ML25216A236
| ML25216A236 | |
| Person / Time | |
|---|---|
| Issue date: | 08/04/2025 |
| From: | James Drake Division of Operating Reactor Licensing |
| To: | |
| Shared Package | |
| ML25216A235 | List: |
| References | |
| Download: ML25216A236 (1) | |
Text
Jason Drake, NRR/DORL Project Manager NRC Experience Reviewing RIPE Submittals
2 Agenda
- RIPE reviews completed to date
- Expanded Example Explanation from Technical Reviewer Perspective (Submittal B, Ming Li)
- Lessons Learned
- RIPE Best Practices
- NRC Conceptual Changes
- Proposed GRIP Initiative 2
3 RIPE Reviews to Date
- Submittal A - [First RIPE application] January 14, 2022, Arizona Public Service Company submitted a request for an exemption from certain requirements in the 10 CFR 50.62(c)(1) regulation for the Palo Verde Nuclear Generating Station (PVNGS) (ML22014A415).
Notes:
- Good example of meeting the baseline criteria for RIPE and providing a comprehensive supporting package.
- Effective use of the pre-submittal meeting phase with effective licensee response to NRC input.
4 RIPE Reviews to Date
- Submittal B - February 2024, Duke Energy submitted a request for a limited exemption from the requirements of 10 CFR 50.55a(h)(2), Protection systems, requiring protection systems meet the requirements of IEEE Standard 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, for Shearon Harris (ML24037A284).
Notes:
- Several internal NRC practices and guidelines were reevaluated following this review.
- Highlighted the importance of receiving complete answers from licensees for the adverse impact screening questions (e.g. proper capture of IDP analysis).
5 RIPE Reviews to Date
- Submittal C - March 2024, Arizona Public Service Company submitted a license amendment request for Palo Verde to extend the TS sections 3.5.1 and 3.5.2, Condition B, Action Time from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to approximately 10 days regarding the safety injection tank TS action statement completion times (ML24068A252).
Notes:
- Another good example of meeting the baseline criteria for RIPE and providing a comprehensive supporting package.
6 RIPE Reviews to Date
- Submittal D - April 2024, Southern Nuclear Operating Company (SNC) submitted a license amendment request to revise the operating license, Appendix A, Technical Specification (TS) 3.6.5, Containment Air Temperature, Actions upon exceeding the containment average air temperature limit and remove an expired Limiting Conditions for Operation Note for Joseph M.
Farley Nuclear Plant (FNP), Units 1 and 2 (ML24110A126).
Notes:
- Serves as an example of the screening responses not meeting the standard or expectation from staff.
- The vulnerability is the subjective allowance in licensee response.
7 RIPE Submittal B Expanded Harris Discussion
- Turbine Trip System (TTS) is non-safety. Two trains of TTS are electrically connected.
- Class 1E RTS cables that trip turbine are terminated in non-Class 1E cabinet. This violates the IEEE 384 independence requirements, and thus IEEE 279 Clause 4.6.
- Although this cable is classified Class 1E, it is isolated from RTS system via the relay.
- Worst case short hazards could disable automatic turbine trip functions but will not impact safety RTS functions.
- The licensee confirms the manual turbine trip could serve as the diverse means to the automatic turbine trip functions.
8 Best Practices Identified from First RIPE Review
- Consider requesting a pre-submittal meeting before submitting a request using RIPE. Staff recommends licensees address the following during the pre-submittal meeting:
- A description of the issue and whether compliance with any regulations will be impacted
- How the screening questions will be answered
- How risk was calculated, including a description of any surrogates used, and the risk results
- Ensure application is consistent with 50.90 and 50.12
- Ensure application is consistent with RG1.174
- Consider providing NRC staff access to the integrated decisionmaking panel (IDP) report on a secure portal when the RIPE request is submitted to support efficient staff review.
9 Lessons Learned
- Provide NRC staff just-in-time training before pre-submittal meetings and when a submittal is received.
- Basis for supporting IDP decision in the initial screening needs to be more robust (appropriately documented).
- NRC needs to clarify expectations to licensees that worst case analysis needs to be defined up front.
- Audit was needed to overcome technical objection for Harris
- NRC use of the audit process for the Harris application provided needed flexibility to continue the review.
10 RIPE Best Practices
- Clarify expectations to licensees during the pre-submittal phase on what is needed for the RIPE review.
- IDP supporting analysis for screening determination
- No impact/worst case analysis to be provided in application
- Maintain the option to utilize the regulatory audit process to provide flexibility and efficiency in the review.
- Ensure NRC and licensee outlook is consistent for what is included in the RIPE package.
11 NRC Conceptual Changes
- Determining safety impact using quantitative analysis:
- The NRC is considering giving allowance for the safety impact of the issue to be considered minimal without providing a PRA result if the issue was screened out of inclusion in the licensees PRA model.
- Safety impact:
- The NRC is evaluating if the change in CDF and LERF can be considered minimal when a safety impact issue can be assessed using the PRA model, but the PRA success criteria continue to be met such that there is no change in calculated CDF and LERF.
12 NRC Conceptual Changes
- Documentation:
- The NRC is assessing clarification statements to more accurately define to licensees what the expectations are for the RIPE submittal package
- IDP supporting analysis for screening determination
- No impact/worst case analysis to be provided in application
- Regulatory Audits:
- The NRC concluded that efficiency and flexibility was gained utilizing the audit process in the Harris review and is evaluating the best methodology to include this as an ongoing option for future RIPE submittals.
13 NRC Conceptual Changes
- Staff Efficiencies:
- Institute training for staff on RIPE process when new applications are received
- Conceptualized standardized templates and project tracking models
- Evaluated the current NTO process and identified opportunities for improvement
- Consider editorial clarifications to ensure justification is provided by licensees for screening the issue out of the PRA, if applicable.
14 Questions?
Development of a Generic Risk-Informed Process (GRIP)
Michelle Kichline Senior Reliability and Risk Analyst Office of Nuclear Reactor Regulation (NRR)
Division of Risk Assessment (DRA)
16 Agenda
- Purpose of GRIP
- GRIP Research Assistance Request
- Process for Identifying the GRIP Safety Significance Criteria
- Proposed Safety Significance Criteria
- Path Forward
17 Purpose of GRIP Purpose
- Develop criteria for a generic risk informed process (GRIP) that can be used to determine when a licensing action that is not within the scope of a traditional PRA model should be considered to have a very low/minimal safety significance.
- These licensing actions would not currently qualify for a streamlined NRC-review under RIPE because they do not meet the quantitative RIPE criteria.
18 GRIP Research Assistance Request GRIP is planned in 2 phases
- Phase 1
- Identify criteria that can be used to expand RIPE as-is to address issues that may not be represented in a plant-specific PRA.
- Phase 2
- Collaborate with EPRI to develop new generic risk-informed criteria to determine when a licensing action that is not within the scope of a traditional PRA model should be considered very low/minimal safety significant.
- Use these criteria to expand RIPE or propose a new RIPE-like process.
19 Identifying the GRIP Criteria
- Assembled a team of NRC and EPRI participants.
- The NRC and EPRI teams each reviewed a variety of documents that discuss risk-informed decision making, including:
- International Standards (IAEA/INSAG)
- EPRI Reports
- Regulatory Guides and NUREGs
- NEI Documents (50.59, 50.69, RICT, SFCP)
- NUMARC Guidelines (Maintenance Rule, Shutdown Risk)
20 Proposed Safety Significance Criteria The teams gathered a list of potential criteria that could be useful in determining the safety significance of an issue and refined them into the following topics:
- 1. Impact on Initiating Events
- 2. Impact on Mitigations of Events
- 3. Impact on Dose Consequences
- 4. Impact on Defense-in-Depth (DID)
- 5. Impact on Safety Margin 20
21 Applicability
- The GRIP criteria are intended to be used to determine if a license amendment or exemption request submitted under 10 CFR Part 50.12 or 50.90 has a very low/minimal safety significance, without using a PRA.
- GRIP is not intended for use with license amendments or exemptions in the following categories:
- Security
- Radiation Protection Exposure Limits under 10 CFR Part 20
- Administrative Changes
22 Criterion 1: Initiating Events Does the proposed change result in a more than minimal increase in the likelihood of an event that could challenge normal plant operations during any mode of operation (e.g., at-power, shutdown)?
- Does the proposed change create a possibility for an accident of a new or different type?
- Does the proposed change adversely affect other units at a multiple unit site?
- Does the proposed change increase the likelihood of failure of an SSC that could cause an initiating event?
- Does the proposed change impact the ability to prevent an initiating event due to internal or external hazards (e.g., fire, internal flooding, and other hazards)?
- Does the proposed change impact the plant's ability to prevent failures of SSCs that could cause or worsen an initiating event (e.g., removal of a flood or fire barrier)?
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23 Criterion 2: Mitigation of Events Does the proposed change result in a more than minimal increase in the likelihood of failure to mitigate challenges to normal or emergency plant operation during any mode of operation?
- Does the proposed change adversely impact the plants ability to restore or maintain any of the key safety functions (e.g., decay heat removal, inventory control, power availability, reactivity control, containment) in any mode of operation?
- Does the proposed change increase the probability that a non-safety related SSC could prevent a safety related SSC from performing its safety function?
- Does the proposed change create the possibility of a malfunction with a different result than previously evaluated?
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24 Criterion 2: Mitigation of Events (continued)
Additional Considerations
- Does the proposed change increase the likelihood of human error or introduce new credible human failure events?
- Does the proposed change increase the failure probability or unavailability of SSCs used to implement abnormal or emergency operating procedures?
- Does the proposed change adversely impact the ability to detect SSC failures?
- Does the proposed change adversely impact the ability to detect or respond to an abnormal operating condition, design basis accident, anticipated operational occurrence, or other regulated event?
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25 Criterion 3: Dose Consequences Does the proposed change result in more than a minimal increase in the dose consequences of an accident?
- Does the proposed change impact the plants ability to prevent or mitigate offsite exposures?
- Does the proposed change impact the inputs or assumptions used in any design basis accident dose calculations?
- Does the proposed change deviate from NRC-approved evaluation methodologies used for dose calculations?
- Does the proposed change increase the probability of a release after a core damage event?
- Does the proposed change adversely impact barrier integrity, making a release more likely?
- Does the proposed change impact the plants ability to mitigate the release of radioactive materials or their radiological consequences, including considerations for release size and timing, etc.?
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26 Criterion 4: Defense-in-Depth (DID)
Does the proposed change result in more than a minimal decrease in DID capability?
- Does the proposed change result in a decrease in redundancy, independence, or diversity of any design features or system functions?
- Does the proposed change decrease the capability of any fission product barrier in any mode of operation?
- Does the proposed change increase the likelihood of common-cause failures?
- Does the proposed change create an overreliance on programmatic activities or other compensatory measures?
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27 Criterion 4: Defense-in-Depth (Continued)
Additional Considerations
- Does the proposed change impact the plants programmatic elements of DID, such as testing and maintenance procedures, operating procedures, inspection procedures, technical specifications, or human performance?
- Does the balance among the layers of defense remain appropriate? Is any decrease in DID commensurate with the expected frequency and consequences of the change?
- Does the proposed change continue to meet the intent of the plants design criteria?
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28 Criterion 5: Safety Margin Does the proposed change result in more than a minimal decrease in safety margin?
- Does the proposed change impact how NRC-accepted codes and standards or NRC-approved alternatives are met?
- Does the proposed change impact any inputs or acceptance criteria used in the safety analysis?
- Does the proposed change provide sufficient safety margin to account for uncertainty?
- Does the proposed change impact how available safety margin is quantified or tracked?
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29 Path Forward
- Determine best way to incorporate the generic criteria into the licensing process.
- Create a new RIPE-like process.
- Incorporate the generic criteria into RIPE or another existing process.
- Evaluate for incorporation into parallel initiatives in development.
- Define very low/minimal safety significance for each of the criteria.
30 Questions?