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# 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*:

**Objective**

**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**

**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:	                 - measurable parameters to monitor performance
                - objective criteria to assess performance
                - flexibility to allow licensees to determine how to meet the performance criteria
                - no immediate safety concern as a result of failure to meet the performance criteria 

Risk-informed activities for operating reactors occur in five broad categories:
              - applicable regulations
              - licensing process
              - revised oversight process
              - regulatory guidance
              - risk analysis tools, methods, and data

The activities in these categories are derived from the Commission's policy statements and guidance, and include revisions to technical requirements in the regulations; risk-informed technical specifications; a new framework for inspection, assessment, and enforcement actions; guidance on other risk-informed applications (e.g., in-service inspections); and improved standardized plant analysis risk models.

**Goals**

**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:
            - Systems-Theoretic Process Analysis (STPA) for Digital Nuclear Safety System Evaluation
            - Revisions to NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness for NPP
            - Revision to NUREG/CR-7002, "Criteria for Development of Evacuation Time Estimate Studies"
            - Power Reactor Cyber Security Program Improvements
            - Consequence-based Security for Advanced Reactors
            - Revision of the Emergency Preparedness Significance Determination Process
            - Baseline Security Program Revision
            - State-of-the-Art Reactor Consequence Analyses
            - Probabilistic Methodologies for Component Integrity Assessment
            - Implementing Lessons Learned from Fukushima
            - Accident Sequence Precursor (ASP) Program
            - Probabilistic Flood Hazard Assessment (PFHA)
            - 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
            - 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
            - Graded Approach to the Use of Safety Significance in the Low Safety Significance Issue Resolution Process
            - Guidance for Unattended Opening Evaluations
            - Emergency Preparedness (EP) Program Review 24-Month Frequency Performance Indicators Development to Satisfy 10 CFR 50.54(t) Requirements

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## Systems-Theoretic Process Analysis (STPA) for Digital Nuclear Safety System Evaluation

**Summary Description**

**Summary Description**
**Summary Description**

NRC staff are investigating the risk-informing potential of Systems-Theoretic Process Analysis (STPA) in the NRC’s regulatory processes. STPA promises to address two of the questions in the NRC’s risk triplet: “what can go wrong?” and “what are the consequences?” Traditional hazard analysis methods (e.g., Failure Modes and Effects Analysis or Fault Tree Analysis) have not proven effective in identifying all systemic causes (e.g., unwanted interactions) for Digital I&C applications, which is the potential advantage of STPA over these traditional methods. Current practice for the safety evaluation and analysis of digital safety systems in nuclear power plants is challenged with changes in the technology, increasing digitization, and increasing interdependencies and interactions across systems and components. This research effort seeks to enable the NRC staff with evaluation tools that improve independent safety analyses, Safety Analysis Report reviews, with consistencies of results for digital systems in a risk-informed environment.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2020**

This research effort will occur between October 2020 and September 2021. Interactive lectures, seminars, and workshops will be developed for the NRC staff. A report detailing the results is expected by September 2021. The report is expected to include information on the limits within which the STAMP based methods can be consistently applied.

**FY 2021**

NRC staff participated in a series of interactive seminars, workshops, and discussions to understand the risk-informing potential of Systems-Theoretic Accident Modeling and Processes (STAMP)-based methods in the NRC’s regulatory processes and the limits within which these methods can be consistently applied. These seminars, workshops, and discussions helped the staff understand how STAMP models are used within the STPA process to identify “what can go wrong?” and “what are the consequences?” Two reports were issued: Investigation of the Use of Causal Analysis based on System Theory (CAST), and Investigation of the Use of System-Theoretic Process Analysis.

**FY 2022**

NRC staff and contractors completed follow-on work to address recommendations from FY 2021 activities.  A scaled-up project began in late FY2022. This project aims to help NRC staff understand the limitations within which risk-related information generated from an STPA process can be consistently evaluated.

Information is to be posted on: Operating Reactors Sub-Arena Webpage.

**FY 2023**
No Update

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

**Summary Description**

**Summary Description**
In December 2019, the NRC and FEMA issued NUREG-0654/FEMA-REP-1, Revision 2, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."  This document provides the overarching guidance for developing and evaluating onsite and offsite emergency plans for nuclear power plants and for State, Local and Tribal government emergency response organizations (EROs). The guidance document was revised to: integrate 35 years of lessons learned within the radiological emergency preparedness program, consolidate and clarify previous guidance.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2020**

The NRC staff continues to review license amendment requests (LAR) related to licensee ERO staffing levels and augmentation timing for licensee emergency responders. In addition, the NRC staff is conducting industry outreach in anticipation of LAR submittals to relocate the implementing procedures for emergency plans. This provides greater licensee flexibility to implement changes to their EP programs without needing NRC’s approval.

**FY 2021**

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 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.

**FY 2022**

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 November 16, 2021, the NRC granted a LAR to the Perry Nuclear Power Plant, Unit 1, to risk inform the ERO staffing composition and increase the staff augmentation response time of certain ERO positions from 30 minutes to 60 minutes and from 60 minutes to 90 minutes. On May 5, 2022, the NRC granted a LAR (ADAMS Accession No ML21286A782) to Beaver Valley Power Station, Units 1 and 2, to change the ERO staffing composition and extend staff augmentation time from 30 minutes and 60 minutes to 60 minutes and 90 minutes.

**FY 2023**
No Update

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

**Summary Description**

**Summary Description**
The NRC staff developed a draft Revision 1 to NUREG/CR-7002, "Criteria for Development of Evacuation Time Estimate Studies," (ML20233A700) based on risk insights from an applied research study on evacuation time estimates (ETE) published in March 2020, as NUREG/CR-7269, "Enhancing Guidance for Evacuation Time Estimate Studies," (ML20070M158). The guidance will be revised to update the considerations for when to include a shadow evacuation, the use of manual traffic control, establishing appropriate boundary conditions, the use of dynamic traffic assignment, updates to modeling adverse weather, and various other parameters of importance.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2020**

Draft Revision 1 to NUREG/CR-7002 was released for public comment on August 27, 2020. NSIR staff held a public meeting on the proposed revisions on September 16, 2020. The staff expects to publish Revision 1 by the end of CY 2020 in time to support decennial updates to ETEs which will start around April 1, 2021.

**FY 2021**

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.

**FY 2022**

The NRC has begun utilizing Revision 1 of NUREG/CR-7002, "Criteria for Development of Evacuation Time Estimate Studies," for the decennial review of licensees’ updated ETEs based on the latest census data.

**FY 2023**
No Update

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## Power Reactor Cyber Security Program Improvements

**Summary Description**

**Summary Description**
**Summary Description**

Based on lessons learned from implementation and oversight of the NRC's cyber security requirements (10 CFR 73.54), the NRC staff is working to improve the efficiency and effectiveness of the power reactor cyber security program.  Specifically, the NRC staff is seeking to further risk-inform the program in areas such as the following: Critical Digital Asset Determination, Protection (Cyber Control Implementation), and Assessment, as well as Cyber Inspection Oversight Program.  Examples of risk-informed activities being performed under this effort include: (1) Evaluating cyber security controls to ensure that they are appropriate for the types of technology (i.e., Information Technology versus Operational Technology) implemented in the licensee's digital infrastructure; and, (2) Providing greater definition to the guidance to screen cyber security controls that are not needed or applicable based on risk-informed analysis (e.g., Critical Digital Asset system capability and complexity).

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2020**

NSIR has issued letters on industry use of guidance for emergency preparedness digital assets (ADAMS Accession No. ML20129J981), balance of plant digital assets (ADAMS Accession No. ML20209A442), and for important-to-safety and safety-related digital assets (ADAMS Accession No. ML20223A256).

**FY 2021**

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.

**FY 2022**

In 2021, the NRC completed inspection of licensees’ full implementation of their cybersecurity programs. The staff subsequently revised Inspection Procedure 73310.10, “Cybersecurity” to shift inspections away from cybersecurity program implementation and instead focus on program maintenance. In June 2022, the NRC issued letters documenting the staff’s review and approval for use of NEI 10-04, Revision 3, “Identifying Systems and Assets Subject to the Cyber Security Rule”, and NEI 13-10, Revision 7, “Cyber Security Control Assessments”. These revisions incorporated the risk-informed changes proposed in the white papers discussed in the FY 2021 and FY2020 updates above.

**FY 2023**
In 2023, the NRC issued Regulatory Guide 5.71, Revision 1, "Cybersecurity Programs for Nuclear Power Reactors," which incorporated lessons learned from initial publication of the regulatory guide and took into consideration the latest revisions to the National Institute of Standards and Technology's cybersecurity standards. Additionally, the staff published Regulatory Guide 5.83, Revision 1, "Cybersecurity Event Notifications," which contains risk-informed and clarified guidance on report of nuclear power plant cyber events. More information can be found at the NRC’s webpage for Cyber Security Requirements.

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## Consequence-based Security for Advanced Reactors

**Summary Description**

**Summary Description**
The NRC first issued its Policy Statement on the Regulation of Advanced Reactors on July 8, 1986, in Volume 51 of the Federal Register, page 24643 (51 FR 24643), with an objective to provide all interested parties, including the public, with the Commission's views concerning the desired characteristics of advanced reactor designs. The NRC revised the policy statement in 2008 (73 FR 60612; October 14, 2008) to specifically include attributes related to physical security that should be considered in advanced designs. In particular, the Commission observed that it would be in the interest of the public as well as the design vendors and the prospective license applicants to address security issues early in the design stage.

Accordingly, and with the resurgence in potential for applications for advanced reactor designs, the NRC is analyzing associated physical security requirements that are commensurate with the potential consequences to public health and safety and common defense and security from the possession and use of special nuclear material at these facilities.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2018**

The staff transmitted a Notation Vote paper (ML18052B032) to the Commission in August, 2018. The paper provides the Commission options for addressing physical security requirements for advanced reactors, and recommends a limited-scope revision of regulations and guidance to reflect the relative risks posed by the technology.

**FY 2019**

On November 19, 2018, the Commission approved the staff's recommendation (Option 3) to initiate a limited-scope rulemaking according to the rulemaking plan (as modified) in Staff Requirements Memorandum SECY-18-0076," Staff Requirements SECY-18-0076, Options and Recommendation for Physical Security for Advanced Reactors." On July 16, 2019, the staff issued the Regulatory Basis for public comment. The comment period closed on August 15, 2019. Nine comments were received and are to be addressed in the proposed rule. Currently, as scheduled, the proposed rulemaking and draft guidance will be provided to the Commission in January 2021, and if approved, will be published for public comment. The Final Rule and Final Guidance are currently scheduled to be provided to the Commission in May 2022.

**FY 2020**

No update.

**FY 2021**

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.

**FY 2022**

On August 2, 2022 the NRC staff submitted SECY-22-0072,"Proposed Rule: Alternative Physical Security Requirements for Advanced Reactors" (RIN 3150-AK19) to the Commission for review and decision. The SECY paper and proposed rule package was released to the public on August 15, 2022. The purpose of this paper is to obtain Commission approval to publish in the Federal Register for public comment the draft proposed rule to establish voluntary alternative physical security requirements for advanced reactors. This paper provides the staff’s recommended draft proposed rule for revising the regulations, primarily 10 CFR 73.55, "Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage," to offer voluntary performance-based alternatives for meeting certain physical security requirements for advanced reactors. In the context of this proposed rulemaking, advanced reactors include non-light-water reactors (non-LWRs) and light-water small modular reactors (SMRs) to be licensed under 10 CFR Part 50 or 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants." Applicants and licensees for these facilities that meet the proposed radiological consequence-based eligibility criterion would have the option to consider implementing one or more of these alternatives rather than complying with certain existing prescriptive physical security requirements under 10 CFR 73.55.

**FY 2023**
On March 1, 2023, the NRC staff submitted SECY-23-0021, "Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 3150-AK31)," to the Commission for review and decision. The SECY paper and proposed rule package was released to the public on March 6, 2023. The purpose of this paper is to obtain Commission approval to publish in the Federal Register for public comment on the draft proposed rule to establish a voluntary risk-informed, performance-based, and technology-inclusive regulatory framework for commercial nuclear plants. This paper includes the staff’s recommended draft proposed revisions for 10 CFR 73.55, "Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage," to establish a new voluntary, technology-inclusive, consequence-based approach for a range of security issues, including physical security, cybersecurity, and access authorization for future commercial nuclear plants."

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## Revision of the Emergency Preparedness Significance Determination Process

**Summary Description**

**Summary Description**
The NRC staff is evaluating possible changes to the process to ensure significance of emergency preparedness inspection findings are appropriately characterized using risk principles. The NRC initiated a focused self-assessment of the significance determination process for emergency preparedness-related licensee performance deficiencies, with the objective of determining whether improvements can or should be made to this established process.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2018**

The staff expects to complete the self-assessment in February 2019 and summarize the results in the annual Reactor Oversight Program's self-assessment paper to the Commission.

**FY 2019**

NSIR completed the Emergency Preparedness (EP) Significance Determination Process focused self-assessment (EP SDP FSA) in February 2019, including recommendations for further action/review (ML18331A374). NSIR staff assisted in incorporating these recommendations into the Reactor Oversight Process Enhancement Project and contributed to the development of SECY-19-0067, "Recommendations for Enhancing the Reactor Oversight Process to obtain Commission direction on the higher priority recommendations. NSIR staff continues to work on other recommendations that do not require Commission approval, for example, revising the EP training program, developing tools for better communication, sharing of regional operating experience, and formalizing knowledge management. The staff expects to complete the recommendation(s) that do not require Commission direction during FY 2020 and FY 2021.

**FY 2020**

The NRC staff revised the Emergency Preparedness (EP) inspector training qualification program and conducted a pilot in October 2019 and a training session in May 2020. The staff developed an online web-based EP Fundamentals course that was available in its internal training system on September 28, 2020. Staff also developed and is using a knowledge management SharePoint file to capture discussions related to EP inspections. Regional EP inspectors regularly add content as issues arise.

**FY 2021**

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.

**FY 2022**

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. The SECY paper was submitted in 4Q FY2022.

**FY 2023**
In February 2023, the Commission approved the staff's recommendation to modernize the Significance Determination Process in the EP cornerstone of the NRC’s Reactor Oversight Process that uses a risk-informed methodology to evaluate the significance of EP inspection findings and align the findings with the associated risk. Once implemented, this revision will enable resource efficiencies by focusing oversight activities on those with the most significant impacts.

 |  

## Baseline Security Program Revision

**Summary Description**

**Summary Description**
The NRC staff continues to use operational and risk-informed insights to ensure that the concept of "high assurance" of adequate protection is equivalent to "reasonable assurance" when it comes to determining what level of regulation is appropriate. Significant accomplishments to date include: revising the associated significance determination process (SDP) to appropriately categorize findings related to the handling of safeguards material; incorporating risk-informed insights to revise the staff's site access authorization process to improve the program's efficiency and effectiveness, standardizing requirements across the agency, and reducing the training requirement to a five-year periodicity.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2018**

In FY 2019, the staff plans to: continue with the SDP revision process to implement additional changes to align the security inspection program with the Reactor Oversight Process, reflect the staff's commitment to the risk-informed process and ensure findings are characterized using risk principles where appropriate; review and revise the baseline security inspection procedures to streamline the inspection process and apply the concept of reasonable assurance; and use risk-informed insights during the evaluation of internal and external stakeholder recommendations for proposed changes to the inspection manual chapters and inspection procedures.

**FY 2019**

NSIR staff continued to use risk-informed insights to ensure that the concept of "high assurance" of adequate protection, as found in NRC security regulations (10 CFR 73.20 and 73.55), is equivalent to "reasonable assurance" when it comes to determining what level of regulation is appropriate. To accomplish this goal, NSIR staff further refined how it revises the significance determination process, implementing additional changes to align the security inspection program with the Reactor Oversight Process, ensuring findings are characterized using risk principles where appropriate. In addition, staff reviewed and revised multiple baseline security inspection procedures to streamline the inspection process, using risk-informed insights during the evaluation of internal and external stakeholder recommendations for proposed changes to the inspection manual chapters and inspection procedures.

**FY 2020**

Due to COVID-19 restrictions, on-site security inspections were interrupted. In an effort to continue to provide regulatory oversight, security inspectors were able to partially complete baseline security inspections remotely. Once travel restrictions were lifted, security inspectors were able to complete the remaining portions of the inspection procedure. NRC staff will leverage lessons learned during the COVID-19 public health emergency to improve the efficiency of the baseline inspection program where appropriate.

**FY 2021**

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.

**FY 2022**

In response to the COVID-19 lessons learned working group, NRC staff are incorporating efficiencies identified during the COVID-19 public health emergency into inspection procedures. Security oversight efforts are being resumed insofar as the local COVID-19 conditions supported additional NRC staff onsite to focus on the licensee training programs to identify any best practices for incorporation into regulatory guidance. The NRC staff submitted SECY-22-0066, "Calendar Year 2022 Triennial Force-on-Force Inspection Program Status Update and Lessons-Learned," that included a summary of force-on-force inspection program improvements identified during the oversight process.

**FY 2023**
In SECY-23-0032, "Reactor Oversight Process Self-Assessment for Calendar Year 2022," dated April 7, 2023, the U.S. Nuclear Regulatory Commission (NRC) staff committed to assessing the Baseline Security Significance Determination Process (BSSDP), Inspection Manual Chapter (IMC) 0609 Appendix E, Part I, "Baseline Security Significance Determination Process for Power Reactors," dated November 2022. Per SECY-23-0032, this assessment will determine whether there are any aspects of the BSSDP that can be improved or further risk-informed. The staff developed a charter to outline key objectives and goals and solicited working group member nominations to begin the assessment.

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## State-of-the-Art Reactor Consequence Analyses

**Summary Description**

**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.

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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2018**

In FY 2018 the staff completed calculations for an updated UA of the Surry unmitigated short-term station blackout scenario. Updated calculations were important to leverage insights from the Sequoyah UA (NUREG/CR-7245).

**FY 2019**

Previously, 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 and a UA for the Peach Bottom unmitigated long-term station blackout is documented in NUREG/CR-7155. In FY 2019, staff completed a UA of the unmitigated STSBO for Surry and prepared formal documentation in a NUREG/CR report. Staff began development of a Research Information Letter to formally document the numerous benefits and uses of the SOARCA project beyond its original objectives including uses by the NRC, reactor licensees and applicants, domestic and international regulatory and research organizations, academia, and other stakeholders. Finally, to complete the work on the SOARCA Project, staff began development of a summary report capturing the most important accident progression and consequence analysis insights from the three SOARCA uncertainty analyses is being compiled.

**FY 2020**

In FY 2020 staff completed a Research Information Letter (RIL-2020-03) that formally documents the numerous benefits and uses of the SOARCA project beyond its original objectives including uses by the NRC, reactor licensees and applicants, domestic and international regulatory and research organizations, academia, and other stakeholders. Staff also completed an updated revision of the SOARCA brochure, NUREG/BR-0359, Rev. 3, "Modeling Potential Reactor Accident Consequences" to capture updates from the more recently completed SOARCA uncertainty analyses. The UA of the unmitigated STSBO for Surry is still awaiting final submittal to publication as NUREG/CR-7262. 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 UA. This summary document is a concise reference that will support risk-informed decisionmaking.

**FY 2021**

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.

**FY 2022**

No update

**FY 2023**
NRC staff published the last two reports documenting the SOARCA body of work. NUREG-2254, "Summary of the Uncertainty Analyses for the State-of-the-Art Reactor Consequence Analyses Project," summarizes the most important accident progression, consequence analysis, and methodological insights from the three SOARCA Uncertainty Analyses (UAs). NUREG/CR-7262, "State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Short-Term Station Blackout of the Surry Power Station," documents the final UA.

 |  

## Probabilistic Methodologies for Component Integrity Assessment

**Summary Description**

**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) development, distribution, and application of the new Fracture Analysis of Vessels – Probabilistic (FAVPRO) code, (2) distribution, maintenance, and application of the Extremely Low Probability of Rupture (xLPR) code, and (3) updating Regulatory guide RG 1.245 on probabilistic fracture mechanics (PFM).

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. FAVOR’s successor FAVPRO includes all the features of FAVOR but also adds state-of-practice ASME flaw analysis capabilities and material embrittlement models, as well as enhanced software quality assurance. In a separate activity, the NRC and the Electric Power Research Institute (EPRI) have collaboratively developed the xLPR Version 2 PFM code to assess the effects of active degradation mechanisms on nuclear power plant piping systems. xLPR Version 2 code development was spurred, in part, as a result of primary water stress-corrosion cracking (PWSCC), which has been discovered in piping systems previously approved for leak-before-break (LBB) in accordance with NRC Standard Review Plan (SRP) Section 3.6.3 and 10 CFR Part 50, Appendix A, General Design Criterion 4. As a result of the discovery of the PWSCC, an extremely low probability of rupture could no longer be demonstrated by the deterministic methods outlined in SRP Section 3.6.3, because these methods assume the absence of active degradation mechanisms. Instead, it was decided that it is more appropriate to quantity the effects of PWSCC on the probability of piping system rupture using PFM analysis techniques, such provided in the xLPR code. Technical development of the full production version of the xLPR code has been completed. Various activities were undertaken during development 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 v20.1.12, FAVPRO, and xLPR v2, 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 developed Regulatory Guide 1.245 for performing and documenting PFM analyses for regulatory applications.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2018**

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 have agreed on a framework for U.S. domestic distribution of xLPR and 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.

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 event of interest.

**FY 2019**

In the nuclear power plant piping area, the NRC completed sensitivity studies concerning the Extremely Low Probability of Rupture (xLPR) Version 2 code to identify which of its models and inputs contribute most to uncertainty in its outputs. As part of this effort, NRC-sponsored sensitivity study methodologies were compared against industry-sponsored methodologies, and it was demonstrated that all of the methodologies produced similar findings that the crack initiation models drive uncertainties. The NRC also began to apply xLPR Version 2 to conduct probabilistic analyses to quantify the risks of primary water stress-corrosion cracking in pressurized-water reactor piping systems which have received NRC approval for leak-before-break under 10 CFR Part 50, Appendix A, General Design Criterion 4. The NRC analysis methodologies and results are being compared against results generated independently by industry-sponsored analysts also using xLPR Version 2. In addition, the NRC transitioned xLPR Version 2 from the development to the maintenance phase of the software lifecycle under a rigorous quality assurance program. To strengthen confidence in the xLPR Version 2 analysis results, the NRC initiated an effort to benchmark the code against the probabilistic fracture mechanics analysis code, PASCAL-SP (Probabilistic Fracture Mechanics Analysis of Structural Components in Aging Light-Water Reactors - Stress-Corrosion Cracking at Welded Joints of Piping), which has been independently developed by the Japan Atomic Energy Agency. Comparison of the analysis results produced by the two codes indicate that they provide similar results when exercising their core fracture mechanics models. The NRC also applied xLPR Version 2 to assess conservatism in loss-of-coolant accident frequencies presented in NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process." The results were discussed with the Advisory Committee on Reactor Safeguards in relation to ongoing NRC evaluations of licensee actions in response to Generic Safety Issue 191, "Assessment of Debris Accumulation on PWR Sump Performance."

In the nuclear power plant vessel area, analysis work using the Fracture Analysis of Vessels-Oak Ridge (FAVOR) Version 16.1 code to address potential safety issues related to shallow subsurface flaws is nearing completion, including investigations of warm pre-stress effects, cladding residual stress modeling, and an assessment of risk-optimized pressure-temperature corridors for vessel heat-up and cool-down. In addition, NRC staff received extensive training in the use and development of the FAVOR probabilistic vessel integrity assessment code. The Reactor Embrittlement Archive Project (REAP) database has also been reactivated as a web-based search tool for worldwide users.

Finally, a Technical Letter Report on important aspects to be considered for probabilistic fracture mechanics was published early in 2019. It lays the foundation for the upcoming publication of the draft probabilistic fracture mechanics regulatory guide and its associated technical basis. The NRC developed a proposal for a graded approach for regulatory submittals based on probabilistic fracture mechanics, largely based on stakeholder feedback received at public meetings during 2019. This approach was presented at the 3rd International Seminar on Probabilistic Methods for Nuclear Applications (ISPMNA'19), which NRC hosted in October 2019 in Rockville, MD.

**FY 2020**

In the nuclear power plant piping area, the NRC engaged in a several activities concerning the Extremely Low Probability of Rupture (xLPR) probabilistic fracture mechanics (PFM) code. Foremost, the NRC publicly released xLPR Version 2.1 to users worldwide. Public release was made possible by re-negotiating distribution arrangements with the Electric Power Research Institute, the code's co-developer. The NRC sponsored five public webinars to announce public release of the xLPR code and provide information to new users. Recordings of these webinars are available for public viewing on the NRC YouTube channel. Public release was also supported by development of xLPR Version 2.1, which marked the first incremental release of the code. Version 2.1 was built and tested following rigorous software quality assurance practices and includes updated leak rate calculations through incorporation of new water property routines supplied by the National Institute of Standards and Technology, among other updates. Public release of the xLPR code garnered significant global interest with further growth of the user base expected. Members of the public may request the code by following the procedures on the NRC's Obtaining the Codes webpage. The NRC also conducted a host of studies using the code to quantify the risks of primary water stress-corrosion cracking in pressurized-water reactor piping systems which have received NRC approval for leak-before-break under 10 CFR Part 50, Appendix A, General Design Criterion 4. In addition, a draft NUREG manuscript, which summarizes the entire xLPR Version 2.0 development effort and underlying theory and operations of the code, was prepared. Finally, a project was initiated through the International Atomic Energy Agency to benchmark xLPR against other PFM codes throughout the world.

In the nuclear power plant vessel area, progress was made towards developing a new version of the Fracture Analysis of Vessels-Oak Ridge (FAVOR) PFM code, named Version 20.1. The updated version will allow for the modeling of as-found flaws and include updated user and theory manuals, as well as the associated software quality assurance and verification and validation documentation and test suite.

To support increased use of PFM in nuclear applications in general, the NRC readied a draft regulatory guide and accompanying technical basis NUREG 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 brought together PFM experts from around the world by hosting the 3rd International Seminar on Probabilistic Methods for Nuclear Applications in Rockville, MD.

**FY 2021**

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.

**FY 2022**

There were many activities completed in the nuclear power plant piping integrity area. The NRC staff and EPRI completed technical development of xLPR v2.2. The updated version will be publicly released in early FY 2023. It includes a faster pre-processor, updated operating environment support, and an increased range of validity of its axial crack opening displacement model. It also includes correction of errors that provide more accurate results for inservice inspection effects, fatigue-related calculations, and circumferential crack opening displacement and rupture calculations. The NRC staff also used the xLPR code to complete additional probabilistic leak-before-break studies considering the effects of primary water stress-corrosion cracking. The results were published in Technical Letter Report TLR-RES/DE/REB-2021-14-R1, "Probabilistic Leak-Before-Break Evaluations of Pressurized-Water Reactor Piping Systems using the Extremely Low Probability of Rupture Code". Efforts were continued to co-lead an international benchmark for piping PFM codes through the Organization for Economic Co-operation and Development, Nuclear Energy Agency, Committee on the Safety of Nuclear Installations. The NRC staff presented some initial benchmark result comparisons at the American Society of Mechanical Engineers 2022 Pressure Vessels & Piping Conference. Another effort was initiated to couple the xLPR code with machine learning models to investigate the impact of using these models to conduct sensitivity analyses more efficiently and effectively, which are recommended for PFM analyses in RG 1.245. Finally, the NRC staff explored using the xLPR code for developing loss of coolant accident frequency estimates. A public meeting was held with industry representatives to discuss potential research and regulatory applications in this area.

In the nuclear power plant vessel area, the development of the FAVPRO code under a modern software quality assurance program and using Agile development practices reached several milestones. First, the three FAVOR sub-programs were integrated into one program. Second, the I/O of the code was revamped to use the Java Script Object Notation (JSON) standard. Finally, the source code was fully modularized ahead of code parallelization. The software quality assurance and verification and validation testing have been enhanced thanks to automatic testing and automatic documentation creation, as well as an expanded test suite that provides better code coverage. It is anticipated that the first beta version of FAVPRO will be released to users in FY 2023.

To support increased use of PFM in nuclear applications in general, the NRC issued Regulatory Guide 1.245 and the accompanying technical basis NUREG/CR-7278 in January 2022. 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 RG-1.245 and NUREG/CR-7278: TLR-RES/DE/REB-2021-15 and RES/DE/REB-2021-16.

**FY 2023**
In the nuclear power plant piping integrity area, the NRC staff and EPRI developed xLPR v2.3. This version includes an optimized module for primary water stress corrosion cracking initiation, expanded software operating environment support, correction of user-reported problems, and various updates to enhance maintainability. It will be publicly released in the beginning of FY 2024. The NRC staff also continued to co-lead the international benchmark for piping PFM codes through the Organization for Economic Co-operation and Development, Nuclear Energy Agency, Committee on the Safety of Nuclear Installations. The NRC staff presented final benchmark result comparisons at the American Society of Mechanical Engineers 2023 Pressure Vessels & Piping Conference. Supplemental studies were published in Technical Letter Report TLR-RES/DE/REB-2023-06, "Assessment of the Performance of the Extremely Low Probability of Rupture Code in an International Benchmark with Insights from Advanced Finite Element Analysis." Additionally, results from coupling the xLPR code with machine learning models to conduct sensitivity analyses were published in TLR-RES/DE/REB-2022-13, ""Autonomous Researcher Feasibility Studies," and highlighted in a technical paper at the 2023 International Association for Probabilistic Safety Assessment and Management Topical Conference on Artificial Intelligence and Risk Analysis for Probabilistic Safety/Security Assessment and Management. Another public meeting was held with industry representatives to discuss potential research and regulatory applications using the xLPR code to generate loss of coolant accident frequency estimates. Further, the NRC staff used the xLPR code to support a risk-informed evaluation of international operating experience involving stress corrosion cracking in PWR emergency core cooling systems.

In the nuclear power plant vessel area, the development of the FAVPRO code was nearly completed by using Agile software development practices that include an enhanced software quality assurance program. A fully integrated beta version of FAVPRO was issued to users in August 2023, with a new modern user interface using the Java Script Object Notation (JSON) standard. The fracture mechanics computational engine was upgraded to use standard ASME correlations wherever possible, and the new ASTM-E900-21 material embrittlement model was added in anticipation of potential work related to SECY-22-0019 and to ASME Code changes. In addition, natural language programming and generative AI were used to begin the development of a visualization tool for FAVPRO output. The production release of FAVPRO is planned for early 2024 once parallelization of the probabilistic portion of the code is complete. FAVPRO constitutes a modern platform that can easily be adapted to support risk-informed regulatory decisionmaking related to Long Term Operations of the current LWR fleet.

 |  

## Implementing Lessons Learned from Fukushima

**Summary Description**

**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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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.

**FY 2018**

The NRC staff received 7 seismic PRA submittals and issued staff assessments of 3 submittals in 2018. More information on this risk-informed initiative can be found on the NRC's Seismic Reevaluations Web page.

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.

**FY 2019**

The NRC staff received 3 seismic PRA submittals and issued staff assessments for 4 submittals (one of the assessments was for a submittal received in FY 2018). The staff uses risk insights from the seismic PRA submittals to evaluate the impact of the site-specific reevaluated seismic hazard and determine whether further regulatory actions are warranted on a plant-specific basis. Several licensees have considered the risk insights from their seismic PRAs to identify and voluntarily undertake plant modifications for safety enhancement. For more information on this risk-informed initiative, please see Plant-Specific Japan Lessons-Learned Activities.

**FY 2020**

The NRC staff received seismic PRA submittals for 6 sites and issued staff assessments for 8 sites (2 assessments were for sites that submitted seismic PRAs in FY 2019). The staff uses risk insights from the seismic PRA submittals to evaluate the impact of the site-specific reevaluated seismic hazard and determine whether further regulatory actions are warranted on a plant-specific basis. The NRC staff completed all its seismic PRA and flooding reviews during FY 2020 using a risk-informed approach. Similarly, the NRC staff received external flooding integrated assessment (IA) submittals for 2 sites and issued staff assessments for 3 sites (1 staff assessment was for a site that submitted its IA in FY 2019). The staff uses information on the impact of flooding on key safety features, the available physical margin, and licensee actions to evaluate the impact of the site-specific reevaluated external flooding hazard and determine whether further regulatory actions are warranted on a plant-specific basis

Several licensees leveraged risk insights from their seismic PRAs and external flooding assessments to identify and voluntarily undertake plant modifications. The NRC staff's efforts, coupled with licensee identified modifications, have resulted in safety enhancements and an improved ability to cope against the reevaluated hazards.

**FY 2021**

No Update

**FY 2022**

The NRC staff has completed the regulatory actions undertaken after the accident at Fukushima Dai-ichi. All applicable licensees have completed the safety improvements associated with the orders for mitigating strategies, spent fuel pool instrumentation, and severe-accident-capable hardened containment vent systems (HCVSs). All applicable operating power reactors have reported compliance with these orders. The NRC has completed all the onsite inspections to verify licensees’ compliance with the orders for mitigating strategies, spent fuel pool instrumentation, and HCVSs. The latter order only applies to boiling-water reactors with Mark I or Mark II containment designs, for which there are 17 sites total.

Also, the NRC has completed its review of the seismic and flooding hazard information and determined that no additional regulatory action related to the seismic and flooding hazards are needed. There was one seismic evaluation associated with a site that had an approved due date deferral beyond its announced permanent shutdown date, and therefore, did not complete the evaluation.

**FY 2023**
No Update

 |  

## Accident Sequence Precursor (ASP) Program

For more information see Accident Sequence Precursor (ASP) Program web page.

 |  

## Probabilistic Flood Hazard Assessment (PFHA)

**Summary Description**

**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 since approximately 2014.

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).

**Previous Fiscal Years**

**Previous Fiscal Years**
**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.

**FY 2018**

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. Two new projects have been initiated via interagency agreement with Oak Ridge National Laboratory and the U.S. Geological Survey. Research projects address probabilistic modeling for precipitation, site-scale flooding, riverine flooding, and coastal flooding, as well as reliability of flood protection features and procedures. Cooperative research efforts with the Electric Power Research Institute (EPRI) have continued under the Flooding Research Addendum to an existing NRC-EPRI MOU. Several technical exchanges with EPRI were held in FY 2018, including a paleoflood hydrology workshop hosted by EPRI and the Tennessee Valley Authority (TVA) on February 21-22, 2018. A cooperative research agreement with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN) was completed and a NRC/IRSN technical exchange was held on March 26-27, 2018. The 3rd annual program review on the progress of PFHA research projects was held as a public meeting on December 4-5, 2017 at NRC headquarters. NRC staff also participated in an international workshop on riverine flooding organized by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA) on March 21-23, 2018.

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.

**FY 2019**

Progress on existing projects initiated via interagency agreements and cooperative research initiatives with other agencies has continued in FY 2019 and several projects have been completed. The technical-basis-phase of the PFHA research is now largely complete. This 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 next phase (pilot projects), focused on integrating the technical basis research into holistic multi-mechanism flooding hazard assessments, was initiated in FY 2019. This phase comprises three new projects: (1) Local Intense Precipitation Flooding PFHA Pilot Study; (2) Riverine Flooding PFHA Pilot Study; and (3) Coastal Flooding PFHA Pilot Study. Cooperative research efforts with the Electric Power Research Institute (EPRI) have continued under the Flooding Research Addendum to an existing NRC-EPRI MOU. The 4th Annual NRC PFHA Research Workshop, held as a public meeting on April 30-May 2, 2019, at NRC headquarters, reviewed progress on NRC-funded projects and updated participants on PFHA research at other agencies. NRC staff participated in a three-day technical exchange with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN) at IRSN Headquarters in Paris on September 23-25, 2019. 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.

**FY 2020**

Progress on existing projects initiated via interagency agreements and cooperative research initiatives with other agencies has continued in FY2020 and several projects have been completed. The first phase of the PFHA research (technical basis) is largely complete. This 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 holistic multi-mechanism flooding hazard assessments, initiated in FY2019, 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), initiated in FY2020, is at the scoping stage. Proceedings from the first four Annual PFHA Research Workshops was published as an NRC Research Information Letter (RIL-2020-01). The 5th Annual NRC PFHA Research Workshop, held as a public meeting on February 19-21, 2020 at NRC headquarters, reviewed progress on NRC-funded projects and updated participants on PFHA research at other agencies. NRC staff participated in a 3-day technical exchange with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN) at IRSN Headquarters in Paris on Sept 23-25, 2019. 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.

**FY 2021**

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.

**FY 2022**

No update

**FY 2023**
Progress on existing projects initiated via interagency agreements and cooperative research initiatives with other agencies has continued in FY2023. Phase 1 of the PFHA research (technical basis) is complete. Several reports were published in 2023 and several more are in press. This phase of the research 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. Phase 2 (pilot projects) is focused on integrating the technical basis research into demonstration multi-mechanism flooding hazard assessments. This phase comprises three projects: (1) Local Intense Precipitation Flooding (LIP) PFHA Pilot Study; (2) Riverine Flooding PFHA Pilot Study; and (3) Coastal Flooding PFHA Pilot Study. The LIP and riverine flooding pilot studies have been completed. The coastal flooding pilot study will be completed in early FY 2024. The third and final phase (guidance development) is in progress. The 8th Annual NRC PFHA Research Workshop held March 21-24, 2023 (Hybrid), reviewed progress on NRC-funded projects and updated participants on PFHA research at other agencies. Proceedings from the 6th and 7th Annual PFHA Research Workshops were published as an NRC Research Information Letters (RIL-2022-10 and RIL-2023-05). NRC staff participated in technical exchanges with the French Institute de Sûreté Nucléaire et de Radioprotection (IRSN). Cooperative research efforts with IRSN and the Electric Power Research Institute (EPRI) have continued under existing agreements. NRC staff participated in activities of the American Society of Mechanical Engineers Joint Committee for Nuclear Risk Management (ASME/JCNRM) External Flood Working Group. NRC staff helped organize and participated in an international workshop on the safety assessment of nuclear installations for combinations of external hazards sponsored by the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA).  NRC staff also participated in activities of the OECD/NEA Working Group on External Events.

 |  

## Risk Assessment of Operation Events (RASP Handbook)

**Summary Description**

**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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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 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.

The staff also revised the RASP Handbook volume on external events that provides methods and guidance on evaluating risk associated with external flooding and seismic. For external flooding, this update uses lessons learned from the analyses of approximately 10 "greater-than-green" findings relating to external flooding that resulted from inspections conducted after the events at Fukushima Dai-Ichi. The revised guidance provides references to methods and datasets along with discussion of common issues with external flooding assessments. For seismic, this revision incorporates updated site-specific seismic information based on licensees' recent seismic hazard reevaluations addressing Near-Term Task Force Recommendation 2.1. The revised guidance also provides references to methods that address key aspects required by the American Society of Mechanical Engineers/American Nuclear Society PRA Standard for a seismic PRA.

**FY 2018**

The staff published NUREG-2225, "Basis for the Treatment of Potential Common Cause Failure in the Significance Determination Process." It describes the basic assumptions and key principles for treating CCF of redundant components in SDP risk assessments when one or more of the redundant components are failed or functionally degraded due to a deficiency in licensee performance. The staff also prepared guidance on estimating the risk metric of Large Early Release Frequency (LERF) from a consequential steam generator tube rupture (C-SGTR) event.

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.

**FY 2023**
While the RASP handbook is still in use, no changes have been made to the RASP handbook since 2018. The latest versions of the RASP handbook are available to the public at the Reactor Oversight Process References web page.

 |  

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

**Summary Description**

**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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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.

**FY 2018**

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." The SAPHIRE development team released three new versions of the SAPHIRE software during FY 2018. A significant new feature that has been added to the code is the ability to load pre-defined groups of model changes (referred to as "change sets") into an event or condition assessment. This feature makes it easier for analysts to perform different what-if scenarios to assess the impacts of changes in modeling assumptions or reliability data input. Another important new code feature allows users to directly post their analyses to a centralized and secure web portal. This enhances the NRC staff's abilities for sharing information and collaboration, which is particularly helpful when staff are collaborating across the Regional Offices, Headquarters, and contractors at the Idaho National Laboratory.

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

**FY 2019**

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." The SAPHIRE development team released the latest version of SAPHIRE, version 8.2.0, in August 2019. A significant improvement that supports the use of this risk tool was the introduction of a cloud-based platform to access SAPHIRE and the SPAR models. This cloud-based platform is hosted on the Idaho National Laboratory Safety Portal. The use of the INL Safety Portal support future development of a cloud-based version of SAPHIRE that is anticipated to become available in the near future. These advances enhance the NRC staff's abilities for sharing information and collaboration, which is particularly helpful when staff are collaborating across the Regional Offices, Headquarters, and contractors at the Idaho National Laboratory. The SAPHIRE computer code is used to develop and run PRA models (e.g., SPAR models) for a variety of risk-informed regulatory applications.

**FY 2020**

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." The SAPHIRE development team released the latest version of SAPHIRE, version 8.2.2, in June 2020. One focus area for the team has been the development of a cloud-based solving platform. With PRA models increasing in scope and complexity, the SAPHIRE team is working to leverage the significant computing resources available on the computer cluster hosted by NRC's contractors at the Idaho National Laboratory. The latest release of SAPHIRE takes a step toward achieving that goal by including the capability to send the model logic to an external solving engine. Although additional development is needed before users can access the cloud-based solving platform, it marks an important development milestone. Beyond the cloud-based development, the SAPHIRE team continues to respond to users' requests for new features and address any identified code errors. The SAPHIRE team has also begun work to develop new features to streamline updates to the model database, which will allow for a more efficient process for keeping the reliability data up-to-date across the entire suite of SPAR models.

**FY 2021**

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.

**FY 2022**

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 2022, the SAPHIRE development team released two new versions of SAPHIRE, version 8.2.5 in November 2021 and version 8.2.6 in April 2022. New features include new pre-formatted results reports and improvements to capabilities for using multiple processors for performing analyses and generating reports. The SAPHIRE team also continues efforts to develop a cloud-based solving platform that will better support solving large and complex models. An initial version with a remote solving capability was released with SAPHIRE version 8.2.6 for testing by NRC users. The remote solving capability allows users to send analyses to be solved using servers hosted at the Idaho National Laboratory. Development of the remote solving capability is planned to continue in FY 2023 as part of the overall effort to move toward a cloud-based architecture for SAPHIRE. The SAPHIRE team will also continue to respond to users’ request for new features and address areas for improvement in the code.

**FY 2023**
In FY 2023, the SAPHIRE development team released two new versions of SAPHIRE, version 8.2.7 in January 2023 and version 8.2.8 in March 2023. 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 addition to bug fixes and stability and performance improvements, new features include an improved Remote Solver and Remote Update function, easier extraction of key data, better automation features, and a capability to generate multi-unit cut sets. The SAPHIRE team also continues efforts to develop SAPHIRE 9, a cloud-based solving platform that will better support solving large and complex models. Initial testing of this feature led to a number of improvements. The remote solving capability allows users to send analyses to be solved using servers hosted at the Idaho National Laboratory. Additional functionality was added in FY 2023, and a new file format and interface will be prototyped in FY 2024 as part of the overall effort to move toward a cloud-based architecture for SAPHIRE. The SAPHIRE team will also continue to respond to users’ request for new features and address areas for improvement in the code.

 |  

## Standardized Plant Analysis Risk Models (SPAR)

**Summary Description**

**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).

**Previous Fiscal Years**

**Previous Fiscal Years**
**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).

**FY 2018**

During FY 2018 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, about 80 percent of the SPAR models were modified. These modifications included: routine updates to reflect recent plant changes, modifications to the models to reflect new Westinghouse Gen 3 seals, and SPAR model modifications to incorporate FLEX modeling. For new reactor designs, the staff continued the efforts to collect information to start building the 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.

**FY 2019**

During FY 2019 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 226 SPAR model modifications were completed. These modifications included: routine updates to reflect recent plant changes, incorporation of new logic associated with external events, SPAR model modifications to incorporate FLEX modeling, and modifications to support Significance Determination Process (SDP) or Events and Conditions Assessment (ECA) analyses. For new reactor designs, the staff continued the initial process of gathering information to develop the 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.

**FY 2020**

During FY 2020 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 150 SPAR model formal modifications were completed. These modifications included: routine updates to reflect recent plant changes, 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 initiated the development of the 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.

**FY 2021**

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.

**FY 2022**

During FY 2022 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 significant number of 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, modifications to SPAR models to reflect the most recent plant operating data, and modifications to support the Significance Determination Process (SDP) or Events and Conditions Assessment (ECA) analyses.

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.

**FY 2023**
During FY 2023 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 significant number of SPAR model formal modifications were completed. These modifications included: routine updates to reflect recent plant changes (7 models were updated with the latest information from licensee PRA models), incorporation of new logic associated with external events, modifications to SPAR models to reflect the most recent plant operating data, and modifications to support the Significance Determination Process (SDP) or Events and Conditions Assessment (ECA) analyses.

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.

 |  

## Full-Scope Site Level 3 PRA

**Summary Description**

**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, a reference site was selected for the Level 3 PRA study. The reference site contains two four-loop Westinghouse PWRs with large dry containments. The Level 3 PRA project team is leveraging the existing and available information on the reference site and the corresponding 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.
              - MELCORE 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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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. Initial versions of reactor, at-power, Level 1 PRA models for internal fires and seismic events were completed in FY 2015, but they need to be significantly revised to incorporate more recent licensee-supplied information.

**FY 2016**

A PWROG-led ASME/ANS PRA Standard-based peer review of the reactor, at-power, internal event and internal flood Level 3 PRA was performed in October 2015. A substantial revision was completed for the reactor, at-power, Level 1 PRAs for internal events and internal floods. The reactor, at-power, Level 1 PRAs for internal fires and seismic events were significantly revised to incorporate more recent licensee-supplied information. 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 will be revised. Substantial progress was made on an initial reactor, low power and shutdown (LPSD), Level 1 PRA for internal events. 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 Units 1 and 2 of the reference plant.

**FY 2017**

The final report was completed for the revised reactor, at-power, Level 1 PRA for internal events. A substantial revision was completed for the reactor, at-power, Level 2 PRA for internal events and internal floods. Significant progress was made on a substantial revision to the reactor, at-power, Level 3 PRA for internal events and internal floods. Internal technical reviews were completed on the reactor, at-power, Level 1 PRAs for internal fires and seismic events. Substantial revisions were completed for the reactor, at-power, Level 1 PRA for high winds and the qualitative screening analyses for other hazards. The DCS PRA for all PRA levels and all hazards was revised. The initial reactor, (LPSD), Level 1 PRA for internal events was completed. 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).

**FY 2018**

Final reports were completed for the revised reactor, at-power, Level 1 PRAs for internal floods and high winds; revised reactor, at-power, Level 2 PRA for internal events and floods; and revised qualitative screening analyses for other hazards. The reactor, at-power, Level 1 PRA for internal fires was submitted to the Level 3 PRA Technical Advisory Group (TAG) for review. The reactor, LPSD, Level 1 PRA for internal events was also submitted to the Level 3 PRA TAG for review. Substantial work was completed on the reactor, at-power, Level 2 PRA for internal fires, seismic events, and high winds; the reactor, LPSD, Level 2 PRA for internal events; and the spent fuel pool Level 1 and Level 2 PRAs (all hazards).

**FY 2019**

The technical work on the Level 3 PRA Project is nearing completion, with a target date for release of the results in FY 2022. Internal reports on the reactor, at-power, Levels 1, 2, and 3 PRAs for internal events and floods; the reactor, at-power, Level 1 PRA for high winds; and the screening analysis of other hazards are essentially complete. The reactor, at-power, Level 1 PRAs for internal fires, seismic events, and low power and shutdown are nearing completion. Substantial work has been completed on the reactor, at-power, Level 2 PRAs for internal fires, seismic events, and high winds; the reactor, LPSD, Level 2 PRA for internal events; and the spent fuel pool Level 1 and Level 2 PRAs (all hazards). Work on the Level 3 PRAs for reactor, at-power internal fires, seismic events, and high winds continues, as does the work on the Level 1, 2, and 3 Dry Cask Storage PRA. Pilot activities have been completed to demonstrate the proposed approach for integrated site risk using a dual-unit model. The pilot activities included Level 1 and 2 PRAs for internal events and floods, Level 1 PRAs for internal fires and seismic events, and a Level 1 PRA for internal events with one unit operating and one unit shut down. With the pilot activities complete, work has been initiated on the reactor, at-power, Level 1, 2, and 3 dual-unit PRA models for internal events and floods.

**FY 2020**

The technical work on the Level 3 PRA Project is nearing completion, with a target date for release of the results in FY22. Internal reports on the reactor, at-power, Levels 1, 2, and 3 PRAs for internal events and floods; the reactor, at-power, Level 1 PRAs for internal fires, seismic events, and high winds; and the screening analysis of other hazards are complete. The reactor, at-power, Level 1 PRA low power and shutdown (internal events only) is nearing completion. Substantial work has been completed on the reactor, at-power, Level 2 and 3 PRAs for internal fires, seismic events, and high winds; the reactor, LPSD, Level 2 PRA for internal events; and 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. Pilot activities have been completed to demonstrate the proposed approach for integrated site risk using a dual-unit model. The pilot activities included Level 1 and 2 PRAs for internal events and floods, Level 1 PRAs for internal fires and seismic events, and a Level 1 PRA for internal events with one unit operating and one unit shut down. With the pilot activities complete, work has been initiated on the reactor, at-power, Level 1, 2, and 3 dual-unit PRA models for internal events and floods.

**FY 2021**

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.

**FY 2022**

The technical work on the Level 3 PRA Project is nearing completion, with a target date for public release of the results in FY23 through early FY24 (results for the reactor, at-power, Level 1, 2, and 3 PRAs for internal events and floods were released publicly in April 2022). Internal reports on the reactor, at-power, Levels 1, 2, and 3 PRAs for all hazards (i.e., internal events, internal floods, internal fires, seismic events, and high winds) and on the screening analysis of other hazards are essentially complete. The reactor, low power and shutdown (LPSD) Level 1 PRA for internal events is undergoing final management review, the Level 2 PRA is nearing completion, and work is continuing on the Level 3 PRA. The spent fuel pool Level 1 and Level 2 PRAs (all hazards) are also nearing completion. The technical work for the spent fuel pool Level 3 PRA is complete and work is continuing on its documentation. The dry cask storage combined Level 1, 2, and 3 PRA (all hazards) is undergoing final documentation. Work is continuing on the analysis of integrated site risk.

**FY 2023**
No Update

 |  

## Data Collection for Human Reliability Analysis (HRA)

**Summary Description**

**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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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.

**FY 2018**

The following are the accomplishments and ongoing tasks in FY 2018:
              - Established an agreement with Vogtle Unit 3 and Unit 4 sites to use the NRC's SACADA system for operator simulator training.
              - Hosted an international HRA Data workshop on March 15 and 16, 2018 at the NRC Headquarters featuring:	                 - 40 participants from 23 organizations of seven countries
                - Presentations from three NRC contractors with proposals for using SACADA data for human error probability estimations
                - Nine technical presentations from workshop participants
                - Workshop documentation via  2018 March 15&16 Human Reliability Analysis Data Workshop 
              - Presented recent SACADA research results at the SACADA Data technical session at the 14th Probabilistic Safety Assessment and Management (PSAM 14) conference (www.PSAM14.org), September 16-21, 2018.
              - Developing a revision to SACADA software to collect operator performance in simulator training, job performance measures, written exams, and actual events (with technical support provided by South Texas Project and Vogtle 3&4 instructors).
              - Continuing outreach to NRC licensees and the international nuclear industry on using SACADA for operator simulator training.

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.

**FY 2019**

The following are the accomplishments and ongoing tasks in FY 2019:
              - Renewed a memorandum of understanding with the Korea Atomic Energy Research Institute on HRA Data Collaboration, effective until November 2024.
              - Hosted an international HRA Data workshop on March 14 and 15, 2019 at the NRC Headquarters featuring:	                 - 25 participants from 15 organizations of four countries
                - Eleven technical presentations and a panel discussion
                - Workshop documentation via  Human Reliability Analysis Data Workshop in the NRC, March 14-15, 2019
                - Released five sets of SACADA data to the public. The data are accessible at:		                   - Data
                  - SACADA Data Release 2019 
              - Continuing outreach to NRC licensees and the international nuclear industry on using SACADA for operator simulator training.

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.

**FY 2020**

RES expanded SACADA's capability to collect operator performance in job performance measures (JPMs). RES continues outreach to nuclear power plants to use SACADA to collect JPM performance information. 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 has collaborated with domestic and international organizations to use simulator data to inform HRA.

**FY 2021**

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.

**FY 2022**

No update

**FY 2022**
The following highlight the HRA data activities:
              - RES initiated the Open Source SACADA (OSS) operation to promote the use of SACADA. OSS provides the SACADA source codes to the organizations who are interested in modifying SACADA for their specific purposes. Sandia National Laboratories (SNL) is the first OSS partnership. SNL plans modify SACADA for its blended (i.e., combination of cyber and physical) attack experiments. Note: SACADA (Scenario Authoring, Characterization and Debriefing Application) was developed by NRC to collect human reliability data in operator simulator training and job performance measures.
              - Provided SACADA tool to Japan Nuclear Regulatory Authority to inform JNRA’s decision on collecting human reliability data.
              - Initiated a project to use Idaho National Laboratory’s simulator facilities to collect human reliability data in digital instrumentation and control environment.
              - Finalizing the project with Pacific Northwest National Laboratory on reviewing  IDHEAS-DATA  (draft).
              - Analyzing data to develop guidance on specifying time uncertainty for IDHEAS-ECA (NUREG-2256).

 |  

## Human Reliability Analysis (HRA) Methods and Practices

**Summary Description**

**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)
              - Developing guidance for HRA dependency analysis

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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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.

NUREG-2114, "Cognitive Basis for HRA" was finalized and published.

**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 staff worked with the Electric Power Research Institute (EPRI) to develop an approach to perform HRA related to main control room abandonment in fire events and published  NUREG-1921, Supplement 1 , "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines: Qualitative Analysis for Main Control Room Abandonment Scenarios"

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" 

**FY 2018**

The following tasks were accomplished in FY 2018:
              - Tested IDHEAS At-Power Application method in collaboration with EPRI, including:	                 - An evaluation of whether the method guidance could be practically applied to produce consistent HRA results, thereby improving HRA quality for risk-informed applications
                - There were three pressurized water reactor (PWR) scenarios:		                   - Two scenarios were adapted from the U.S. Scenario 1 described a standard steam generator tube rupture (SGTR), and Scenario 2 described a total loss of feedwater (LOFW) with a misleading indicator of flow to the steam generators.
                  - The third scenario was developed from an actual event in which an electrical fire caused a reactor trip and subsequent loss of reactor coolant pump (RCP) seal injection and cooling.
                  - Overall, the results of the testing indicate that the IDHEAS At-Power method:			                     - provides a structured analysis framework and traceable quantification approach to HRA
                    - had some instances where the method was not applied consistently with method guidance
                    - exhibited the ability to capture a broad range of failure modes, contextual conditions, and influences on behavior associated with the difficult operator actions and complex scenarios in the study
                    - has the capability to translate qualitative findings into reasonable HEPs. 
              - Documentation of the results and lessons learned from the testing in NUREG-2199, Vol.3. (to be published)

The following are activities are in process in FY 2018:
              - Working with the Electric Power Research Institute to develop an approach to perform HRA related to main control room abandonment in fire events in order to improve fire PRA realism in risk-informed applications:	                 - Development of NUREG-1921, Supplement 2, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines: HRA Quantification for Main Control Room Abandonment Scenarios" which addresses:		                   - Operator actions taken before abandonment
                  - Operator decision to abandon for both loss of habitability and loss of control scenarios
                  - Operator actions taken after abandonment (including command and control contributions to operator failure) 
                - Presentation to the ACRS PRA Subcommittee and peer review
                - Final report documenting the guidance (to be published in 2019) 
              - In order to support risk-informed license amendment requests (LARs), notice of enforcement discretion (NOED) evaluations, event evaluations, and significance determination process (SDP) evaluations, the NRC staff needs an HRA method suitable for the unique FLEX context. This method will enable the staff to assign human error probabilities (HEPs) for such actions and incorporate the associated human failure events (HFEs) in the NRC's SPAR models. Research activities in 2018 include:	                 - Development of the IDHEAS application for event and condition analysis to support the NRC's inspection, licensing, and enforcement activities.
                - Development of a simplified HRA tool that can be used to quantify the HEPs of human actions in FLEX strategies
                - Performance of an expert elicitation to develop human error probabilities with the following results:		                   - Identification of the unique performance shaping factors associated with the use of FLEX equipment,
                  - evaluation of the contribution of the these performance shaping factors on HEPs, and
                  - development of HEPs associated with a few typical strategies for using FLEX equipment for added defense in depth during non-FLEX-designed accident scenarios and during FLEX-type scenarios (such as transportation, placement, connection, and local control of portable pumps and generators, refilling water storage tanks using alternate water sources, declaration of ELAP, and deep DC load shedding) 
                - Drafting a NUREG report documenting the expert elicitation process and results (to be published in FY2019. 

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.

**FY 2019**

The following tasks were accomplished in FY 2019:
              - Completed the development of the Draft IDHEAS General Methodology (IDHEAS-G):	                 - Developed a new function of IDHEAS-G to generalize and integrate human error data of various sources to calculate human error probabilities (HEPs)
                - Made the Draft IDHEAS-G report publicly available
                - Presented IDHEAS-G to ACRS Subcommittee on 9/18/2019 
              - Completed the development of the Draft IDHEAS Events and Conditions Assessment Method (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 an NPP control room—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.
                - The IDHEAS-ECA Software has been evaluated by NRR users, tested by the development team, and approved by the NRC Information Technology for installation on NRC computers.
                - A draft Users' Manual and training materials on IDHEAS-ECA were developed. The development team administrated two training session to NRC and industry PRA/HRA analysts that will use IDHEAS-ECA to perform HRA of human actions in FLEX Strategies. 
              - The following are activities in progress:	                 - Working with the industry to analyze human actions in FLEX Strategies using the IDHEAS-ECA method:		                   - In order to support risk-informed license amendment requests (LARs), notice of enforcement discretion (NOED) evaluations, event evaluations, and significance determination process (SDP) evaluations, the NRC staff needs an HRA method suitable for the FLEX context. IDHEAS-ECA enables the staff to assign human error probabilities (HEPs) for such actions and incorporate the associated human failure events (HFEs) in the NRC's SPAR models.
                  - In this joint project, the staff and industry developed representative FLEX scenarios for the testing of the IDHEAS-ECA method. 

The purposes of the HRA method efforts are to improve HRA methods and consistency among HRA practitioners. This will help improve HRA/PRA quality and provide a stronger basis for risk-informed regulatory actions.

**FY 2020**

The following tasks were accomplished in FY 2020:
              - Updated the Draft IDHEAS General Methodology (IDHEAS-G):	                 - Revised the report to address ACRS and public members' comments
                - Updated the new function of IDHEAS-G for generalizing and integrating human error data of various sources to calculate human error probabilities (HEPs)
                - Developed the conceptual IDHEAS dependency model
                - Presented IDHEAS-G to ACRS PRA Subcommittee on 9/23/2020 
              - Completed the development of IDHEAS Events and Conditions Assessment Method (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 an NPP control room—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.
                - The IDHEAS-ECA Software has been evaluated by NRR users, tested by the development team, and approved by the NRC's OCIO for installation on NRC computers.
                - IDHEAS-ECA method and guidance report was published as an NRC Research Information Letter RIL-2020-02.
                - IDHEAS-ECA Software Version 1.0 and 1.1 was released to the public. 
              - Completed HRA evaluation of representative human actions in FLEX Strategies using the IDHEAS-ECA Software:	                 - NRC staff and industry jointly developed two scenarios using FLEX equipment for evaluation, a FLEX-designed scenario in a severe seismic event and the other a non-FLEX-designed scenario of using FLEX equipment in an outage.
                - A group of NRC and industry HRA analysts estimated the HEPs of the selected human actions in the two FLEX-use scenarios using IDHEAS-ECA Software.
                - The NRC staff documented the scenarios and evaluation results in a draft (publicly available) report and presented the results to ACRS PRA Subcommittee on 09/23/2020. 

The following activities are on-going:
              - The NRC staff are collecting feedback and suggestions from the NRC and industry PRA/HRA practitioners on improving IDHEAS-ECA.
              - The NRC staff are planning on testing the IDHEAS dependency model and developing guidance for NRC HRA applications.

The purposes of the HRA method efforts are to improve HRA methods and consistency among HRA practitioners. This will help improve HRA/PRA quality and provide a stronger basis for risk-informed regulatory actions.

**FY 2021**

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.

**FY 2022**

Completed IDHEAS General Methodology (IDHEAS-G)
              - Published the final report (NUREG-2198)

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
              - Published the Final Report for IDHEAS-ECA (NUREG-2256) in November 2022

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.
              - RIL 2021-14 - Integrated Human Event Analysis System Dependency Analysis Guidance (IDHEAS-DEP) was published on  11/18/2021 .

The following activities are on-going:
              - The NRC staff is developing guidance for handling HRA data, estimating time, and for considering HRA recovery within IDHEAS.

**FY 2023**
Rolled out the IDHEAS Method for Events and Conditions Assessment (IDHEAS-ECA).
              - In 2019, the U.S. Nuclear Regulatory Commission (NRC) developed a new human reliability analysis (HRA) method—Integrated Human Event Analysis System for Event and Condition Assessment (IDHEAS-ECA). The method is based on state-of-the-art cognitive research and can improve the technical basis, analysis detail, and transparency of key assumptions for estimating the human error probabilities (HEPs) of human failure events (HFEs). 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 cognitive failure modes which model failure of human actions, and performance-influencing factor (PIF) structure, which models the context of a human failure event (HFE), are comprehensive and technology-neutral. The method covers all the PIFs in existing HRA methods and the factors reported in broad literature and in nuclear-specific human events. IDHEAS-ECA can be used in PRA applications; for example, SPAR models, 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.

Implementation of IDHEAS-ECA to human failure events (HFEs) in SPAR models.
              - The NRC has begun exploring the application of IDHEAS-ECA in various risk-informed activities. In addition, the NRC is assessing the transition from using SPAR-H HRA Method to IDHEAS-ECA in risk assessment of initiating events and/or degraded conditions, which are known as event and condition assessments (ECAs). However, HRA evaluations performed by the NRC staff for ECAs have historically been difficult and time consuming due to the limitations in accessing plant-specific data and documentation needed to fully implement HRA methods. As part of a pilot activity for increased use of IDHEAS-ECA, the NRC is building a knowledge base of application examples. An initial activity for building this knowledge base was to identify and evaluate several of the most risk significant HFEs that are commonly used in most standardized plant analysis risk (SPAR) models or that have been identified as risk significant during NRC-conducted ECAs. In addition, using IDHEAS-ECA to evaluate these HFEs in base SPAR models could improve the understanding of uncertainties associated with the use of industry-average (i.e., not plant-specific) HEPs currently in the SPAR models and specifying the contextual differences between accident sequences/cut sets.
              - The NRC staff analyzed an initial set of selected HFEs in SPAR models and documented the results in an evaluation report. The NRC staff is currently looking into case studies to understand the differences between currently used HEP values and those produced by IDHEAS-ECA.
              - The staff summarized the process and results of IDHEAS-ECA SPAR model applications in a technical paper and presented the paper at 2023 PSA conference, titled " Base Standardized Plant Analysis Risk (SPAR) Model Human Failure Event Application of Integrated Human Event Analysis System for Event and Condition Assessment (IDHEAS-ECA )". This paper describes the process of applying IDHEAS-ECA to an HFE with consideration of potential variability due to design differences and differing scenario contexts. In addition, this paper also discusses some general insights on the use of IDHEAS-ECA and illustrates the documentation generated during an IDHEAS-ECA analyses.

Application of IDHEAS-ECA in digital instrumentation and control (DI&C) environment.
              - Many operating U.S. plants are planning modernization projects to replace their analog instrumentation and control systems and human-system interfaces with new digital systems. Nuclear power plant control room modernization introduces digital instrumentation and control (DI&C) systems and digital human-system-interfaces to operators. These new systems expectedly will offer functions and capabilities that are vital for performance and plant safety. Although digital technology potentially can improve operational performance, there are challenges to using this technology. Moreover, introducing new technologies to control rooms would introduce new operator actions, change existing operator actions, and change the context of actions. The impact of such changes on operator performance and plant safety should be evaluated as new technologies are being introduced. This activity is to establish the basis that IDHEAS-ECA is applicable to evaluate operator actions in DI&C environment.
              - Because IDHEAS-ECA is cognition-centered and technology-neutral, in principle, the method can be used for HRA of human actions with DI&C technologies in advanced control rooms and DI&C modernization. The NRC staff analyzed IDHEAS-ECA applications in a DI&C environment and demonstrated its use with a set of human events in control room DI&C upgrades.
              - The work was summarized in a technical paper, presented at 2023 HMIT&NPIC Conference, titled as " Application of Human Reliability Analysis to DI&C Control Room Modernization ".
              - It describes the process and two case studies of applying the NRC’s human reliability method, the Integrated Human Event Analysis System for Event and Condition Analysis (IDHEAS-ECA), to the analysis of changing operator actions with the introduction of control room digital systems. The process with the case demonstration can be used along with human factors engineering process to systematically identifying and analyze potential risks associated with DI&C control room modernization. This paper also demonstrates the applicability of IDHEAS-ECA in human reliability analysis of DI&C working environment.

The following activities are on-going:
              - The NRC staff is developing guidance for handling HRA data, estimating time, and for considering HRA recovery within IDHEAS.

 

 |  

## National Fire Protection Association (NFPA) Standard 805

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.

Forty-six operating nuclear power reactors committed to transition to NFPA 805, and all have received license amendments. All reactor units have fully completed the transition. Transition completion is controlled by license condition and transition is considered completed when all implementation items and modifications required by NFPA 805 have been completed. Although there are no additional licensees scheduled to submit license amendment requests to transition to NFPA 805, the NRC staff has received 23 requests from NFPA 805 licensees requesting additional changes. Of these 23 requests, 21 have been completed, 1 has been withdrawn, and 1 remains to be completed.

As a result of NFPA 805, two NRC guidance documents have been updated. Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, Revision 2, was issued in May of 2021, and Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown, Revision 1, was issued in January of 2021.

 |  

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

**Summary Description**

**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.

As described below, the staff has identified several options including risk-informed options to address this generic safety issue. When using the risk-informed option to disposition GSI-191, the licensees use inputs from their PRA models which may including bounding analyses. Also, when using risk-informed options, licensees must address all five principles of risk-informed decision making in Regulatory Guide (RG) 1.174.

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).

**Previous Fiscal Years**

**Previous Fiscal Years**
**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.

**FY 2018**

The staff is currently expecting license amendment requests for eight additional units.

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."

**FY 2019**

GL 2004-02 has requested licensees to perform an evaluation of the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions considering potential post-LOCA debris impacts and, if appropriate, take additional actions to ensure system function.

On November 30, 2018, the Director of the Office of Nuclear Reactor Regulation (NRR) signed a letter to the PWR Owners Group Chairman (ADAMS Accession Number ML18311A297) stating that the NRC staff was reevaluating the closure path for GSI-191 and Generic Letter (GL) 2004-02, particularly related to in-vessel downstream debris effects. On June 13, 2019, the NRC staff issued a technical evaluation report of in-vessel downstream debris effects (ADAMS Accession Number ML19178A252), and subsequently, on July 23, 2019, the NRC closed out GSI-191 (ADAMS Package Accession Number ML19203A303) primarily because the NRC staff concluded that complex technical and safety impacts of downstream and chemical effects are well understood and that all safety significant issues have been adequately addressed by most licensees.

Although NRC closed out GSI-191, the staff continues to review licensee responses to GL 2004-02.

Licensees pursuing risk-informed approaches to respond to GL 2004-02 can also submit license amendment requests following the guidance in RG 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession Number ML17317A256).

**FY 2020**

The staff continues working with individual licensees on closure of GL 2004-02. Closure requires plants to address both strainer and in-vessel issues. This year the staff continued work to facilitate closure of the in-vessel portion. The staff developed guidance (ADAMS Accession No. ML19228A011v) for NRC staff review of in-vessel submittals. This in-vessel guidance risk-informs plant-specific resolution by describing several defined paths licensees may follow to resolve GL 2004-02. These are based on plant specific configurations, including the relative risk and available safety margin for that configuration. This, in turn affects the type and depth of supporting information needed for each. The staff also worked with the PWROG to develop related guidance for licensee submittals for in-vessel closure. About one-third of the plants closed both aspects of the issue using deterministic methods. Most of the plants completed their strainer evaluations using deterministic methods. Some of the plants that have not finalized their strainer evaluations plan to use a transition break size methodology that accounts for the risk associated with breaks of different sizes. A few plants plan to use fully risk-informed evaluations to close the strainer portion of GL 2004-02 using methods like the STP pilot discussed above. All of the remaining plants will use the risk-informed in-vessel guidance to close out the in-vessel portion of GL 2004-02. The staff will review submittals as they are submitted by each licensee.

**FY 2021**

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.

**FY 2022**

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 the Point Beach and Callaway license applications.

**FY 2023**
No Update

 |  

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

**Summary Description**

**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," allows licensees to use risk insights to extend the "completion times" by which an inoperable SSC controlled by technical specifications must be restored. The risk-informed completion time could be shorter than that required by technical specifications given actual plant conditions, or could be much longer, up to a "backstop" of 30 days. From a safety perspective, the program uses a real-time view of plant conditions to determine an appropriate time for key equipment to be out of service and focuses plant attention on issues of the highest safety significance. Licensees also benefit from the ability to schedule maintenance activities over a longer time period if appropriate. The plant-specific Probabilistic Risk Analysis (PRA) model is utilized to generate the risk metrics following a change in the plants configuration resulting in a quantifiable change in risk allowing for a flexible completion time for the Conditions in the Technical Specifications of the nuclear plant.

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. The associated Technical Specifications Task Force traveler (TSTF-505) to revise the STS was published in March 2012.

TSTF-505, Revision 2, was submitted July 2, 2018 (ADAMS Accession No. ML18183A493).  TSTF-505, Revision 2, was approved on November 21, 2018 (ADAMS Package Accession No. ML18269A041).  As of October 2021, the NRC has approved nine applications adopting TSTF-505, Revision 2.  Seven additional applications are currently being reviewed by NRC staff.

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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**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**

NRC suspended adoption of TSTF-505, detailing challenges with continuing to review amendments against it. Six LARs to implement a plant-specific 4b amendment were under review in FY 2017. Additionally, NRC met with industry representatives in several public meetings to discuss resolution of TSTF-505 issues, such as conditions which were accepted for use in program, and guidance for addressing other aspects of the application. The NRC staff, however, completed and issued the safety evaluation approving Vogtle Electric Power Generating Plants (Units 1& 2) to adopt Tech Spec 4b.

**FY 2018**

NRC issued the model safety evaluation for TSTF-505 alongside Revision 2 of TSTF-505. NRC received seven additional License Amendment Requests for Review. To enhance the efficiency of review, NRC staff conducted a number of on-site audits and disseminated lessons-learned at various public and industry led meetings. Based on information received from industry survey, a large majority of plants are planning to submit LARs and request staff approval to implement TSTF-505. NRC completed and issues several license amendments (Turkey Point, Calvert Cliffs).

**FY 2019**

The NRC has issued amendments for TSTF-505 (St. Lucie, Palo Verde, Farley), bringing the total of issued TSTF-505 and Initiative 4b amendments to 13.

**FY 2020**

The NRC continued receiving and reviewing License Amendment Applications for TSTF-505 Revision 2. The NRC issued six additional amendments, bringing the total of issued amendments to nineteen.

**FY 2021**

See Summary Description

**FY 2022**

The staff continues to work on the risk-informed technical specifications (RITS) initiatives to add a risk-informed component to the STS. The following summary highlights these activities:
              - Initiative 4b, "Risk-Informed Completion Times," (RICT) allows licensees to use risk insights to extend the "completion times" by which an inoperable SSC controlled by technical specifications must be restored. The RICT could be shorter than that required by technical specifications given actual plant conditions, or could be much longer, up to a "backstop" of 30 days. From a safety perspective, the program uses a real-time view of plant conditions to determine an appropriate time for key equipment to be out of service and focuses plant attention on issues of the highest safety significance. Licensees also benefit from the ability to schedule maintenance activities over a longer time period, if appropriate. The plant-specific Probabilistic Risk Analysis (PRA) model is utilized to generate the risk metrics following a change in the plants configuration resulting in a quantifiable change in risk allowing for a flexible completion time for the Conditions in the Technical Specifications of the nuclear plant.

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. The associated Technical Specifications Task Force traveler (TSTF-505) to revise the STS was published in March 2012.

TSTF-505, Revision 2, was submitted July 2, 2018. TSTF-505, Revision 2, was approved on November 21, 2018.  As of October 2022, the NRC has approved twenty three applications adopting the RICT program.  Six additional applications are currently being reviewed by NRC staff.

Initiative 5b, "Relocation of all Surveillance Requirement Frequencies out of TS," would permit SR frequencies to be determined in and relocated to a licensee-controlled TS program. PRA analysis is used to determine the risk impact of the intervals. A multi-disciplinary independent decisionmaking panel evaluates revised surveillance frequencies, based on operating experience, test history, manufacturers recommendations, codes and standards, and other factors, in conjunction with the risk insights from the PRA. Results and bases for the decision are also documented. Sensitivity studies are performed on important PRA parameters.  The staff has recently approved an amendment for the one remaining plant to adopt the program.

**FY 2023**
As of December 2023, the NRC has approved twenty nine applications adopting the RICT program. Seven additional applications are currently being reviewed by NRC staff.

 |  

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

**Summary Description**

**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. The provisions of 10 CFR 50.69 allow for the adjustment of the scope of SSCs subject to special treatment requirements based on an integrated and systematic risk-informed process. Special treatment refers to those requirements that provide increased assurance beyond normal industry practices that SSCs perform their design basis functions (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For SSCs categorized as low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed. 10 CFR 50.69 and its implementation relies on risk insight and metrics, such as importance measures, to categorize the safety significance of systems, structures, and components (SSCs). 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.

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).

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2015**

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

**FY 2016**

No new submittals seeking to implement 10 CFR 50.69 were received by the NRC in FY 2016. The NRC staff met with industry representatives in August 2016 to discuss future 50.69 LAR submittals and their content (ADAMS Accession Number ML16250A548).

**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, ML17265A020).

**FY 2018**

NRC issued the safety evaluation for Limerick Nuclear Generating Station. NRC received seven additional License Amendment Requests for Review. In FY 2018 the NRC licensing staff was in the process of reviewing 13 license amendment requests applicable to 24 additional operating units. To enhance the efficiency of review, NRC staff conducted a number of on-site audits and disseminated lessons-learned at various public and industry led meetings. Based on information received from industry survey, a large majority of plants are planning to submit LARs and request staff approval to implement 10 CFR 50.69. The NRC staff continued its meetings with licensees to discuss industry's proposed approach to implementing 10 CFR 50.69 for licensees that do not have seismic probabilistic assessments of seismic margins analysis (ADAMS Accession Numbers ML17305A242, ML18025B737, ML18143B668, and ML18250A193).

**FY 2019**

The NRC has issued nine amendment for 50.69 (ADAMS Accession Numbers ML18243A280, ML18264A092, ML18263A232, ML18289A378, ML19149A471, ML19192A012, ML19179A135, ML19176A421, and ML19205A289), bringing the total of issued 50.69 amendments to eleven. An additional five are under technical review.

The NRC staff made significant progress in reviewing new alternative approaches for addressing fire and seismic risk in the categorization process. For plants that did not develop a fire PRA the industry proposed to use the fire safe shutdown equipment list developed to demonstrate compliance with 10 CFR Part 50, Appendix R. One plant has received the amendment to use this alternative fire approach. With regards to the seismic risk, the industry proposed a three-tiered approach to implementing 10 CFR 50.69 for plants with low, medium and high seismic hazard/margin (EPRI Report 3002012988) for those licensees that do not have seismic probabilistic risk assessments of seismic margins analysis. One lead plant submitted for the Tier 1 of the approach and is under technical review. Another lead plant expected to submit for the Tier 2.

**FY 2020**

The NRC continued receiving and reviewing 50.69 License Amendment Applications. The NRC issued additional amendments for 50.69, bringing the total of issued amendments to seventeen. The NRC completed the review for one plant that applied the alternative seismic methodology proposed in EPRI Report 3002012988, for Tier 1 (low) seismic risk (ADAMS Accession Number ML19330D909). The NRC is in process of reviewing an application for a plant in Tier 2 (medium) seismic risk.

**FY 2022**

No Update

**FY 2023**
The NRC continued receiving and reviewing 50.69 License Amendment Applications. The NRC issued additional amendments for 50.69, bringing the total of issued amendments to twenty nine with eight under active review. The NRC approved multiple plants with the so-called Tier 1 (low seismic risk) and Tier 2 (medium seismic risk) alternative seismic methodology. These alternative seismic methodologies are supported by information in EPRI Report 3002017583.

 |  

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

**Summary Description**

**Summary Description**
NRC continues to explore how it can enhance existing processes and procedures to better focus licensee and NRC regulatory attention on design and operational issues commensurate with their importance to public health and safety, common defense, and security.  The NRC uses quantitative and qualitative risk insights that may be gleaned from assessing the risk triplet (What can go wrong? How likely is it? What are the consequences?) or when appropriate using Probabilistic Risk Analysis models. The following processes have been developed to better focus resources on matters important to safety: (1) The Very Low Safety Significance Issue Resolution (VLSSIR) process which establishes criteria for staff to use to stop screening and evaluating issues that have both very low safety significance and would take significant resources to examine whether the issue is within the licensing basis to determine whether a violation exists, and (2) The COM-106 Technical Assistant Request (TAR) process which provides a structured risk-informed process to inform decision making such as when it would be appropriate to stop screening and evaluating using the VLSSIR process, to expend staff resources to analyze the licensing basis to determine the state of regulatory compliance, or to conduct a backfit analysis.

While VLSSIR and TAR address issues during the inspection process, the Risk-Informed Process for Evaluations (RIPE) provides an approach to address compliance issues that are within the licensing basis. Staff developed RIPE to address low safety significant compliance issues using existing regulations under 10 Code of Federal Regulations (CFR) 50.12, "Exemptions, and 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit." RIPE is considered as an extension to LSSIR. 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 leverages 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.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2019**

Work to date has focused on revisions to inspection guidance, the COM-106 process, and licensing processes, as they pertain to the type of issues described above. Working groups were formed, external stakeholder feedback has been solicited via three public meetings, some process enhancements are being implemented, and some recommendations for additional changes are being made. As an example, the aforementioned COM-106 process is being revitalized to feature a streamlined 3- step graded approach commensurate with issue significance, with timelier responses. Further the enhancement is expected to include an integrated team approach to address issue screening, scoping, evaluation and early alignment in the issue lifecycle. This screening and evaluation include consideration of an issue's safety significance, inclusions of additional licensing basis expertise, consideration of a resolution if easily achievable, and consideration of whether there is another agency process that is more suitable to resolving the issue, all prior to entering in to an in-depth evaluation. These changes to COM-106, in concert with accompanying changes to inspection guidance, will promote resource expenditures commensurate with an issue's significance. Separately, additional work is being undertaken to scope process improvements that would have an analogous benefit for compliance issues.

**FY 2020**

Staff completed the update to NRR Office instruction COM-106, "Technical Assistance Request" in August 2020 (ADAMS Accession No. ML19176A098). The recommendations of the LSSIR working group were provided to the NRR Office director in a memorandum (ADAMS Accession No. ML19260G224), and subsequently endorsed in their entirety (ADAMS Accession No. ML20022A032). A major recommendation from the working group resulted in the revision of Inspection Manual Chapter (IMC) 0612, Appendix B, "Issue Screening," and IMC 0611, "Power Reactor Inspection Reports" to enable inspection efforts associated with very LSS issues, for which there is a lack of clarity regarding its licensing basis standing, to be discontinued early in the inspection process. The revision to IMC 0612B was issued on 12/12/2019 and the revision to IMC 0611 was issued on 1/7/2020.

In FY 2020, staff began an initiative to entitled Risk Informed Process for Evaluations (RIPE) to address low safety significant compliance issues using existing regulations under 10 Code of Federal Regulations (CFR) 50.12, "Exemptions, and 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit." RIPE is considered as an extension to LSSIR. 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.

**FY 2021**

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).

**FY 2022**

Very Low Safety Significance Issue Resolution (VLSSIR) Process.The staff revised IMC 0612 Appendix B, "Issue Screening Directions" in August 2022 to clarify that issues that would screen to Traditional Enforcement can be closed by using the VLSSIR process.  This revision allows inspection efforts associated with very LSS issues, for which there is a lack of clarity regarding its licensing basis standing, to be discontinued early in the inspection process even if the issue would screen as Traditional Enforcement.

Risk-informed Process for Exemptions (RIPE)On May 10, 2022, RIPE was expanded to allow its application to license amendments involving changes to the technical specifications (TSs). This expansion does not substantively change RIPE, and it continues to meet the same process constraints and attributes that were identified in the previous issuances, with the exception of excluding TSs. In FY 2022, the NRC successfully reviewed the first-of-a-kind application using RIPE. The review was completed ahead of the prescribed schedule and the resources expended were commensurate with the safety significance of the issue under review. This demonstrated that RIPE can be successfully implemented if the criteria established by the process are met and the licensees follow the published guidance.

**FY 2023**
No Update

 |  

## Guidance for Unattended Opening Evaluations

**Summary Description**

**Summary Description**
See FY 2019.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2019**

The NRC staff began interactions with external stakeholders on revising current guidance regarding the evaluation of unattended openings (i.e., those that intersect a security boundary at a facility, such as underground pathways) to incorporate risk information, notably the dimensions for openings that must be protected. The updated guidance will consider three-dimensional information, rather than the current guidance that is based on a two-dimensional opening. The staff is reviewing the revision to the industry guidance document (NEI 09-05) for acceptance.

**FY 2020**

NEI 09-05, (Guidance on the Protection of Unattended Openings that Intersect a Security Boundary, Supplement) revision is being evaluated by staff.

**FY 2021**

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 (SNL) 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.

**FY 2022**

Following a peer-review process, SNL intends to issue the final report in Q1 FY2023. The next steps for this topic will be determined based on the results of the report and subsequent industry/licensee submissions.

**FY 2023**
Based on the results of the SNL Unattended Opening (UAO) study that was reviewed by representatives from the Department of Energy and NRC staff, NEI made a determination not to pursue changes to existing NRC-endorsed guidance regarding UAOs. Based on the final SNL report, NEI and the NRC determined that the report conclusions and data collected do not support revising the existing regulatory reasonable assurance standard (i.e., greater than 96 square inches) for UAOs.

 |  

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

**Summary Description**

**Summary Description**
See FY 2019.

**Previous Fiscal Years**

**Previous Fiscal Years**
**FY 2019**

The NRC staff is reviewing a White Paper submitted by the Nuclear Energy Institute (NEI), "Implementing a 24-Month Frequency for Emergency Preparedness Program Reviews," September 2019. Prior to 2018, 10 CFR 50.54(t), which requires licensees to review all program elements of their emergency preparedness programs no less frequently than every 12 months. However, this regulation was changed to allow for a 24-month frequency with the monitoring of EP Performance Indicators (PI). This change was made prior to the development of the Reactor Oversight Process (ROP) and the subsequent EP Cornerstone PIs. The NEI white paper proposes to adopt the ROP PIs in support of the 50.54(t) 24-month review frequency. The staff is using a risk-informed approach to analyze the merits of this proposal. (Note that the current ROP EP Cornerstone PIs may need to be revised as a result of this risk-informed activity). The staff expects to complete its analysis in FY 2020.

**FY 2020**

The NRC staff reviewed the White Paper submitted by the Nuclear Energy Institute (NEI), "Implementing a 24-Month Frequency for Emergency Preparedness Program Reviews," September 2019 and conducted public meetings during FY 2020 Qs 1/2. NRC is planning on endorsing the White Paper in the upcoming revision to RG 1.101, which is expected to be in draft for issuance during FY2021. (Note that the current ROP EP Cornerstone Performance Indicators may need to be revised as a result of this risk-informed activity).

**FY 2021**

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.

**FY 2022**

No update

**FY 2023**
No update. However, staff anticipates providing updates for FY 2024.

 |  

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