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:

• Objective
• Basis
• Goals
• List of Risk-Informed and Performance-Based Activities

Objective

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.

Basis

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:
1. measurable parameters to monitor performance
2. objective criteria to assess performance
3. flexibility to allow licensees to determine how to meet the performance criteria
4. 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.

Goals

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:

• State-of-the-Art Reactor Consequence Analyses
• Probabilistic Methodologies for Component Integrity Assessment
• Implementing Lessons Learned from Fukushima
• Accident Sequence Precursor (ASP) Program
• Design Compliance Enforcement Discretion (DCED): a Risk-Informed Approach for Addressing Low Risk, Low Safety Significance Design Compliance Issues
• Probabilistic Flood Hazard Assessment (PFHA)
• Methods, Tools and Guidance for Including Digital Systems in Nuclear Power Plant PRAs
• Risk Assessment of Operation Events (RASP Handbook)
• 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
• Consequential Steam Generator Tube Rupture Probability and Consequence Assessment
• National Fire Protection Association (NFPA) Standard 805
• Assess Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance, Generic Safety Issue (GSI)-191
• 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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State-of-the-Art Reactor Consequence Analyses

Summary Description

The state-of-the-art reactor consequence analyses (SOARCA) project was initiated to evolve our understanding of the consequences of important severe accident scenarios at selected U.S. nuclear power plants including Peach Bottom, a BWR in Pennsylvania; Surry, a PWR with a large dry containment in Virginia; and Sequoyah, a PWR with an ice condenser containment in Tennessee. The project has focused on detailed modeling of accident progression using MELCOR and offsite consequences using MACCS (MELCOR Accident Consequence Code System). MELCOR models the severe accident processes within the plant to the point of release of fission products to the environment. MACCS models the atmospheric transport and deposition of radionuclides released to the environment as well as emergency response and long-term protective actions, exposure pathways, dosimetry, and health effects for the affected population. Staff conducted uncertainty analyses (UA) of a subset of the scenarios to better understand the range of potential outcomes for these accidents and what drives key phenomena. Each UA included hundreds of simulations to account for uncertainty in MELCOR and MACCS input parameters and the results help corroborate the project's overall conclusions.

The staff completed deterministic and sensitivity analyses of Peach Bottom and Surry which are documented in NUREG-1935, NUREG/CR-7110, and NUREG/BR-0359. A UA was conducted for the Peach Bottom unmitigated long-term station blackout (LTSBO) scenario and was documented in NUREG/CR-7155. Subsequently the Commission approved in SRM-SECY-2012-0092 limited additional analyses to further address the SOARCA objectives and to also support agency projects such as evaluation of Fukushima Near Term Task Force recommendations and the Full-Scope Site Level 3 PRA project.

FY 2017

In FY 2017 staff completed an updated uncertainty analysis (UA) of the Sequoyah unmitigated short-term station blackout (STSBO) scenario and briefed the NRC's ACRS. This work will be published in a NUREG/CR report. In FY 2018 the staff is continuing to work on an updated UA of the Surry unmitigated STSBO leveraging insights from the Sequoyah UA.

Risk-Informed Basis

By its nature SOARCA focuses on the consequences of accidents rather than on their likelihood or on the many redundant safety systems, components, procedures, training, strategies, or the recently added backup mitigation equipment required following the Fukushima Dai-ichi nuclear power plant accident in Japan. Plant safety features and added mitigation capability drive down the likelihood of a severe accident but not necessarily the consequences. The study of the unmitigated consequences of a severe accident does not dismiss or under-value those safety features, rather it sheds light on their importance by providing insights into the possible consequences they are intended to prevent. SOARCA project's results, insights, computer code models, and modeling best practices have supported NRC rulemaking, licensing, and oversight efforts. SOARCA supported SECY-15-0137 and SECY-16-0041 which closed NRC's evaluation of post-Fukushima recommendations related to containment vents and hydrogen control and mitigation.

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Probabilistic Methodologies for Component Integrity Assessment

Summary Description

The U.S. Nuclear Regulatory Commission (NRC) has considered insights drawn from probabilistic methodologies for component integrity assessment as part of its regulatory decision-making for several decades. The use of probabilistic methods moves the agency further towards risk-informed decision-making, which is a stated policy goal of the NRC. Furthermore, the NRC needs methodologies and procedures that enable it to perform an educated, thoughtful review of probabilistic methods proposed by the industry. The NRC currently has several active projects related to probabilistic methodologies for component integrity assessment: (1) maintenance and improved verification and validation of the Fracture Analysis of Vessels – Oak Ridge (FAVOR) code, (2) development and release of the Extremely Low Probability of Rupture (xLPR) code, and (3) development of a probabilistic fracture mechanics (PFM) Regulatory Guide.

The NRC and the U.S. nuclear industry have used probabilistic methods to inform their evaluation of postulated pressurized thermal shock (PTS) of reactor pressure vessels (RPVs) since the 1980s. In the original PTS rule (10 CFR 50.61) probabilistic evaluations provided complementary information to deterministic evaluations, and the reference temperature (RTPTS) screening criteria in 10 CFR 50.61 relate to a vessel failure frequency of ≈5×10-6 events / reactor operating year. Several PFM codes were used in the 1980s, including VISA (Vessel Integrity Simulation Analysis) and OCA-P (Over Cooling Accident - Pressurized). In the mid-1990s these codes were combined to generate the FAVOR code, which later provided computational support for the technical basis of the alternate PTS rule, 10 CFR 50.61a. FAVOR has since found other applications (e.g., risk-informed pressure-temperature limits, evaluation of nil-ductility transition [RTNDT] uncertainties, and evaluation of quasi-laminar flaws), although these other applications have not garnered generic regulatory acceptance.

In a separate activity, NRC and the Electric Power Research Institute (EPRI) have collaboratively developed the xLPR Version 2.0 PFM code to assess the effects of active degradation mechanisms on nuclear power plant piping systems approved for leak-before-break (LBB). Specifically, beginning around the year 2000, primary water stress corrosion cracking (PWSCC) was discovered in systems that had previously been approved for LBB based on the assumed absence of active degradation mechanisms, in accordance with the General Design Criteria in 10 CFR 50. As a result of the discovery of the PWSCC active degradation mechanism, an extremely low probability of rupture could no longer be demonstrated by the deterministic methods outlined in NUREG-0800, but would instead need to be addressed probabilistically, for instance by using a PFM code such as xLPR. Technical development of the full production version of the code is now complete. Various activities were undertaken during the development phase to build confidence into the code, including a broad team of experts from diverse backgrounds, a rigorous quality assurance program, comprehensive verification and validation, and extensive documentation.

With the release of FAVOR v16.1 and xLPR v2.0, PFM use by the U.S. nuclear industry is expected to increase, as PFM may be used to develop a technical basis for relief requests, license amendments, and topical reports. Uncertainty is addressed differently in PFM when compared to deterministic fracture mechanics. In PFM, a single deterministic (usually conservative) analysis is replaced by many deterministic analyses that use randomly sampled inputs. Statistical analyses are then performed on the collection of outputs obtained to determine the probability of an event of interest. Unfortunately, it is difficult for NRC staff to reproduce or verify PFM calculations submitted by licensees, thus resulting in complex regulatory reviews. In particular, NRC staff has often perceived PFM codes as ‘black boxes' with insufficient vetting of the models and the uncertainty framework. This has resulted in low confidence in the results of PFM analyses. As a result, the NRC has begun developing guidance for performing and documenting PFM analyses for regulatory applications. Specifically, NRC's Office of Nuclear Regulatory Research has been tasked with developing a PFM Regulatory Guide (RG). The process of developing the RG involves publication of a Technical Letter Report, a technical basis NUREG, and the Draft RG itself. The Technical Letter Report will be released publicly before the end of 2017.

FY 2017

A recent release of FAVOR, Version 16.1, includes updated fracture driving force solutions for surface-breaking flaws and the ability to analyze both heat-up and cool-down transients in the shell coarse region of both pressurized water reactor (PWR) and boiling water reactor designs. Planned efforts are underway to assess potential safety issues related to shallow subsurface flaws, including warm pre-stress effects, cladding residual stress modeling, and an assessment of risk-optimized pressure-temperature corridors for RPV heat-up and cooldown. NRC and EPRI are currently pursuing coordinated efforts to apply xLPR to conduct probabilistic LBB studies for the U.S. fleet of PWRs. A Technical Letter Report on important aspects to be considered for PFM has been produced and lays the foundation for the upcoming development of the PFM RG and its technical basis.

Risk-Informed Basis

PFM is typically used to determine the likelihood of a component failure, or the likelihood of a precursor to component failure. As such, PFM can answer one of the two fundamental questions in risk assessment: what is the initiating event frequency or likelihood of occurrence? The other question that PFM does not address is: what are the consequences of such an event occurring? In addition to the likelihood of an event, PFM can also be used to determine confidence bounds on the probability of an even of interest.

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Implementing Lessons Learned from Fukushima

Summary Description

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

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.

FY 2016

The NRC staff made significant progress in developing the infrastructure to support its review of licensees' submittals of the results of their seismic probabilistic risk assessments (PRAs). The first such submittal is expected to be received in the first quarter of calendar year 2017.

FY 2017

The NRC completed the development of the infrastructure to support the review of licensees' seismic PRA submittals. The NRC received three seismic PRA submittals, on a staggered schedule over the course of the year, and began implementing the review process. The first staff assessment of a seismic PRA submittal is expected to be issued by the NRC in the first quarter of calendar year 2018. The NRC expects to receive five more seismic PRA submittals in 2018, and the remainder of the seismic PRA submittals in 2019, all on a staggered schedule. More information on this risk-informed initiative can be found on the NRC's Seismic Reevaluations Web page.

Risk-Informed Basis

Seismic PRAs will be submitted to and reviewed by the NRC staff for about 20 sites. The risk insights from the seismic PRAs will be used by the staff to evaluate the impact of the site-specific reevaluated seismic hazard and determine whether further regulatory actions are warranted.

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Accident Sequence Precursor (ASP) Program

Summary Description

In 1979, the U.S. Nuclear Regulatory Commission (NRC) established the Accident Sequence Precursor (ASP) Program in response to the Risk Assessment Review Group report issued in September 1978 (NUREG/CR-0400, "Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Commission"). The evaluations performed for events that occurred between 1969 and 1979 were the first efforts in this type of analysis. The ASP Program systematically evaluates U.S. nuclear power plant operating experience to identify, document, and rank operational events by calculating a conditional core damage probability (CCDP) or an increase in core damage probability (ΔCDP).

The ASP Program identifies potential precursors by reviewing operational events from licensee event reports on a plant unit basis. An operational event can be one of two types: (1) the occurrence of an initiating event, such as a reactor trip or a loss of offsite power, with or without any subsequent equipment unavailability or degradation; or (2) a degraded plant condition characterized by the unavailability or degradation of equipment without the occurrence of an initiating event.

For the first type of event, the staff calculates a CCDP. This metric represents a conditional probability that a core damage state is reached given the occurrence of the observed initiating event (and any subsequent equipment failures or degradations). For the second type of event, the staff calculates a ΔCDP. This metric represents the increase in core damage probability for the time period during which a component or multiple components were deemed unavailable or degraded.

Starting in 2006, in an effort to minimize overlap and improve efficiency, Significance Determination Process (SDP) results have been used in lieu of independent ASP analyses to the extent practical and consistent with the overall objectives of both programs. More information regarding the details of this change is documented in NRC Regulatory Issue Summary 2006-24.

FY 2015

The ASP Program independently identified five precursor events in Fiscal Year (FY) 2015. In addition, four precursor events were analyzed by the SDP and accepted into the ASP Program (as described in NRC Regulatory Issue Summary 2006-24). See SECY-15-0124, "Status of the Accident Sequence Precursor Program and the Standardized Plant Analysis Risk Models," for more information on the status of the ASP Program for FY 2015.

FY 2016

In FY 2016, the ASP Program implemented a variety of administrative changes. In accordance with Project AIM, and by direction of the Commission, the status of the ASP Program will no longer be reported in the annual SECY paper "Status of the Accident Sequence Precursor Program and the Standardized Plant Analysis Risk Models." Instead, an annual summary of the ASP Program will be provided as a publicly available document. In addition, the ASP Program transitioned from a FY reporting cycle to a calendar year (CY) reporting cycle. Operational events will be organized based on the CY in which the licensee event report is submitted to the NRC. As part of this transition, the FY 2015 annual report was combined with the CY 2016 report. Annual summary reports will be made available to the public in the following CY (e.g., the CY 2016 annual report was made available to the public in CY 2017).

The NRC's Risk-Informed Steering Committee initiated an internal evaluation of the ASP Program in July 2016, performed by staff within the Office of Nuclear Reactor Regulation. A public meeting was held on October 13, 2016, to solicit feedback from external stakeholders and members of the public.

FY 2017

In FY 2017, the ASP Program published the "U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program 2016 Annual Report," which summarizes the results of ASP analyses for events reported between October 2014 and December 2016. Twenty-three events were determined to be precursors. Of these 23 precursors, 15 precursors utilized SDP results in accordance with RIS 2006-24 and the remaining 8 precursors were identified via independent ASP analyses. Three of the events identified by ASP analyses had a CCDP or ΔCDP greater than or equal to 1x10^-5.

The NRC continues its internal evaluation of the ASP Program with a focus on identifying resource efficiencies through process changes, increasing the use of ASP results in other NRC processes, and ensuring timeliness of ASP analyses to support internal and external stakeholder needs. Recommended changes to the ASP Program will likely be communicated in early FY 2018.

Risk-Informed Basis

The ASP Program analyzes potential precursors by calculating the probability of an event leading to a core damage state. The analyses of operational events are conducted using the NRC's Standardized Plant Analysis Risk (SPAR) models and the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software.

The ASP Program is one of three agency programs that assess the risk significance of issues and events. The other two programs are the Reactor Oversight Process (ROP) SDP and the event response evaluation process, as defined in Management Directive (MD) 8.3, "NRC Incident Investigation Program." In contrast to the other two programs, a comprehensive and integrated risk analysis under the ASP Program includes all anomalies observed at the time of the event or discovered after the event. These anomalies may include unavailable and degraded plant structures, systems, and components (SSCs); human errors; and/or an initiating event (e.g., reactor trip). An unavailable or degraded SSC does not have to be a performance deficiency (PD) or an analyzed condition in the plant design basis, as required in the SDP. The ASP Program analyzes concurrent, multiple PDs or degraded conditions together, unlike the SDP that analyzes PDs individually.

The ASP Program results are used to support programmatic and regulatory decisions. Specifically, RES provides recommendations for any programmatic or regulatory reviews based on results of adverse ASP trends and results of precursor analyses identifying a potentially generic issue. The ASP program provides unique and independent inputs to the Report to Congress on Abnormal Occurrences (NUREG-0090), Congressional Budget Justification (NUREG-1100), Performance and Accountability Report (NUREG-1542), Strategic Plan (NUREG-1614), and the NRC's Agency Action Review Meeting (AARM).

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Design Compliance Enforcement Discretion (DCED): a Risk-Informed Approach for Addressing Low Risk, Low Safety Significance Design Compliance Issues

Summary Description

The agency is developing a risk-informed approach to resolve licensee design issues that render a technical specification structure, system or component inoperable and 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 design non-conformance 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 and safety significance of the non-compliance. If the issue is determined to be of low risk and low safety significance, 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

In September 2015, a working group with members from NRR, the Regions, OGC, and OE was formed, and began 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 also looked at the process for implementing this new approach. One implementation method that was considered was modifying the Notice of Enforcement Discretion (NOED) process to provide a process for addressing for low risk, low safety significance design compliance issues in a risk-informed manner.

FY 2016

Three public meetings were held to discuss this initiative. The meetings were held at NRC Headquarters on February 3, 2016, April 11, 2016, and May 23, 2016. The Commission was also briefed on the initiative during the Operating Reactor Business Line briefing on July 7, 2016. A draft outline of the proposed process was developed and circulated within NRR, the Regional Offices, OE and OGC for comment.

FY 2017

After modification of the draft outline based on internal feedback, the outline was made publicly available for feedback from external stakeholders. Based on feedback received from internal and external stakeholders from the review of the draft outline for the proposed DCED process, a draft DCED procedure was developed and circulated internally for review and a draft Commission Notation Vote paper was prepared. However, the DCED Commission Paper due date was extended to October 2018 for the following reasons:

1. So the staff can examine new guidance documents that are under development (e.g., backfit guidance resulting in part from Commission direction in SRM-COMSECY-16-0020 and operability guidance under development by NEI) and evaluate their potential for reducing the number of low risk and low safety significance operability issues created by non-compliances with design requirements.
2. If the staff concludes that the new guidance documents are unlikely to significantly reduce the number of DCED candidate issues, the staff will explore additional options consistent with feedback received from both internal and external stakeholders. The options will seek to:
   1. better balance the public's hearing rights with risk-informing the agency's response to low risk, low safety significance operability issues,
   2. align on the extent to which technical specifications can and should be risk-informed, and
   3. engage more extensively with external stakeholders.

Risk-Informed Basis

The proposed process will utilize risk insights as one of the criteria to determine if a design issue is a candidate for the licensee to request enforcement discretion under the proposed DCED process.

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Probabilistic Flood Hazard Assessment (PFHA)

Summary Description

The PFHA research program is a wide-ranging effort to establish a sound technical basis for transitioning flood hazard assessment guidance and tools from deterministic to probabilistic approaches. The PFHA research is guided by a joint NRO-NRR 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 in 2014. RES has been implementing the research plan for approximately 3 years.

By supporting development of risk-informed licensing and oversight guidance and tools for assessing flooding hazards and consequences, this research addresses a significant gap in the probabilistic basis for external hazards since seismic and wind hazard assessments are currently conducted on a probabilistic basis. The PFHA research program is designed to support both new reactor licensing (e.g. design basis flood hazard assessments for new sites or facilities) and oversight of operating reactors (e.g. significance determination process analyses for evaluating inspection findings or event assessments involving flood hazards, flood protection, or flood mitigation at operating facilities).

FY 2015

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).

FY 2016

Thirteen research projects have been initiated via interagency agreements with the US Army Corps of Engineers, the US Geological Survey, the Bureau of Reclamation, Idaho National Laboratory (INL), and Pacific Northwest National Laboratory (PNNL). A fourteenth project is being conducted with the University of California at Davis via a cooperative research contract with USGS under authority of the Water Resources Research Act. A fifteenth research activity has been implemented as a commercial contract. Cooperative research efforts have been initiated with the Electric Power Research Institute (EPRI) under a Flooding Research Addendum to an existing NRC-EPRI MOU. A cooperative research agreement is under development with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN).

FY 2017

Progress has continued on the existing projects initiated via interagency agreements and cooperative research contracts with other agencies and commercial contracts, as reported last year. A number of technical reports have been completed. Two new projects have been initiated via interagency agreement with Oak Ridge National Laboratory. The 2nd annual program review on the progress of PFHA research projects was held on January 23-25, 2017 at NRC headquarters. Cooperative research efforts with the Electric Power Research Institute (EPRI) have continued under the Flooding Research Addendum to an existing NRC-EPRI MOU. Two technical exchanges with EPRI were held in FY 2017. The technical aspects of a cooperative research agreement with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN) were completed and the agreement is under review by IRSN and NRC management.

Risk-Informed Basis

This activity is risk-informed because it addresses several aspects of risk: (1) probability or frequency of occurrence for various flooding scenarios; (2) fragility of flood protection features; and (3) reliability of flood protection and mitigation procedures.

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Methods, Tools and Guidance for Including Digital Systems in Nuclear Power Plant PRAs

Summary Description

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

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 FY 2016.
• 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 FY 2016.

More background on this approach can be found in the transcripts from an ACRS subcommittee meeting held in November 2014.

FY 2016

In collaboration with the Korea Atomic Energy Research Institute, the staff completed the development of an approach for quantifying software reliability using a Bayesian Belief Network (BBN)-based model. A NUREG/CR report describing the BBN approach was submitted for publication in FY 2016. The PRA-based statistical testing method was applied to the INL Advanced Test Reactor Loop Operating Control System. A NUREG/CR report describing the statistical testing application was submitted for publication in FY 2016.

FY 2017

In May 2017, the NRC published NUREG/CR-7234, "Development of a Statistical Testing Approach for Quantifying Safety-Related Digital System on Demand Failure Probability." At this time, there are no plans for future work in this area as under Project AIM, support for work in this area was eliminated.

Risk-Informed Basis

This research program aims to develop methods to quantify safety related digital I&C system failure probabilities that enable the inclusion of digital I&C components into current NPP PRAs.

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Risk Assessment of Operation Events (RASP Handbook)

Summary Description

Provide methods and guidance for the risk-informed analysis of operational events and licensee performance issues including internal and external events during both full power and low-power/shutdown operations.

Risk-Informed analyses are performed in response to needs identified in: Management Directive 8.3, "Incident Investigation Program"; Reactor Oversight Process; the Significance Determination Process (SDP); and the Accident Sequence Precursor (ASP) program. State-of-the-practice methods and guidance support risk analysts and senior reactor analysts from various NRC offices (NRR, RES, NRO, and the Regions) that use risk analysis software (SAPHIRE) and plant-specific PRA model (SPAR models).

The Risk Assessment Standardization Project (RASP) handbook and associated internal web site provides guidance and a description of the methods the NRC staff uses to achieve consistent results in the performance of risk-informed studies 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 incorporates best practices gleaned from experience on accident precursor events performed in ASP reviews and other insights gained from SDP analyses.

FY 2015

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.

FY 2016

The staff prepared for the publication of a NUREG on the application of Common Cause Failure (CCF) Analysis in Event and Condition Assessment. The intent of this report is to provide acceptable methods that the staff will accept in the area of CCF when applied to identified component and system failures which typically occur as part of SDP and ASP evaluations.

FY 2017

The staff prepared for the publication of a NUREG on the basis for the treatment of potential common-cause failure in risk-informed analysis. The intent of this report is to provide acceptable methods that the staff will perform in the area of potential CCF when applied to identified component and system failures which typically occur as part of SDP and ASP evaluations.

The staff revised the RASP handbook volume on internal events that provides additional guidance on how to credit alternate mitigating strategies (e.g., FLEX) in risk assessments. These mitigating strategies employ plant responses which utilize portable equipment to restore or maintain various safety functions during beyond design basis conditions and the loss of permanently installed plant equipment.

Risk-Informed Basis

This activity helps to put a risk perspective on operational events and inspection findings. It is not always obvious how much actual risk is associated with identified violations or component/system failures. This activity attempts to take advantage of insights gained using PRA modeling as applied to operational events discovered during normal operations, which have the potential to contribute to nuclear plant risk. As such, it provides a different and independent perspective on nuclear plant performance than would be available simply by tracking compliance with plant technical specifications and operational directives.

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Maintenance and Development of the Systems Analysis Programs for Hands-on Analysis Integrated Reliability Evaluations (SAPHIRE) Code

Summary Description

The NRC has developed and maintains the SAPHIRE 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), Accident Sequence Precursor (ASP) program, risk-informing 10 CFR Part 50, vulnerability assessment, advanced reactor assessment, operational experience, generic issues, and regulatory backfit.

FY 2015

A summary of recent activities regarding the status of the SAPHIRE computer code can be found in SECY 15-0124, "Status of the Accident Sequence Precursor Program and the Standardized Plant Analysis Risk Models."

FY 2016

The SAPHIRE development team continues to maintain the code's performance and add new features in accordance with the SAPHIRE software quality assurance program. During FY 2016 two new SAPHIRE versions were released for use by NRC staff. Improvements include enhanced seismic hazard modeling capability and development of a new quantification approach with improved accuracy for models involving high failure probabilities.

FY 2017

The SAPHIRE development team released one new version of the SAPHIRE software during FY 2017. A number of improvements were made to the reporting capabilities and user options. In addition, the number of modeled accident sequences that SAPHIRE can store was increased from 2,000,000 to 4,500,000, which was necessary as the size and complexity of models continues to grow. The new version release coincided with a significant update to all the SPAR models. The SAPHIRE team performed extensive testing with the new SAPHIRE version to identify and resolve any issues prior to releasing the updated models. The SAPHIRE development team continues to maintain the code's performance and add new features in accordance with the SAPHIRE software quality assurance program.

Risk-Informed Basis

The SAPHIRE computer code is used to develop and run PRA models (e.g., SPAR models) for a variety of risk-informed regulatory applications.

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Standardized Plant Analysis Risk Models (SPAR)

Summary Description

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. Nuclear Power Plant (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

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."

FY 2016

The staff continued to develop new SPAR model capabilities and provide technical support for SPAR model users and risk-informed programs. The staff maintains and implements a quality assurance (QA) plan for the SPAR models to ensure that the models appropriately represent the as-built, as-operated nuclear plants to support the assessment of operational events within the staff's risk-informed regulatory activities. The SPAR QA Plan provides mechanisms for model benchmarking and reviews, validation and verification, and configuration control of the SPAR models. In addition, about half of the SPAR models are updated to reflect significant plant modifications or other plant or modeling changes.

The staff also continued developing the SPAR model for the AP1000 new reactor design, adding a low power shut down model and a level 2 PRA model for the AP1000 reactor design.

FY 2017

The staff continued to maintain all SPAR models, with the implementation of the QA plan to represent the as built-as operated nuclear plants; and continued to provide technical support for SPAR model users and risk-informed programs. During FY 2017, the staff updated all SPAR models to reflect the most recent plant reliability data. For new reactor designs, the staff continued to work on expanding the AP1000 SPAR model capabilities (e.g., shutdown and Level 2 model); and initiated work on plant specific SPAR models for Vogtle (AP1000).

Risk-Informed Basis

The SPAR models are used by NRC staff in support of risk-informed activities related to the inspection program, incident investigation program, license amendment reviews, performance indicator verification, accident sequence precursor program, generic safety issues, and special studies. These models also support and provide rigorous and peer reviewed evaluations of operating experience, thereby demonstrating the agency's ability to analyze operating experience independently of licensees' risk assessments and enhancing the technical credibility of the agency.

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Full-Scope Site Level 3 PRA

Summary Description

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

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.

FY 2016

A substantial revision was completed for the reactor, at-power, Level 1 PRAs for internal events and internal floods, and the associated reports are nearing completion. The reactor, at-power, Level 1 PRAs for internal fires and seismic events were significantly revised to incorporate more recent licensee-supplied information, and our currently undergoing internal technical review. The dry cask storage (DCS) PRA was completed for all PRA levels and all hazards, and reviewed internally. In response to review comments, the consequence analysis for the DCS PRA is now undergoing revision. An initial reactor, low power and shutdown, Level 1 PRA for internal events is nearing completion. An approach was developed for modeling integrated site risk and a pilot application of this approach was performed based on the results of the revised Level 1 PRAs for internal events for Vogtle, Units 1 and 2. A similar pilot application is being performed based on the results of the initial Level 2 PRAs for internal events for Vogtle, Units 1 and 2.

FY 2017

A substantial revision was completed for the reactor, at-power, Level 2 PRA for internal events and internal floods, and is currently undergoing final project management review. Work is nearing completion on a substantial revision to the reactor, at-power, Level 3 PRA for internal events and internal floods. The reactor, at-power, Level 1 PRAs for internal fires and seismic events have completed their internal technical reviews, and are currently undergoing project management review. Substantial revisions were completed for the reactor, at-power, Level 1 PRA for high winds and the qualitative screening analyses for other hazards, and both are currently undergoing final project management review. The DCS PRA for all PRA levels and all hazards was revised and is currently undergoing project management review. An initial reactor, low power and shutdown (LPSD), Level 1 PRA for internal events was completed and is currently in queue for project management review. Two-unit pilot applications of the integrated site risk approach were completed for the Level 2 PRA for internal events, the Level 1 PRA for seismic events, and the Level 1 PRA for LPSD (one unit in operation, and one unit in shutdown).

Risk-Informed Basis

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 PRA project are expected to benefit a variety of ongoing risk-informed regulatory initiatives.

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Data Collection for Human Reliability Analysis (HRA)

Summary Description

Consistent with the Commission's policy statements on the use of probabilistic risk assessment (PRA) and for achieving an appropriate PRA quality for NRC risk-informed regulatory decision-making, the NRC has ongoing activities to improve the quality of human reliability analysis (HRA). The adequacy of data available for HRA is a concern on the credibility and consistency of human error probability estimates. To address this need, NRC's Office of Nuclear Regulatory Research (RES) has developed the Scenario Authoring, Characterization, and Debriefing Application (SACADA) system to collect operator performance information in simulator exercises. RES has collaborated with nuclear power plants and research institutes to use the SACADA system to collect their simulator training, examination, and experiment data. In addition, RES reviews literature and operations experience, and plans to collaborate with nuclear power plants to collect the human performance information of actions performed outside of the main control room. This includes actions to implement FLEX strategies to support the data needs identified by the Office of Nuclear Reactor Regulation (NRR).

FY 2015

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.

FY 2016

The following two SACADA collaborations were established in FY 2016:

• The Taiwan Power Company (TPC): To support this agreement, RES, with support of TPC, developed a Chinese version of SACADA for TPC plants to use. RES, with the support of the South Texas Project Nuclear Operating Company and the Idaho National Laboratory, provided SACADA training to the TPC instructors. TPC is piloting the SACADA system.
• The Advanced Test Reactor (ATR) of the Department of Energy: The ATR has used the SACADA system and has made data accessible to the NRC since June 2016.

FY 2017

The following are tasks accomplished in FY 2017:

• Established an agreement with the Grand Gulf Nuclear Generating Station to use the NRC's SACADA system to collect the licensed operator performance information in simulator training and to share the information with the NRC for improving HRA techniques.
• Awarded two contracts to perform independent analysis of the SACADA data for HRA. The results will be presented at a NRC-hosted HRA data workshop on March 15 and 16, 2018 at the NRC headquarters.

The following are activities are either in process or performed:

• Establishing an agreement for the Vogtle Unit 3 and Unit 4 site to use the NRC's SACADA system for operator simulator training. After the operators are licensed, the performance data will be shared with the NRC to improve HRA techniques.
• Performing literature and operations experience review to inform human performance assessment of FLEX strategy implementation.
• Plan to host a SACADA data workshop in March 2018 to discuss SACADA data analysis results and improvements.
• In negotiation with Entergy to collaborate on expanding the SACADA scope to collect operator performance in simulator training, on the job training, written tests, and actual events.
• Continue outreach to NRC licensees on using SACADA for operator simulator training.

Risk-Informed Basis

Human reliability analysis results are used in the NRC's risk-informed regulatory activities such as the reactor oversight process. The collected data would improve the reliability of NRC's HRA methods. That, in turn, improves the reliability of the NRC's risk-informed decision-making.

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Human Reliability Analysis (HRA) Methods and Practices

Summary Description

The purpose of the HRA method effort is to improve the methods for regulatory applications. This enhancement involves improving 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 available to evaluate human performance. The ongoing activities include:

• Developing the Integrated Human Event Analysis System (IDHEAS) for risk analyses of all nuclear-related HRA applications (SRM-M061020)
• Developing IDHEAS application for event and condition analysis (IDHEAS-ECA)

Regulatory Guide (RG) 1.200 provides an acceptable approach for determining the technical adequacy of probabilistic risk assessment (PRA) results for risk-informed regulation. HRA is a key element in the PRA. Because various HRA methods often have different assumptions and approximations that could lead to significant variability in results affecting regulatory decisions, enhancing the consistency and quality of HRA could improve regulatory decision-making.

FY 2015

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.

FY 2016

The following are tasks accomplished in FY 2016:

• Published NUREG-2199, Vol.1, "An Integrated Human Event Analysis System (IDHEAS) for Nuclear Power Plant Internal Events At-Power Application".
• Completed the testing of IDHEAS for internal at-power applications.
• Published NUREG-2156, "The U.S. HRA Empirical Study – Assessment of HRA Method Performances against Operating Crew Performance on a U.S. Nuclear Power Plant Simulator".
• Published NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detector Systems in Nuclear Facilities (DELORES-VEWFIRE).

FY 2017

The following are tasks accomplished in FY 2017:

• Published NUREG-2170, "A Risk-informed Approach to Understanding Human Error in Radiation Therapy"

The following are activities are in process:

• Completing the IDHEAS framework for risk analyses of all nuclear-related HRA applications.
• Developing the IDHEAS application for event and condition analysis to support the NRC's inspection, licensing, and enforcement activities.
• Working with the Electric Power Research Institute to develop an approach to perform HRA related to main control room abandonment in fire events:
  • In publication: NUREG-1921, Supplement 1, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines: Qualitative Analysis for Main Control Room Abandonment Scenarios"
  • Completing development of NUREG-1921, Supplement 2, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines: HRA Quantification for Main Control Room Abandonment Scenarios"

Risk-Informed Basis

The purpose of the HRA method efforts is to improve the methods to be used for regulatory applications and the consistency among HRA practitioners in performing HRA. This will help improve HRA/PRA quality and provide a basis for risk-informed regulatory actions.

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Consequential Steam Generator Tube Rupture Probability and Consequence Assessment

Summary Description

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

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.

FY 2016

The "Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes – Draft Report for Comment (NUREG-2195)," was issued for public comments. Public comments were received and were addressed. A final NUREG is expected to be issued late in 2017.

FY 2017

The final NUREG-2195 is in the publication process and will be available by early calendar year 2018.

Risk-Informed Basis

This project provides a method to assess the conditional SGTR probability given SG tube challenge (temperature-induced) during severe accidents or as an initiating event (pressure-induced). This probability can be used to assess the potential plant risk.

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National Fire Protection Association (NFPA) Standard 805

Summary Description

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 9.5.1.2, "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 SRM-SECY-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 28 applications to date and expects another application by April of 2018.

FY 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. Additional FY 2015 information is available.

FY 2016

The NRC staff issued seven non-pilot NFPA 805 license amendments. Five license amendment requests (LARs) are currently under review. Additional FY 2016 information is available.

FY 2017

The NRC staff issued five non-pilot NFPA 805 license amendments. Two LARs are currently under review. Additional FY 2017 information is available.

Risk-Informed Basis

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.

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Assess Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance, Generic Safety Issue (GSI)-191

Summary Description

This generic issue concerns the possibility that following a Loss of Coolant Accident (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 sump strainer(s) 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-SECY-12-0093, dated December 14, 2012. Since the Commission's endorsement, 11 licensees (18 units) have proposed to implement a risk-informed approach to address GSI-191 concerns. In consideration of the additional time required to implement risk informed approaches and/or complete further testing, subject licensees have implemented mitigative measures to address the potential for debris blockage of the strainer or reactor core.

SRM-SECY-12-0093, Title 10 of the Code of Federal Regulations (CFR) Section 50.46c, addresses ECCS performance during a LOCA. SECY-12-0034, dated January 7, 2013, directed that a provision allowing NRC licensees, on a case-by-case basis, to use risk informed alternatives should be included as part of proposed revisions to 10 CFR 50.46c. The proposed rule containing this provision was published in the Federal Register on March 24, 2014 (79 FR 16106).

In accordance with SRM-COMSECY-13-006, dated May 9, 2013, draft guidance related to implementation of the GSI-191 risk informed alternative was developed in parallel with its review of the STP pilot submittal, and published it in the Federal Register for public comment on April 20, 2015 (75 FR 21658).

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.

FY 2016

Preparations were made to ensure that final regulatory guidance (RG 1.229, "Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident") could be issued concurrent with the revised 10 CFR 50.46c rule. Proposed 50.46c rule changes were still pending Commission approval at the end of FY16. Several pre-submittal public meetings were conducted in preparation for forthcoming GSI-191 risk-informed closure submittals.

FY 2017

The staff completed its review of the STP pilot and issued a safety evaluation and license amendment approving the risk informed closure of GSI-191 for STP. Currently, eight additional units are expected to request similar risk-informed closures.

Risk-Informed Basis

Site-specific closeout of GSI-191 according to the risk-informed approach involves the use of a systematic processes to evaluate the risk from debris in terms of core damage frequency (CDF) and large early release frequency (LERF). The systematic risk assessment would rely on, at minimum, a plant-specific at-power, internal events probabilistic risk assessment (PRA) and take into consideration all hazards, initiating events, and plant operating modes. The risk attributable to debris would be compared to the risk calculated assuming debris is not present yielding values for the change in CDF and LERF (∆CDF and ∆LERF, respectively).

Licensees pursuing risk-informed approaches to address GSI-191 concerns, will be submitting license amendment requests subject to RG 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision on Plant-Specific Changes to the Licensing Basis."

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Develop Risk-Informed Improvements to Standard Technical Specifications (STS)

Summary Description

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 ADAMS Accession Nos. 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. Five additional applications to implement TSTF-505 have been received and are currently being reviewed by the technical staff. The five additional applications were received on November 25, 2013; December 5, 2014; December 23, 2014; July 31, 2015 and February 25, 2016. The five 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, 2014, and October 6, 2014, after the NRC requested additional information and the participating licensees decided not to pursue these initiatives.

FY 2015

The NRC staff continued review of STS initiatives as they were received. Additional FY 2015 information is available.

FY 2016

The NRC staff performed reviews of STS initiatives based license amendment applications as they were received. Additional FY 2016 information is available.

FY 2017

In late 2016 TSTF-505 was suspended pending updates required to address technical issues not previously identified. These issues are still being addressed. Although work is nearing completion, TSTF-505 remains suspended.

Risk-Informed Basis

The activity uses risk insights and results to identify appropriate improvements to the current STS and to determine appropriate compensatory risk management actions associated with plant equipment that is deemed inoperable per STS. Decisions concerning changes to STS are reached in an integrated fashion, considering traditional engineering and risk information, and may be based on qualitative factors as well as quantitative analyses and information.

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Implement 10 CFR 50.69: Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

Summary Description

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).

FY 2015

Completed the pilot application for the Vogtle Electric Generating Plant (VEGP) in December 2014. Additional FY 2015 information is available.

FY 2016

Although no new submittals seeking to implement 10 CFR 50.69 were received by the NRC in FY 2016, the industry has expressed interest in its widespread implementation. The NRC staff met with industry representatives in August 2016 to discuss future 50.69 LAR submittals and their content. Additional FY 2016 information is available.

FY 2017

NRC has received seven submittals to implement 10 CFR 50.69 in FY 2017. Additionally NRC met with industry representatives in several public meetings to discuss 50.69 topics of interest, such as content of License Amendment Requests, and industry proposed deviations from RG 1.201 guidance for addressing seismic and fire risk (ADAMS Accession Numbers ML17027A251, ML17177A063, ML ML17265A020). Additional licensee submittals to implement 10 CFR 50.69 are anticipated in FY 2018. The NRC staff met with industry representatives in October 2017 to discuss industry proposed approach to implementing 10 CFR 50.69 for licensees that do not have seismic probabilistic risk assessment or a seismic margins analysis (ADAMS Accession Number ML17305A242). Public meetings will continue in FY 2018.

Risk-Informed Basis

10 CFR 50.69 and its implementation relies heavily on risk insight and metrics, such as importance measures, to categorize the safety significance of systems, structures, and components (SSCs). The rule revises requirements with respect to 'special treatment,' that is, those requirements that provide increased assurance (beyond normal industrial practices) that SSCs perform their design basis functions. This rule permits licensees (and applicants for licenses) to remove SSCs of low safety significance, as determined based on the risk metrics, from the scope of certain identified special treatment requirements and to revise requirements for SSCs of greater safety significance. The plant-specific Probabilistic Risk Analysis (PRA) model is utilized to generate the risk metrics used for the categorization resulting in a quantifiable method of determining the risk significance of the components on the safe operation of the nuclear plant.

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Page Last Reviewed/Updated Thursday, May 10, 2018

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