&W), and Westinghouse Electric Company (Westinghouse) plants have been approved, and revisions to the BWR, CE, B&W, and Westinghouse STS are available at ML093570241 and ML103360003).
Initiative 4b, "Risk-Informed Completion Times," modifies technical specification completion times to reflect a configuration risk-management approach that is more consistent with the approach described in the Maintenance Rule, as specified in 10 CFR 50.65(a)(4). As reported previously in SECY-07-0191, "Implementation and Update of the Risk-Informed and Performance-Based Plan," dated October 31, 2007, the staff issued the license amendment for the first pilot plant, South Texas Project (STP), in July 2007.
In July 2010, Southern Nuclear Company (SNC) submitted a letter of intent for Vogtle Electric Generating Plant (VEGP) (Units 1 and 2) to implement RITS Initiative 4b. The NRC granted an associated fee waiver request and received a pilot application in September 2012. The NRC staff is nearing completion of its review of the application, and is actively working to resolve the remaining technical issues. The associated Technical Specification Task Force guidance (TSTF-505) to revise the STS was published in March 2012. Four additional applications to implement TSTF-505 have been received and are currently being reviewed by the technical staff. The four additional applications were received on November 25, 2013; December 5, 2014; December 23, 2014; and July 31, 2015. The four additional applications are not classified as "pilot applications."
Initiative 6, "Add Actions to Preclude Entry into LCO 3.0.3," modifies technical specification action statements for conditions that result in a loss of safety function related to a system or component included within the scope of the plant technical specifications. The staff approved the industry's TR for CE nuclear power plants (Revision 2 to WCAP-16125-NP-A, "Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown") in August 2010. The associated Technical Specification Task Force (TSTF) guidance (Revision 5 of TSTF-426) to revise the CE STS was submitted for NRC review by letter dated November 2011. Based on the approved CE TR, the industry has also submitted requests to revise the B&W STS (Revision 0 of TSTF-538) and the STS for BWRs (Revision 0 of TSTF-540) in March 2012 and May 2012, respectively. However, these TSTFs were withdrawn per letters dated January 6 and October 6, 2014 after the NRC requested additional information and the participating licensees decided not to pursue these initiatives. Additional information is available.
FY 2015 StatusThe NRC staff continued review of STS initiatives as they were received.
Risk-Informed Licensing Reviews. Consistent with the Commission's policy statement on technical specifications and the use of probabilistic risk assessment (PRA), the NRC and the industry continue to develop more fundamental risk-informed improvements to the current system of technical specifications. Initiatives for fundamental improvements to the STS are being developed by the industry and discussed with the NRC staff in public meetings.
Implement 10 CFR 50.69: Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
In 1998, the Commission decided to consider issuing new regulations that would provide an alternative risk-informed approach for special treatment requirements in the current regulations for power reactors. The NRC published the final rule (10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors") in the Federal Register on November 22, 2004 (69 FR 68008). The NRC staff issued Regulatory Guide (RG) 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance," in May 2006.
By letter dated December 6, 2010, the Southern Nuclear Company (SNC) informed the NRC of its intent to submit a license amendment request for implementation of 10 CFR 50.69 for Vogtle Electric Generating Plant (VEGP) Units 1 and 2, and requested pilot plant status and a waiver of review fees. By letter dated June 17, 2011, the staff informed SNC that the NRC granted the fee waiver request for the proposed licensing action in accordance with 10 CFR 170.11(b). SNC submitted a pilot plant application to implement 10 CFR 50.69 on August 31, 2012. By letter dated December 17, 2014, the NRC staff issued a License amendment to SNC revising the licensing basis for the VEGP by adding license conditions that allow for the voluntary implementation of 10 CFR 50.69. Lessons learned from the application review will be used to revise the associated industry guidance and RG 1.201.
In addition, the NRC staff issued draft Inspection Procedure 37060, "10 CFR 50.69 Risk Informed Categorization and Treatment of Structures, Systems, and Components Inspection," on February 16, 2011. The Nuclear Energy Institute (NEI) and one licensee provided comments on the procedure. The NRC staff addressed the comments and issued the revised inspection procedure in 2011. The NRC will focus its inspection efforts on the most risk-significant aspects related to implementation of 10 CFR 50.69 (i.e., proper categorization of SSCs and treatment of Risk-Informed Safety Class (RISC)-1 and RISC-2 SSCs).
As part of the Regulatory Guide Periodic Review, the NRC reviewed RG 1.201 to determine whether changes were necessary to incorporate lessons learned from the VEGP pilot application. The review concluded that the RG could be updated, but identified no safety concerns if the guide is not updated. The NRC staff did not recommend an update because no additional LARs have been submitted. Additional information is available.
FY 2015 Status
Completed the pilot application for the Vogtle Electric Generating Plant (VEGP) in December 2014.
Risk-Informed Licensing Reviews. The purpose of this activity was to review a pilot application of 10 CFR 50.69, grant the amendment as appropriate, and apply any lessons to future reviews and update RG 1.201, if necessary.