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 with expanding menus:

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:

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Risk-Informed Reviews of Instrumentation and Control (I&C) Systems and Components: Integrating Risk Insights into the Digital I&C Regulatory Framework

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The objective of this research was to provide support in developing the technical basis for integrating risk insights into the regulatory framework for DI&C systems and components by: (1) assessing the technical feasibility of risk-informed approaches and gaps associated with further integrating risk insights into regulatory reviews for DI&C systems, and (2) develop recommendations to enhance the use of risk insights within the existing risk-informed regulatory framework for DI&C systems. A report titled “Assessment of Technical Feasibility of Risk-Informed Approaches and Gaps Associated with Further Integrating Risk Insights into Regulatory Reviews for Digital I&C Systems and Components” was produced. The report can be found in ADAMS (Accession No.: ML20296A259).

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Use of Systems-Theoretic Accident Model and Processes (STAMP)-based Methods for Digital Nuclear Safety System Evaluation

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NRC staff participated in a series of interactive seminars, workshops, and discussion to understand the risk-informing potential of STAMP-based methods in the NRC’s regulatory processes and the limits within which these methods can be consistently applied. Two reports detailing the results of this effort are expected by December 2021

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Technical Assistance for Integration of Risk-Informed Performance Based Approach to Seismic Safety of Nuclear Facilities

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This activity is completed as described in “Feasibility Study on a Potential ?Consequence-Based Seismic Design Approach for Nuclear Facilities”, ML21113A066.

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Revisions to NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness for NPP

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The NRC issued two significant emergency preparedness-related license amendment requests using the December 2019 risk-informed guidance in NUREG-0654/FEMA-REP-1, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,” Revision 2.  Specifically, on August 26, 2021, the NRC issued an 11-unit license amendment requests (LAR) to Duke Energy Common to replace the site-specific emergency plans with a Duke Energy Common Emergency Plan with site-specific annexes.  On September 21, 2021, the NRC issued an six-unit LAR to Southern Nuclear Corporation to risk inform the ERO staffing composition and increase the staff augmentation response time of certain ERO positions.   On September 27, 2021, the NRC granted an LAR (ADAMS Accession No ML21217A021) to Vogtle Electric Generating Plant, Units 3 and 4, to change the ERO staffing composition and extend staff augmentation time from 75 to 90 minutes.

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Revision to NUREG/CR-7002, "Criteria for Development of Evacuation Time Estimate Studies"

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On February 9, 2021, the NRC issued Revision 1 of NUREG/CR-7002, "Criteria for Development of Evacuation Time Estimate Studies," which reflects the importance of various evacuation time estimate (ETE) model parameters based on the results of an applied research study of ETEs published in NUREG/CR-7269, "Enhancements to Evacuation Time Estimate Guidance." The guidance was enhanced to risk-inform the size of the evacuation models, the impact of a shadow evacuation, modeling adverse weather, the use of manual traffic control, and various other parameters of importance.  The format and criteria provided in the guidance will support consistent application of the ETE methodology and will facilitate consistent NRC review of ETE studies.

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Power Reactor Cyber Security Program Improvements

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In a letter dated June 30, 2021, the NRC staff completed its review of the NEI white paper and concluded that the methods in the white paper for identifying and protecting critical digital assets associated with security functions are consistent with NEI 08-09, "Cyber Security Plan for Nuclear Power Reactors, "Revision 6.

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Ensure Force-on-Force (FoF) Scenarios Are Realistic and Reasonable

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The NRC staff initiated a pilot program for a scoring system for the related to exercise scenarios to determine credibility and applicability across the industry. The staff plans to make adjustments as necessary in its development to normalize the scoring, given the site-specific capabilities of the licensees.

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Consequence-based Security for Advanced Reactors

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The objective of this rulemaking is to permit future applicants and licensees to demonstrate a safety case and technical basis to meet alternative physical security requirements commensurate with the potential lower associated risks of advanced reactor designs. In response to comments from public stakeholders, including those from the Nuclear Energy Institute, the NRC staff revised proposed preliminary rule language that established risk-informed, performance-based requirements that are alternatives to the prescriptive requirements in NRC’s security regulations for power reactors. This language was made publicly available on September 14, 2020 to facilitate further stakeholder engagement (FR 2020-19907; 85 FR 56548). That stakeholder engagement focused on three primary areas: 1) determining appropriate eligibility criteria and their use; 2) target set identification process; 3) process for completing a consequence analysis considering a design basis threat-initiated event. Several possible alternatives to prescriptive security requirements are considered in the revised preliminary proposed rule language, including alternatives for armed responders, physical barriers, and an onsite secondary alarm station.

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Revision of the Emergency Preparedness Significance Determination Process

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SECY-19-0067, “Recommendations for Enhancing the Reactor Oversight Process” included a recommendation to revise the EP Significance Determination Process (SDP) such that only those planning standard (PS) functions that have an impact on public health and safety would have performance deficiencies assessed to have greater than green (GTG) safety significance. With the retraction of SECY-19-0067 in FY2021, NRC staff initiated a path forward to submit a separate SECY paper to request Commission approval to revise the risk-informed principles of the EP SDP. The staff's recommendation is to revise the EP SDP risk informed methodology such that only those planning standard functions (10 CFR 50.47(b)(1) – (b)(16)) that have an impact on public health and safety would be assessed a GTG safety significance. It is anticipated that the SECY paper would be submitted in 2Q FY2022.

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Baseline Security Program Revision

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Ongoing Coronavirus-2019 (COVID-19) mitigation measures throughout FY 2021 have allowed the NRC staff to conduct security baseline inspections; however, the focus for the year was to evaluate the FY 2020 lessons learned and inspector feedback from the COVID-2019 pandemic response for opportunities to revise the inspection procedures (IPs). On February 8, 2021, the NRC revised its Force-on-Force inspection program procedures, IP 71130.03 Contingency Response – Force-on-Force Testing (ML21012A329) and IP 92707 “Security Inspection of Facilities Impacted by a Local, State, or Federal Emergency Where the NRC's Ability to Conduct Triennial Force On-Force Exercises is Limited” (ML21019A452). IP 71130.03 and IP 92707 will become effective in CY2021.

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State-of-the-Art Reactor Consequence Analyses

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NRC staff authored an article providing an overview of the SOARCA Uncertainty Analysis (UA) summary project. This article was published in the American Nuclear Society journal, Nuclear Technology, Volume 207, Issue 3. Staff continues to make progress on the SOARCA UA summary report which seeks to capture the most important accident progression and consequence analysis insights from the three SOARCA UAs. This summary document is a concise reference that will support risk-informed decision-making. In addition, the UA of the unmitigated STSBO for Surry is still awaiting final submittal to publication as NUREG/CR-7262.

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Probabilistic Methodologies for Component Integrity Assessment

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In the nuclear power plant piping area, the NRC engaged in a several activities involving the xLPR code. The NRC and EPRI initiated development of xLPR Version 2.2.  This version will provide incremental improvements, including support for the latest computation framework engine, expanded preprocessor platform support, and correction of high-priority user-reported problems.  These updates began implementation and testing following the NRC's and EPRI's rigorous software quality assurance and maintenance practices.  The NRC and EPRI also held a public webinar with stakeholders to discuss various xLPR code applications, lessons learned, and formation of a user group.  In addition, the NRC and EPRI co-published a capstone report on xLPR Version 2.  The report outlines the basic design and operation of the software, its underlying models and theory, quality assurance practices, testing, trial inputs and analyses, and other development-related topics.  The NRC version of the report was published as NUREG-2247, "Extremely Low Probability of Rupture Version 2 Probabilistic Fracture Mechanics Code" (ML21225A736); the EPRI version was published as Technical Report 3002013307 of the same title.

In the applications area, the NRC published findings from sensitivity studies and analyses involving the xLPR code in Technical Letter Report, TLR‐RES/DE/CIB‐2021‐11, "Sensitivity Studies and Analyses Involving the Extremely Low Probability of Rupture Code" (ML21133A485).  The NRC also completed a two-part study using the xLPR code to quantify the effects of PWSCC on pressurized-water reactor piping systems that have received prior NRC approval for LBB under 10 CFR Part 50, Appendix A, General Design Criterion 4.  The results were published in Technical Letter Reports TLR-RES/DE/REB-2021-09, "Probabilistic Leak-Before-Break Evaluation of Westinghouse Four-Loop Pressurized-Water Reactor Primary Coolant Loop Piping using the Extremely Low Probability of Rupture Code" (ML21217A088), and TLR-RES/DE/REB-2021-14, "Probabilistic Leak-Before-Break Evaluations of Pressurized-Water Reactor Piping Systems using the Extremely Low Probability of Rupture Code" (ML21266A045).  Finally, efforts continued through the International Atomic Energy Agency to benchmark the xLPR code against other PFM codes throughout the world.  The PFM code models were evaluated and compared deterministically.  These results will be used to develop probabilistic benchmarks to further assess the codes.

In the nuclear power plant vessel area, a new and final version of the Fracture Analysis of Vessels-Oak Ridge (FAVOR) PFM code was released: FAVOR v20.1.12. The updated version allows for the modeling of as-found flaws and includes updated user and theory manuals, as well as the associated software quality assurance and verification and validation documentation and test suite.  FAVOR v20.1.12 implemented many software improvements that constitute the first step towards a full modernization of the code.  FAVOR will be sunset and replaced by a new code called FAVPRO (Fracture Analysis of Vessels - Probabilistic).  FAVPRO is being developed under a modern software quality assurance program, using Agile development practices, and implements modern parallel 0bject-oriented Fortran 2018 features.  FAVPRO will be more user friendly, more easily adapted to model new technologies and new aging phenomena, and will provide a large performance increase over FAVOR.

To support increased use of PFM in nuclear applications in general, the NRC issued draft regulatory guide DG-1382 / RG-1.245 and the accompanying technical basis NUREG/CR-7278 for public comment. This work expanded significantly upon the foundations laid in and stakeholder feedback on the prior Technical Letter Report on important aspects to be considered for PFM applications. Additionally, the NRC published two Technical Letter Reports that illustrate some of the concepts in DG-1382/RG-1.245 and NUREG/CR-7278.

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Implementing Lessons Learned from Fukushima

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No Update

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Accident Sequence Precursor (ASP) Program

For more information see existing public web page.

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Probabilistic Flood Hazard Assessment (PFHA)

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Progress on existing projects initiated via interagency agreements and cooperative research initiatives with other agencies has continued in FY2021. The first phase of the PFHA research (technical basis) is largely complete. Several reports were published in 2021 and several more are in press. This phase of the research has comprised projects focused on specific technical details of probabilistic modeling for precipitation, site-scale flooding, riverine flooding, and coastal flooding, as well as reliability of flood protection features and procedures. The second phase (pilot projects), focused on integrating the technical basis research into demonstration multi-mechanism flooding hazard assessments, is making good progress. This phase comprises three projects: (1) Local Intense Precipitation Flooding PFHA Pilot Study; (2) Riverine Flooding PFHA Pilot Study; and (3) Coastal Flooding PFHA Pilot Study. The third and final phase (guidance development) is in progress. The 6th Annual NRC PFHA Research Workshop, held virtually February 22-25, 2021, reviewed progress on NRC-funded projects and updated participants on PFHA research at other agencies. Proceedings from the 5th Annual PFHA Research Workshops was published as an NRC Research Information Letter (RIL-2021-01). NRC staff participated in virtual technical exchanges with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN). Cooperative research efforts with the Electric Power Research Institute (EPRI) and the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN) have continued under existing agreements.

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Risk Assessment of Operation Events (RASP Handbook)

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No Update

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

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The SAPHIRE development team continues to make software improvements in accordance with the software quality assurance program, which is described in NUREG/CR-7039, Volume 6, "SAPHIRE Version 8: Quality Assurance." In FY 2021, the SAPHIRE development team released two new versions of SAPHIRE, version 8.2.3 in February 2021 and version 8.2.4 in September 2021. New features include improvements to the workspace for assessing initiating event occurrences, enhanced capabilities for using multiple processors for solving sensitivity analyses, and a new viewer for reviewing and editing fault tree logic. The SAPHIRE team continues to develop a cloud-based solving platform that will better support solving large and complex models. The team’s current focus is building the external solving capability and leveraging the high-performance computing resources available at the Idaho National Laboratory. The cloud-based development is planned to continue in FY 2022. The SAPHIRE team will also continue to respond to users’ request for new features and address areas for improvement in the code.

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Standardized Plant Analysis Risk Models (SPAR)

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During FY 2021 the staff continued to maintain and improve SPAR models to reflect the as-built-as operated nuclear power plants and continued to provide technical support for SPAR model users and risk-informed programs. In an effort to meet the objective or reflecting the as-built-as-operated plant, a total of 173 SPAR model formal modifications were completed. These modifications included: routine updates to reflect recent plant changes (6 models were updated with the latest information from licensee PRA models), incorporation of new logic associated with external events, SPAR model modifications to incorporate FLEX modeling, and modifications to support the Significance Determination Process (SDP) or Events and Conditions Assessment (ECA) analyses. For new reactor designs, the staff finalized the reactor, at-power, internal events plant specific PRA model for Vogtle Units 3 & 4 (AP1000).

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

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The technical work on the Level 3 PRA Project is nearing completion, with a target date for release of the results in FY23 through early FY24. Internal reports on the reactor, at-power, Levels 1, 2, and 3 PRAs for internal events and floods; the reactor, at-power, Level 1 and 2 PRAs for internal fires, seismic events, and high winds; and the screening analysis of other hazards are complete. The reactor, low power and shutdown (LPSD) for internal events, Level 1 and 2 PRAs are nearing completion. Internal reports have been completed on the reactor, at-power, Level 3 PRAs for internal fires, seismic events, and high winds, which are currently undergoing management review. Substantial progress has been made on completing the spent fuel pool Level 1 and Level 2 PRAs (all hazards). Work is continuing on the Level 3 PRAs for the reactor, LPSD, for internal events and the spent fuel pool, as well as on the Level 1, 2, and 3 Dry Cask Storage PRA. The staff has actively begun performing the integrated site risk approach building off the completed work for reactor, at-power, Levels 1, 2, and 3 PRAs for internal events and floods.

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Data Collection for Human Reliability Analysis (HRA)

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RES expanded SACADA's capability to collect operator performance in job performance measures (JPMs). While continuing its outreach to nuclear power plants to use SACADA to collect JPM performance information, RES plans to expand the SACADA tool to collect operator performance data in written exams. Also, RES reviews literature and operational experience, and plans to collaborate with nuclear power plants to collect the human performance information related to the operations of nuclear facilities and document the information in the IDHEAS-DATA report (draft, ADAMS Accession Number: ML20238B982). RES contracted the Pacific Northwest National Laboratory to review the IDHEAS-DATA report and develop recommendations on the estimates of action timing. RES continues to collaborate with domestic and international organizations to use simulator data to inform HRA.

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Human Reliability Analysis (HRA) Methods and Practices

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Completed IDHEAS General Methodology (IDHEAS-G)

  • Published the final report (NUREG-2198)
  • Presented IDHEAS-G to ACRS full Committee on 2/24/2021 and received ACRS endorsement of the method

Rolled out the IDHEAS Method for Events and Conditions Assessment (IDHEAS-ECA).

    • IDHEAS-ECA is a human reliability analysis (HRA) method that is based on the General Methodology of an Integrated Human Event Analysis System (IDHEAS-G). The method is intended to be used in event and condition assessment (ECA) of nuclear power plants. IDHEAS-ECA supports probabilistic risk assessment (PRA) applications by analyzing human events and estimating human error probabilities. The application scope of IDHEAS-ECA is broad because the performance-influencing factor (PIF) structure, which models the context of a human failure event (HFE), is comprehensive. The method covers all the PIFs in existing HRA methods and the factors reported in broad literature and in nuclear-specific human events. Because of the comprehensiveness of the PIF structure, IDHEAS-ECA can model the context of HFEs inside and outside NPP control rooms—including the use of flexible and coping strategies (FLEX) equipment—and during different plant operating states (i.e., at-power and shutdown). IDHEAS-ECA can be used in PRA applications; for example, the significance determination process (SDP), accident sequence precursor (ASP) program, and risk-informed licensing reviews. To facilitate the use of IDHEAS-ECA, the NRC staff developed the IDHEAS-ECA Software Tool to calculate the HEPs of HFEs analyzed in the IDHEAS-ECA process.
    • Presented IDHEAS-ECA at 2021 Regulatory Information Conference
    • Presented IDHEAS-ECA at 4/8/2021 public meeting
    • Presented IDHEAS-ECA to EPRI HRA User Group meeting
    • Conducted and completed public comments on IDHEAS-ECA report RIL-2020-02
    • Collected feedback from multiple sources on using the IDHEAS-ECA method and software.

Completed development of guidance of HRA dependency analysis (IDHEAS-DEP)

    • IDHEAS-DEP is a new method for HRA dependency analysis. It is based on the dependency model in IDHEAS-G and the calculation of human error probabilities (HEPs) in IDHEAS-ECA.
    • IDHEAS-DEP analyzes dependency between two human failure events by assessing dependency factors inherited from the relationships between the events; the impact of the dependency factors on HEPs is represented by performance influencing factors. The relationships, dependency factors, and impacted performance influencing factors together provide explanation on why there is dependency and how the dependency increases the likelihood of human errors.
    • IDHEAS-DEP includes three stages of dependency analysis: Pre-Determination analysis assesses relationships between human failure events; Screening Analysis assesses dependency factors and assigns screening values of dependent HEPs; Detailed Analysis assesses dependency factors and calculates dependent HEPs using IDHEAS-ECA.
    • The IDHEAS-DEP guidance document was developed collaboratively by NRC staff from RES, NRR, Regions, and industry participants from the Electric Power Research Institute and their contractors.

The following activities are on-going:

  • The NRC staff are consolidating comments and suggestions from the NRC and industry PRA/HRA practitioners on improving IDHEAS-ECA; the staff are updating the IDHEAS-ECA report (RIL-2020-02) which will be published as a NUREG.

The NRC staff are finalizing the IDHEAS-DEP report for publication as a Research Information Letter in 2021.

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National Fire Protection Association (NFPA) Standard 805

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See Summary Description

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

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There has not been a significant change in the technical knowledge or guidance for GL 2004-02 in the last year and it is unlikely that there will be significant changes moving forward. The most significant event related to risk-informed resolution of GL 2004-02 in the last year was the issuance of the risk-informed amendment for Vogtle Units 1 and 2 and the associated ACRS letter that determined that the methodology used by the licensee was acceptable. The staff continues working with individual licensees on closure of GL 2004-02. Closure requires plants to address both strainer and in-vessel issues. NRC Staff in-vessel guidance risk-informs plant-specific resolution by describing several defined paths licensees may follow to resolve GL 2004-02 depending on their plant-specific configuration. The guidance considers the relative risk and available safety margin for each configuration and considers these in its recommendations for the type and depth of supporting information required for closure. Currently, six plants have opted to use fully risk-informed evaluations to close GL 2004-02. Of the six, Vogtle and STP have received amendments and exemptions to close out the issue. Approximately half of the PWR fleet have already closed the GL. All plants, with the exception of the six risk-informed plants, used or plan to use deterministic methods to resolve the generic letter. Some plants that have not yet closed the GL plan to use a transition break size methodology that accounts for the risk associated with breaks of different sizes. The staff continues to review submittals as they are submitted by each licensee. With respect to risk-informed submittals, the staff is currently reviewing a Callaway LAR and is awaiting a supplement from Wolf Creek to address acceptance issues.

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

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See Summary Description

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

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No Update

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Graded Approach to the Use of Safety Significance in the Low Safety Significance Issue Resolution Process

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The staff completed a major overhaul and updated to NRR Office instruction COM-106, Revision 6, "Technical Assistance Request (TAR) Process."(ML19228A001). Revision 6 established a more structured process with enhanced instructions, process-related templates, a safety significance determination tool to address questions raised by other agency organizations and then to inform timely and resource appropriate regulatory decision making commensurate with the significance of the underlying issue. The TAR process can be used to inform the discission to suspend inspection and evaluation under the Very Low Safety Significance Issue Resolution (VLSSIR) process described in Inspection Manual Chapter (IMC) 0612 Appendix B, "Issue Screening Directions," to continue analysis to determine whether the issue is part of the current licensing basis for that licensee, or to determine whether the issue should be referred to the backfit process for consideration.

The staff completed a VLSSIR process self-assessment (i.e., Results of a Calendar Year 2020 Reactor Oversight Process Self-Assessment Effectiveness Review of the Very Low Safety Significance Issue Resolution Process(ML21070A334)). The self-assessment concluded that the VLSSIR process is meeting its goal, having reduced the number of unresolved inspection items and the agencies focus on issues having very low safety significance, thereby allowing those resources to be allocated to matters having greater significance.

The staff revised IMC 0612 Appendix B, "Issue Screening Directions" (ML21203A356) to address feedback received during the initial rollout of the VLSSIR process and to align the VLSSIR process with the TAR process.

In FY2021, as part of the Low Safety Significance Issue Resolution initiative, the NRC developed the Risk-Informed Process for Evaluations (RIPE) to resolve very low safety significance compliance issues commensurate with their risk significance using existing regulations under 10 CFR 50.12 or 10 CFR 50.90 and risk information. Its objective is to focus NRC and licensee resources on the most safety significant issues by addressing low safety compliance issues in an efficient and predictable manner consistent with our Principles of Good Regulation. The process plans to leverage existing regulations and risk initiatives to allow licensees to justify plant-specific exemptions or license amendment requests using a streamlined NRC review process. In addition, RIPE incentivizes further development and use of probabilistic risk assessment and risk-informed applications. The RIPE guidance was approved for use on January 7, 2021 (ADAMS Accession No. ML21006A324). If a licensee elects to use RIPE to resolve a non-compliance, it would characterize the risk associated with the proposed exemption or amendment and submit a request to the NRC for approval. Licensees can use the RIPE process to justify plant-specific licensing actions to address the issue without imposing undue burden. Most recently, on June 30, 2021, the RIPE process was expanded to allow licensees with additional approved risk-informed informed initiatives to use the process (ADAMS Accession No. ML21180A011). 

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Guidance for Unattended Opening Evaluations

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The proposed revision to NEI 09-05 was withdrawn by the Nuclear Energy Institute based on the lack of technical justification. Throughout FY 2021, Sandia National Laboratory conducted performance testing for accessibility of various opening sizes. The analysis is scheduled to be completed with the report to be issued in the second quarter of FY 2022. The next steps for this topic will be determined based on the results of the report.

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Risk-Informed Adversary Timeline Calculations

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An industry proposal was submitted on July 9, 2021. The NRC staff conducted an initial review and Identified the methodology used to incorporate passive physical protection program features had value; however, the methodology used to calculate the impact of active delay features associated with security force response needs to be substantiated with a technical justification. Staff plans to continue to engage industry representatives in Q1 FY 2022 on the next steps.

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Transition from Physical Security Plan to Safeguards Contingency Plan

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No activity was performed in FY2021. However, on September 14, 2020, the NRC issued SFAQ 20-01SFAQ 20-01, “Utilization of Armed Responders as Compensatory Measures Including During Transition from Physical Security Plan (PSP) to Safeguards Contingency Plan (SCP).” This supplements and clarifies the information in SFAQ 09-07, dated February 15, 2010 and SFAQ 09-07, revision 1, dated December 20, 2010, related to armed responders being used for compensatory measures.

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Emergency Preparedness (EP) Program Review 24-Month Frequency Performance Indicators Development to Satisfy 10 CFR 50.54(t) Requirements

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On June 1, 2021, the NRC published Revision 6 of Regulatory Guide (RG) 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors.” RG 1.101 endorses and updates guidance that is available to licensees and applicants on methods acceptable to the NRC staff for complying with the NRC’s regulations for emergency response plans and preparedness at nuclear power reactors. It endorses the Nuclear Energy Institute’s White Paper entitled, “Implementing a 24-Month Frequency for Emergency Preparedness Program Reviews.” The White Paper discusses risk-significant activities that will achieve compliance with 10 CFR 50.54(t)(ii), including the monitoring of performance indicators, adequacy of interfaces with State and local governments, and identification of a change in personnel, procedures, equipment, or facilities that potentially could adversely affect EP.

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Page Last Reviewed/Updated Tuesday, October 11, 2022