Light-Water Reactors Sub-Arena

Objective

Implement risk-informed and performance-based activities to address the PRA elements of Title 10, Part 52, of the Code of Federal Regulations (10 CFR Part 52), and to increase the effectiveness and efficiency of the design certification, licensing, and oversight activities that the NRC staff conducts for new LWRs.

This objective has two main parts:

• First, this objective involves using the plant-specific PRA to implement risk-informed and performance-based programs. For example, the maintenance rule (10 CFR 50.65) will utilize the PRA to a great extent. Other examples include initiatives that a new reactor licensee may voluntarily pursue, such as risk-informed technical specification completion time, risk-informed in-service inspection, or special treatment under 10 CFR 50.69.
• Second, this objective involves using risk insights and PRA results to improve the NRC's effectiveness and efficiency in the licensing and oversight processes. For example, the staff will use risk insights, in conjunction with other considerations, to focus its review of a new reactor license application on those aspects that are important to risk. Other examples include developing risk-informed acceptance criteria for applications and adopting a risk-informed approach to sampling the inspection, testing, analysis, and acceptance criteria (ITAAC) to confirm the acceptability of the as-built plant.

Basis

The risk-informed and performance-based activities (listed below) for this sub-arena satisfy the following screening criteria:

• The stated objective will help to improve the effectiveness and efficiency of the NRC's regulatory process, while increasing nuclear plant safety and reducing unnecessary regulatory burden.
• The bases for developing a risk-informed and performance-based regulatory structure for licensing and oversight of new LWRs are articulated in several Commission documents, policy statements, and processes (including the 10 CFR Part 52 rulemaking).
• Goals and activities to meet the objective for this sub-arena will be performance-based, to the extent that they meet the following four criteria:
  1. measurable parameters to monitor performance
  2. objective criteria to assess performance
  3. flexibility to allow licensees to determine how to meet the performance criteria
  4. no immediate safety concern as a result of failure to meet the performance criteria

An applicant for a combined license (COL) for a new LWR is required to perform a PRA. The NRC staff expects such PRAs to be used for the following purposes:

• Identify risk-informed safety insights.
• Demonstrate how risk compares to the Commission's goals.
• Assess the balance between accident prevention and mitigation.
• Identify and address vulnerabilities, reduce risk contributors, and select among design alternatives during the design phase.
• Demonstrate that the plant design represents a reduction in risk (compared to existing operating plants).
• Demonstrate that the design addresses the requirements in 10 CFR 50.34(f), as they relate to Three Mile Island (TMI).

PRA results and insights are used to support the following programs (among others):

• Regulatory Treatment of Non-Safety Systems (RTNSS)
• Inspection, test, analysis, and acceptance criteria (ITAAC)
• Reliability Assurance Program (RAP)
• Future aspects of regulatory oversight, technical specifications, the maintenance rule (10 CFR 50.65), and others

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:

• Ensure (during the design certification phase) that the applicant used risk-informed safety insights to select among alternative features, operational strategies, and design options to reduce or eliminate the significant risk contributors of existing operating plants.
• Ensure that the risk associated with the design compares favorably with the Commission's goals of less than 1E-04/year for core damage frequency (CDF) and less than 1E-06/year for large release frequency (LRF).
• Using the results and insights from the PRA, ensure that the COL applicant supported the RTNSS process, including the identification of structures, systems, and components (SSCs).
• Using the results and insights from the PRA, ensure that the COL holder supported regulatory oversight processes, as well as programs associated with plant operations (such as technical specifications, reliability assurance, human factors, and maintenance rule implementation).
• Using the results and insights from the PRA, ensure that the applicant identified and supported the development of specifications and performance objectives for plant design, construction, inspection, and operation (such as the ITAAC, RAP, technical specifications, and COL action items and interface requirements).

Summary Description

In response to the staff requirements memorandum (SRM) on SECY-12-0081, "Risk-informed Regulatory Framework for New Reactors," the staff submitted SECY-13-0137, "Recommendations for Risk-Informing the Reactor Oversight Process (ROP) for New Reactors." In that SECY paper the staff recommended the development of an integrated risk-informed approach for evaluating the safety significance of inspection findings for new reactor designs. In its SRM on SECY-13-0137, the Commission approved the staff's recommendation to develop appropriate performance indicators (PIs) and thresholds for new reactors. The Commission requested that the staff develop, with appropriate stakeholder input, the necessary updates to the PIs, including any new PIs or changes to thresholds, and submit them to the Commission for approval before power operation for the first new reactor units.

The Commission disapproved the staff's recommendation to develop an integrated risk-informed approach for evaluating the safety significance of inspection findings for new reactor designs. The Commission directed the staff to enhance the Significance Determination Process (SDP) by developing a structured qualitative assessment for events or conditions that are not evaluated in the supporting plant risk models, such as those conditions that might arise with passive safety systems, digital instrumentation and control (I&C), and human performance issues. The Commission requested that the staff submit a paper to the Commission with its proposed approach for any revisions to the SDP for new reactors at least 1 year before the scheduled implementation of any changes to the Reactor Oversight Program (ROP).

Previous Fiscal Years

FY 2015

The staff continued to work on the Commission's directions from the SRM on SECY-13-0137. The staff worked with stakeholders and the public to develop appropriate PIs and enhance the SDP. In May 2015, the staff discussed its approach and plans for responding to the SRM on SECY-13-0137 with stakeholders during a ROP working group public meeting. Another public meeting with stakeholders was held in September 2015 to discuss updates on the staff's activities and to obtain stakeholder feedback.

FY 2016

The staff continued to hold meetings internally and with stakeholders to develop documents related to the ROP program for new reactors. A draft white paper was issued on September 7, 2016.

FY 2017

The staff continued to hold meetings internally and with stakeholders. The staff drafted a response to the commission direction in SRM-SECY-13-0137 which has been reviewed by internal and external stakeholders including the ACRS. The draft SECY is planned to be submitted to the Commission by December 2017.

FY 2018

On September 12, 2018, the staff submitted SECY 18-0091, "Recommendations for Modifying the Reactor Oversight Process for New Large Light Water Reactors with Passive Safety Systems Such as the AP1000 (Generation III+ Designs)," in response to the Commission direction in the SRM to SECY 13-0137. In this paper, the staff has provided to the Commission its proposed approach to modifying the SDP for new reactors, specifically the AP1000. The staff intends to modify Inspection Manual Chapter (IMC) 0609, Appendix A, "The Significance Determination Process for Findings At-Power," Appendix G, "Shutdown Operations Significance Determination Process," Appendix H, "Containment Integrity Significance Determination Process," and Appendix M, "Significance Determination Process Using Qualitative Criteria." The staff intends to have the revisions to those guidance documents completed by November 2019.

In addition to the planned changes to the SDP, the staff provided to the Commission options for revising the Performance Indicators (PIs) for the AP1000. Staff analysis concluded that the five Mitigating Systems Performance Index (MSPI) PIs used for the currently operating fleet will not be valid for monitoring performance for the AP1000 design. In the paper, the staff recommended for Commission approval elimination of the MSPI indicators for the AP1000, with no new PIs to replace them. The baseline inspection program will provide additional guidance to compensate for the elimination of the MSPI indicators.

The paper also provided an analysis of the baseline inspection program for the AP1000, and determined that most inspection procedures will need to be modified to adjust sample sizes and resource estimates because of the fewer active components and lower overall risk in the design. The inspection procedures will also be modified to provide specific guidance unique to the AP1000 design.

FY 2019

In FY 2019, the staff completed a review of all inspection procedures for AP1000. The staff is planning to revise required sample sizes in many of those inspections based on far fewer safety-related structures, systems, and components, as well as the reduced risk profile of the new design. The staff is also adding guidance for inspectors to consider systems classified as Regulatory Treatment of Non-Safety Systems (RTNSS) for inspection samples because of their importance to defense-in-depth.

The staff continued updating Inspection Manual Chapter (IMC) 0609, Appendix A, "The Significance Determination Process for Findings At-Power," Appendix G, "Shutdown Operations Significance Determination Process," and Appendix H, "Containment Integrity Significance Determination Process," to incorporate procedure enhancements associated with AP1000. Since the changes identified to incorporate AP1000 are not substantial, the IMC procedure revisions are merged with other updates as part of the routine procedure change process. Appendix M, "Significance Determination Process Using Qualitative Criteria," was updated in January 2019. During its review, the staff determined that no changes were necessary to Appendix M to accommodate AP1000.

FY 2020

The staff completed updating Inspection Manual Chapter (IMC) 0609, Appendix A, "The Significance Determination Process for Findings At-Power," Appendix G, "Shutdown Operations Significance Determination Process," and Appendix H, "Containment Integrity Significance Determination Process," to incorporate procedure enhancements associated with AP1000. Since the changes identified to incorporate AP1000 were not substantial, the IMC procedure revisions were merged with other updates as part of the routine procedure change process. Appendix M, "Significance Determination Process Using Qualitative Criteria," was updated in January 2019. During its review, the staff determined that no changes were necessary to Appendix M to accommodate AP1000.

FY 2021

No Update

FY 2022

The NRC staff completed the update of baseline inspection procedures to incorporate unique sample requirements, to applicable procedures, that considers the risk profile of the AP1000 Units. In SECY-20-0050, the staff proposed the current inspection sample sizes (minimum, nominal, and maximum) for Units 1 and 2 as well as the proposed sample sizes for Units 3 and 4, and revised sample sizes for those Baseline Inspection Procedures (BIPs) for which the Vogtle Electric Generating Plant (VEGP) will be treated as a four-unit site. The staff also proposed the approach to inspecting the regulatory treatment of non-safety system systems (RTNSS) of structures and components within the Reactor Oversight Process.

For common sitewide program BIPs, the staff identified that many BIPs are not dependent on system design or numbers of components. If the applicable organizational structure is common to the four units (e.g., site security is the same for all units), the staff intends to treat the facility as a single four-unit site. For these BIPs, the staff determined that the current sample range was sufficient with the expectation that inspections will be performed at the maximum sample range and resource estimate at the VEGP.