• CURRENT RISK-INFORMED ACTIVITIES
  • Operating Reactors
  • New Light-Water Reactors
  • Advanced Reactors
  • Research and Test Reactors
  • Fuel Cycle
  • Byproduct Materials
  • Spent Fuel Storage and Transportation
  • Low-Level Waste and Decommissioning

Spotlight

FY 2018

FY 2018

Upon Commission approval, the NRC will publish the proposed rule in the Federal Register for public comment and hold a public meeting.

FY 2018

FY 2018

In FY2019, the staff intends to take a number of initiatives on this topic. As directed in SRM-SECY-17-0100 (ML18283A072), the staff will develop and implement a revised Force-on-Force inspection process, using risk-informed insights during the evaluation of internal and external stakeholder recommendations for proposed improvements. In the process, the staff will evaluate alternatives for the FoF program mock adversary force. Finally, the staff anticipates endorsing an industry-initiated FoF exercise controller guidance document.

FY 2018

FY 2018

This is anticipated to be accomplished through approval of an industry proposal.

FY 2018

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 2018

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 2018

FY 2018

In FY2019, 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.

Previous Fiscal Years

FY 2017

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

FY 2018

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. The updated Surry UA will be published in a NUREG/CR report likely in FY 2019. Because the three UA NUREG/CR reports for Peach Bottom, Surry, and Sequoyah are very long, technical, and detailed, the staff is planning to produce a separate SOARCA UA summary report which serves as a more useful and practical reference document for identifying and explaining the most important accident progression and consequence analysis insights across the various analyses.

FY 2018

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.

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

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.

Previous Fiscal Years

FY 2015

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

FY 2016

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

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

FY 2017

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

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

FY 2018

FY 2018

In FY 2017, the ASP Program published the "U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program 2017 Annual Report," which summarizes the results of ASP analyses for events reported in CY 2017. Eleven events were determined to be precursors. Of these 11 precursors, 3 precursors utilized SDP results in accordance with RIS 2006-24 and the remaining 8 precursors were identified via independent ASP analyses. Six of the events identified by ASP analyses had a CCDP or ΔCDP greater than or equal to 1x10^-5.

The NRC completed its internal evaluation of the ASP Program in February 2018. The ASP Program has implemented the recommendations made by NRC management. Additional information on the internal ASP Program review, including conclusions and recommendations, is available.

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

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

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

Previous Fiscal Years

FY 2015

In September 2015, a working group with members from NRR, the Regions, OGC, and OE was formed, and began evaluating the feasibility of the proposed approach, including verifying the legality of the approach determining how the risk significance would be evaluated, and gaging the industry's interest in participating in the process once developed. The working group also looked at the process for implementing this new approach. One implementation method that was considered was modifying the Notice of Enforcement Discretion (NOED) process to provide a process for addressing for low risk, low safety significance design compliance issues in a risk-informed manner.

FY 2016

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

FY 2017 - FY 2018

FY 2017 - FY 2018

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

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

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

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

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.

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

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.

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

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.

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

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.

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

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).

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 NRC's ADAMS Accession number ML18079A156
• 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.

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

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.

Previous Fiscal Years

FY 2015

A draft report is being finalized to document the research results from this study. It is expected that the report will be issued for public review and comment in late calendar year 2015 and finalized in 2016. This work was presented to the ACRS Metallurgy and Reactor Fuels Subcommittee on April 7, 2015. A draft version of the report was provided to the ACRS.

FY 2016

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

FY 2017

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

FY 2018

FY 2018

The C-SGTR project, based on the work request by NRR, has been completed in 2018.

The following main deliverables were issued:

• NUREG-2195
• RASP Manual Volume 5 on LERF, including C-SGTR. (draft, not publicly available).
• C-SGTR Calculator

A presentation on the subject and on the project was given to NRC staff on September 5, 2018.

A project closure letter from RES to NRR is to be issued in the first quarter of 2019.

Previous Fiscal Years

FY 2015

The NRC staff issued six non-pilot NFPA 805 license amendments with three more expected to be completed by the end of the year. Thirteen LARs are currently under review. Additional FY 2015 information is available.

FY 2016

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

FY 2017

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

FY 2018

FY 2018

The NRC staff issued one non-pilot license amendment and received one additional license amendment request. Two requests are currently under review.

Risk-Informed Licensing Reviews. NFPA 805 is a performance-based standard, endorsed via 10 CFR 50.48(c) that critically depends on risk information in the form of Fire PRA to enable licensees to transition from existing "deterministic" fire protection programs to ones that are "risk-informed, performance-based." Fire PRA is an integral part of the new licensing basis, and includes both quantitative evaluations of base risk and changes to base risk in accordance with RG 1.174 guidelines as well as supporting qualitative considerations, such as traditional defense in depth and safety margin, also as per RG 1.174.

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

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."

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

In late 2016 TSTF-505 was suspended pending updates required to address technical issues not previously identified. These issues are still being addressed. Although work is nearing completion, TSTF-505 remained suspended. The NRC staff, however, completed and issued the safety evaluation approving Vogtle Electric Power Generating Plants (Units 1 & 2) to adopt Tech Spec 4b (ADAMS Accession No. ML15127A669).

FY 2018

FY 2018

The NRC staff is reviewing four additional license amendment requests to adopt Tech Spec 4b, applicable to 11 operating units. Staff is also planning to lift its suspension on TSTF-505 since staff and other stakeholders' comments have been addressed to staff's satisfaction. Staff expects to receive a number of additional license amendment requests once the suspension on TSTF-505 is lifted.

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

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

FY 2018

NRC issued the safety evaluation for Limerick Nuclear Generating Station (ADAMS Accession Number ML18165A162). 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 Number ML17305A242, ML18025B737, ML18143B668 and ML18250A193).

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