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| Hazard(s) | HLR(s) | SR(s) | Category | Methods | Applicability | Limitations/Clarifications | Source of Technical Basis | Evidence of Use |
|---|---|---|---|---|---|---|---|---|
| Internal Events | IE-A | IE-A1 IE-A2 IE-A3 IE-A4 IE-A5 IE-A6 IE-A7 IE-A8 IE-A9 IE-A10 |
Initiating Events | Identification of events that challenge normal plant operation and that require successful mitigation to prevent core damage | In addition to outlining general categories for the spectrum of internal-event challenges, ASME/ANS RA-S1.1, through cited SRs, identifies structured, systematic approaches (e.g., master logic diagrams, FMEA) as well as other types of approaches (e.g., operational experience reviews, comprehensive engineering evaluations) for identifying initiating events. NUREG/CR-2300 (e.g., Section 3.4.2, 3.6), NUREG-1150 (e.g., Appendix A), and NUREG/CR-4550 (e.g., Section 3) provides additional detailed guidance on these and other approaches as well as their use in the compilation of a comprehensive list of initiating events, including those resulting from multiple failures. | EPRI 1016741 provides additional clarification on the identification of support system initiating events. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, December 1990 (ML040140729). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal events methodology, Revision 1, Volume 1, January 1990. EPRI 1016741, Support System Initiating Events, December 2008. |
NUREG-1150, NUREG/CR-2300, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | IE-B | IE-B1 IE-B2 IE-B3 IE-B4 IE-B5 |
Initiating Events | Grouping of initiating events so that events in the same group have similar mitigation requirements | ASME/ANS RA-S1.1, through cited SRs, identifies structured, systematic approaches (e.g., master logic diagrams, FMEA) for grouping initiating events and related considerations. NUREG/CR-2300 (e.g., Section 3.4.2, 3.6), NUREG-1150 (e.g., Appendix A), and NUREG/CR-4550 (e.g., Section 3) provides additional detailed guidance on grouping initiating events. | Initiating events, as indicated by ASME/ANS RA-S1.1 through cited SRs, should only be grouped when events can be considered similar in terms of plant response, success criteria, timing, and the effect on the operability and performance of operators and relevant mitigating systems or when events can be bounded by the worst-case impacts within the group. EPRI 1016741 provides additional clarification on grouping of support system initiating events. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, December 1990 (ML040140729). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. EPRI 1016741, Support System Initiating Events, December 2008. |
NUREG-1150, NUREG/CR-2300, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | IE-C | IE-C1 IE-C2 IE-C3 IE-C4 IE-C5 IE-C6 IE-C7 IE-C8 IE-C9 IE-C10 IE-C11 IE-C12 IE-C13 IE-C14 IE-C15 |
Initiating Events | Quantification of the annual frequency of initiating events | NUREG/CR-6823 provides guidance on the evaluation of initiating events probability models (Section 2), component failure and boundary definitions (Section 3), data sources (Section 4), plant-specific data collection and interpretation (Section 5), parameter estimation and model validation (Section 6), trends and aging (Section 7), and parameter estimation using data from different sources (Section 8). Additional guidance can be found in NUREG/CR-2300 (e.g., Section 5), NUREG-1150 (e.g., Appendix A), NUREG/CR-2815 (e.g., Section 5.5), and NUREG/CR-4550 (e.g., Section 8). Note that certain initiating event frequencies may not be derived from data calculations but rather be developed through modeling (e.g., support system initiating event fault trees) to reflect the plant-specific or unique failure modes. EPRI 1016741 provides additional clarification on the quantification of support system initiating events. |
NUREG/CR-6928 (as updated by INL/EXT-21-65055) and NUREG/CR-5750 (as updated by INL/RPT-23-72818) present an analysis of initiating event frequencies at U.S. NPPs. NUREG-1829 provides additional clarification on the estimation of LOCA frequencies. Guidance specific to LOOP and SBO events, including recovery, can be found in NUREG/CR-6890 (as updated by INL/RPT-22-68809). The frequency analysis of ISLOCA events is informed by WCAP-17154-P, NUREG/CR-5928, and NSAC-154. Specific attention to the performance of relief valves, including their role as a potential initiating event, is discussed in NUREG/CR-7037. Section 1-4 of ASME/ANS RA-S1.1, NUREG-1489 (e.g., Section C.5), and NUREG-2255 provide guidance on the formal use of expert judgment. ASME/ANS RA-S1.1, through cited SRs and associated commentary (e.g., Table 2-A.2.1-4), provide guidance on screening criteria as applied to initiating event frequencies. |
NUREG/CR-6823, Handbook of Parameter Estimation for Probabilistic Risk Assessment, September 2003 (ML032900131). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, December 1990 (ML040140729). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, September 2007 (ML070650650). INL/EXT-21-65055, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants: 2020 Update, November 2021. NUREG/CR-5750, Rates of Initiating Events at U.S. Nuclear Power Plants: 1987-1995, February 1999. INL/RPT-23-72818, Initiating Event Rates at U.S. Nuclear Power Plants: 2022 Update, June 2023. NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, April 2008 (ML082250436, ML081060300). EPRI 1016741, Support System Initiating Events, December 2008. WCAP-17154-P, ISLOCA Risk Model, Revision 1, August 2013. NUREG/CR-5928, ISLOCA Research Program, July 1993 (ML072430731). NSAC-154, ISLOCA Evaluation Guidelines, September 1991. NUREG/CR-6890, Reevaluation of Station Blackout Risk at Nuclear Power Plants, December 2005 (ML060200477, ML060200479, ML060200510). INL/RPT-22-68809, Analysis of Loss-of-Offsite Power Events 2021 Update, August 2022. NUREG/CR-7037, Industry Performance of Relief Valves at U.S. Commercial Nuclear Power Plants through 2007, March 2011 (ML110980205). NUREG-2255, Guidance for Conducting Expert Elicitation in Risk-Informed Decisionmaking Activities, September 2022 (ML22242A277). |
NUREG-1150, NUREG-1829, NUREG/CR-2300, NUREG/CR-2815, NUREG/CR-4550, NUREG/CR-5750 (as updated by INL/RPT-23-72818), NUREG/CR-6823, NUREG/CR-6928 (as updated by INL/EXT-21-65055), and NUREG/CR-6890 (as updated by INL/RPT-22-68809) provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | AS-A | AS-A1 AS-A2 AS-A3 AS-A4 AS-A5 AS-A6 AS-A7 AS-A8 AS-A9 AS-A10 AS-A11 |
Accident Sequence Analysis | Accident sequence model event tree structure and sequence definition | The accident sequence analysis captures, within an event tree structure, those operator actions, mitigation systems, and phenomena that can alter the accident sequence progression, including key safety functions, for modeled initiating events (or groups thereof). ASME/ANS RA-S1.1, through cited SRs, provides guidance on accident sequence definition, delineation, and development. Other more detailed guidance, including approaches and techniques for performing accident sequence analysis, can be found in NUREG/CR-2300 (e.g., Section 3.4), NUREG-1489 (e.g., Appendix C.3.3), NUREG-4550 (e.g., Sections 3 and 4), and NUREG/CR-2815 (e.g., Section 4). |
Construction of accident sequence models, including associated parameters (e.g., timing, temperature, pressure, steam) that could potentially affect the operability of mitigating systems, is informed using thermal-hydraulic and other analyses. Example tools include MELCOR, MAAP, SCDAP/RELAP5/MOD3.3, and RETRAN. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG-1489, A Review of NRC Staff Uses of Probabilistic Risk Assessment, June 2007 (ML063540593). |
NUREG-1489, NUREG/CR-2300, NUREG/CR-2815, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | AS-B | AS-B1 AS-B2 AS-B3 AS-B4 AS-B5 AS-B6 AS-B7 |
Accident Sequence Analysis | Dependencies that can impact the ability of the mitigating systems to operate and function. | As indicated by ASME/ANS RA-S1.1, through cited SRs, the accident sequence analysis is to account for dependencies that can impact the ability of the mitigating systems to operate and function. This includes the impact of the initiating event itself; the success or failure of preceding systems, functions, and human actions; and phenomenological conditions created by the accident progression. Detailed guidance on addressing these dependencies appears in NUREG-1489 (e.g., Section C.3.3.4), NUREG-4550 (e.g., Section 6), NUREG/CR-2300 (e.g., Section 3.7), and NUREG/CR-2815 (e.g., Section 4.3.3). Treatment of time-phased dependencies, exampled in ASME/ANS RA-S1.1, through cited SRs, is further explored in EPRI 1009187. |
WCAP-16882-NP provides a specific example of an approach for addressing phenomenological conditions, namely the transport and accumulation of debris in containment following a LOCA, and their impact on the operation of the ECCS in operating PWRs. NEI 16-06, as clarified by the NRC’s 2022 memorandum, provides guidance on the treatment of plant mitigating strategies within the accident sequence analysis. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG-1489, A Review of NRC Staff Uses of Probablistic Risk Assessment, June 2007 (ML063540593). EPRI 1009187, Treatment of Time Interdependencies in Fault Tree Generated Cutset Results, October 2003. WCAP-16882-NP, Revision 1, PRA Modeling of Debris-Induced Failure of Long-Term Core Cooling via Recirculation Sumps, November 2009. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). |
NUREG-1489, NUREG/CR-2300, NUREG/CR-2815, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. |
| Internal Events | SC-A | SC-A1 SC-A2 SC-A3 SC-A4 SC-A5 SC-A6 SC-A7 |
Success Criteria | Development of success criteria consistent with the features, procedures, and operating philosophy of the plant. | To ensure that safety functions are satisfied, success criteria are established, delineating the minimum number or combinations of systems or components required to operate, operator actions, or minimum levels of performance per component during a specific period of time. This also includes defining core damage. Guidance on the overall development of success criteria can be found in NUREG/CR-2300 (e.g., Section 3.4 and 3.5), NUREG-4550 (e.g., Sections 3 and 4), and NUREG/CR-2815 (e.g., Section 3). ASME/ANS RA-S1.1, through cited SRs and Section 1-2, provides guidance on definition of core damage and additional insight on accident sequence mission times. |
For the purposes of establishing success criteria, a minimum accident sequence mission time of 24 hours is set by the ASME/ANS RA-S1.1, but if a safe stable state would not be achieved within this time, additional evaluation or modeling is needed. NEI 16-06, as clarified by the NRC’s 2022 memorandum, provides guidance on the treatment of plant mitigating strategies in the development of success criteria. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). |
NUREG/CR-2300, NUREG/CR-2815, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. The NRC staff found that licensees that reference and apply the guidance identified in NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. |
| Internal Events | SC-B | SC-B1 SC-B2 SC-B3 SC-B4 SC-B5 |
Success Criteria | Justification of success criteria using thermal-hydraulic, structural, and other supporting engineering bases. | Thermohydraulic, structural, and other supporting engineering bases, as noted in ASME/ANS RA-S1.1 through cited SRs, may be used for success criteria and event timing applied within the internal events PRA, provided that the analysis models are consistent with the level of detail of the initiating-event grouping and accident sequence modeling, have sufficient capability to model the conditions of interest, and yield results that are reasonable and acceptable. Related guidance appears in NUREG/CR-2300 (e.g., Section 3.4 and 3.5), NUREG-4550 (e.g., Sections 3 and 4), and NUREG/CR-2815 (e.g., Section 3). | Examples of computer codes and tools that are used to support the determination of success criteria include MELCOR, MAAP, SCDAP/RELAP5/MOD3.3, and RETRAN, among others. Note that per ASME/ANS RA-S1.1, sufficient margin on code-calculated values should exist to allow for limitations of the code, sophistication of the models, and uncertainties in the results. For situations in which there is lack of available information regarding the condition or response of a modeled SSC or a lack of analytical methods on which to base a prediction of SSC condition or response, expert judgment may be used. Section 1-4 of ASME/ANS RA-S1.1, NUREG-1489 (e.g., Section C.5), and NUREG-2255 provide guidance on the formal use of expert judgment. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG-1489, A Review of NRC Staff Uses of Probabilistic Risk Assessment, June 2007 (ML063540593). NUREG-2255, Guidance for Conducting Expert Elicitation in Risk-Informed Decisionmaking Activities, September 2022 (ML22242A277). |
NUREG-1489, NUREG/CR-2300, NUREG/CR-2815, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | SY-A | SY-A1 SY-A2 SY-A3 SY-A4 SY-A5 SY-A6 SY-A7 SY-A8 SY-A9 SY-A10 SY-A11 SY-A12 SY-A13 SY-A14 SY-A15 SY-A16 SY-A17 SY-A18 SY-A19 SY-A20 SY-A21 SY-A22 SY-A23 SY-A24 |
Systems Analysis | Development of System Logic Models | System logic models represent the various system alignments, success criteria, and mission times as well as include the failure modes associated with system maintenance, component actuation and functionality, and associated HFEs. General guidance on the development of such models for use in the internal events PRA can be found in NUREG/CR-2300 (e.g., Sections 3.5 and 3.6), and NUREG-4550 (e.g., Section 5), and NUREG/CR-2815 (e.g., Section 4), whereas guidance specific to fault tree development appears in NUREG-0492. | ASME/ANS RA-S1.1, through cited SRs, provides clarification on several topics related to system logic models, including the definition of system model boundaries and inclusion of components, the level of systems model detail, the definition of component boundaries, the use of supercomponent and module failure events, the effect of variable success criteria, etc. This guidance extends to the inclusion (or exclusion) of certain failure modes as well as potential repair. For complex, risk-significant systems, components, or aspects thereof that make use of novel technologies to reduce risk, sufficient technical basis (e.g., engineering analysis, testing, operating experience, etc.) is needed to inform the development of detailed PRA approaches for use within the internal events PRA. One example of this is PRA models developed to address the performance of RCP seal packages. Along with associated safety evaluations, WCAP-15603-NP, WCAP-16175-P, and PWROG-14001-P/NP document PRA models applicable to various PWR RCP seal designs. NEI 16-06, as clarified by the NRC’s 2022 memorandum, provides guidance on the treatment of plant mitigating strategies within the systems analysis. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NUREG-0492, Fault Tree Handbook, January 1981 (ML100780465). WCAP-15603-NP, WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs, Revision 1, May 2002 (ML021500485) Safety Evaluation of Topical Report WCAP-15603, Revision 1, “WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs,” May 2003 (ML031400376). WCAP-16175-P, “Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants,” Revision 0, January 2004, (ML040340226). Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report WCAP-16175-P, Revision 0, (CE NPSD-1199, Revision 1), “Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants,” February 2007 (ML070240429). PWROG-14001-P/NP, PRA Model for the Generation Ill Westinghouse Shutdown Seal, Revision 1, October 2017 (ML18019A215). Final Safety Evaluation by the Office of Nuclear Reactor Regulation PWROG-14001-P, Revision 1, “PRA Model for The Generation III Westinghouse Shutdown Seal” Pressurized Water Reactor Owners Group Project No. 694, August 2017 (ML17200C876). NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). |
NUREG-0492, NUREG/CR-2300, NUREG/CR-2815, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. |
| Internal Events | SY-B | SY-B1 SY-B2 SY-B3 SY-B4 SY-B5 SY-B6 SY-B7 SY-B8 SY-B9 SY-B10 SY-B11 SY-B12 SY-B13 |
Systems Analysis | CCFs and both intersystem and intrasystem dependencies | CCFs and both intersystem and intrasystem dependencies could influence system performance and thus should be evaluated. This includes evaluating functional, human, and phenomenological effects. Guidance on CCFs appears in NUREG/CR-5485, which is identified as one acceptable means of addressing CCFs within the systems analysis by ASME/ANS RA-S1.1. Detailed guidance on addressing intersystem and intrasystem dependencies appears in NUREG-1489 (e.g., Section C.3.3.4), NUREG-4550 (e.g., Section 6), NUREG/CR-2300 (e.g., Section 3.7), and NUREG/CR-2815 (e.g., Section 4.3.3). |
In evaluating hazards that may impact multiple systems or redundant components in the same system, computer codes and tools are often used (e.g., MELCOR, MAAP, RETRAN, and SCDAP/RELAP5/MOD3.3). For instance, room-heat-up analyses are frequently performed using GOTHIC to inform the assessment of system performance under elevated temperatures. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG/CR-5485, Guidelines on Modeling Common-Cause Failures in Probabilistic Risk Assessment, November 1998. NUREG-1489, A Review of NRC Staff Uses of Probabilistic Risk Assessment, June 2007 (ML063540593). |
NUREG-1489, NUREG/CR-2300, NUREG/CR-2815, NUREG/CR-4550, and NUREG/CR-5485 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | HR-A | HR-A1 HR-A2 HR-A3 |
Pre-Initiator HRA | Identification of those specific routine activities that, if not completed correctly, may impact the availability of equipment necessary to perform system functions. | If not completed correctly, certain activities may impact the availability of equipment necessary to perform system functions modeled in the PRA. These activities include test, inspection, and maintenance activities that require realignment of equipment outside its normal operational or standby status; calibration activities that, if performed incorrectly, can have an adverse impact on the initiation and control of risk-significant SSCs; and activities that simultaneously affect equipment in either different trains of a redundant system or diverse systems. Guidance on the identification of such pre-initiator HFEs can be found in EPRI 3002008094 as well as NUREG-1792 (e.g., Section 4.1), which outlines HRA good practices. | Relevant guidance can also be found in NUREG/CR-2300 (e.g., Section 4) and NUREG/CR-4550 (e.g., Section 7). Additionally, some specific approaches to HRA, including THERP (NUREG-1278 and NUREG/CR-2254) and IDHEAS (NUREG-2256 and NUREG-2198), as well as SHARP1 (EPRI 101711 and EPRI 3583), which is a framework for performing HRA, address this topic, providing a source of additional clarification. Note that IDHEAS (NUREG-2256 and NUREG-2198) is more current than the other approaches and is an attempt by the NRC staff to update and consolidate HRA techniques. EPRI 3002013018, as supplemented by EPRI KBA 2021-001, provides guidance specific to the pre-initiator HRA of plant mitigating strategies as discussed in NEI 16-06 and clarified by the NRC’s 2022 memorandum. |
EPRI 3002008094, Data and Modeling of Pre-Initiator Human Failure Events in Probabilistic Risk Assessment, March 2017. NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, August 1983 (ML071210299). NUREG/CR-2254, A Procedure for Conducting a Human Reliability Analysis for Nuclear Power Plants, May 1983 (ML072430772). NUREG-2198, The General Methodology of An Integrated Human Event Analysis System, May 2021 (ML21127A272). NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment, October 2022 (ML22300A117). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). EPRI 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 2018. EPRI KBA 2021-001, Guidance for Pre-Initiator HRA for FLEX and Portable Equipment, September 2021. |
NUREG-1278, NUREG-1792, NUREG-2198, NUREG-2256, NUREG/CR-2300, NUREG/CR-2254, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in EPRI 3002013018, EPRI KBA 2021-001, and NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. |
| Internal Events | HR-B | HR-B1 HR-B2 |
Pre-Initiator HRA | Screening out activities that need not be addressed explicitly modeled | ASME/ANS RA-S1.1, through cited SRs and Table 1-1.8-1, provides guidance and criteria related to the screening of pre-initiator HFEs. Additional guidance is also provided in EPRI 3002008094, SHARP1 (EPRI 101711 and EPRI 3583), and NUREG-1792 (e.g., Section 4.1), which outlines HRA good practices. | Activities that could simultaneously have an impact on multiple trains of a redundant system or on diverse systems should not be screened out. | EPRI 3002008094, Data and Modeling of Pre-Initiator Human Failure Events in Probabilistic Risk Assessment, March 2017. NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). EPRI 101711, SHARP1A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. |
NUREG-1792 provides guidance specifically developed by the NRC to inform the development of an internal event PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | HR-C | HR-C1 HR-C2 HR-C3 |
Pre-Initiator HRA | Definition of pre-initiator HFEs | HFEs should be defined and modeled within the context of the PRA as well as at the function, system, train, or component level. Guidance for doing so appears in EPRI 3002008094 as well as NUREG-1792 (e.g., Section 4.1), which outlines HRA good practices. | Relevant guidance can also be found in NUREG/CR-2300 (e.g., Section 4) and NUREG/CR-4550 (e.g., Section 7). Additionally, some specific approaches to HRA, including THERP (NUREG-1278 and NUREG/CR-2254) and IDHEAS (NUREG-2256 and NUREG-2198), as well as SHARP1 (EPRI 101711 and EPRI 3583), which is a framework for performing HRA, address this topic, providing a source of additional clarification. Note that IDHEAS (NUREG-2256 and NUREG-2198) is more current than the other approaches and is an attempt by the NRC staff to update and consolidate HRA techniques. EPRI 3002013018, as supplemented by EPRI KBA 2021-001, provides guidance specific to the pre-initiator HRA of plant mitigating strategies as discussed in NEI 16-06 and clarified by the NRC’s 2022 memorandum. |
EPRI 3002008094, Data and Modeling of Pre-Initiator Human Failure Events in Probabilistic Risk Assessment, March 2017. NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NUREG-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, August 1983 (ML071210299). NUREG/CR-2254, A Procedure for Conducting a Human Reliability Analysis for Nuclear Power Plants, May 1983 (ML072430772). NUREG-2198, The General Methodology of An Integrated Human Event Analysis System, May 2021 (ML21127A272). NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment, October 2022 (ML22300A117). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). EPRI 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 2018. EPRI KBA 2021-001, Guidance for Pre-Initiator HRA for FLEX and Portable Equipment, September 2021. |
NUREG-1278, NUREG-1792, NUREG-2198, NUREG-2256, NUREG/CR-2300, NUREG/CR-2254, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in EPRI 3002013018, EPRI KBA 2021-001, and NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. IDHEAS has not been used in any LAR submittal reviewed by the NRC to date. |
| Internal Events | HR-D | HR-D1 HR-D2 HR-D3 HR-D4 HR-D5 HR-D6 |
Pre-Initiator HRA | Assessment of the probabilities of the pre-initiator HFEs | The assessment of the probabilities of the pre-initiator HFEs is performed by using a systematic process that addresses the plant-specific and activity-specific influences on human performance. The following HRA methods are available: THERP (NUREG-1278 and NUREG/CR-2254); the ASEP HRA procedure (NUREG/CR-4772); SPAR-H (NUREG/VR-6883 and INL/EXT-10-18533); and IDHEAS (NUREG-2256 and NUREG-2198). EPRI 3002008094 also provides an approach for collecting utility-specific data to support quantification of pre-initiator HFEs. Furthermore, NUREG-1792 (e.g., Section 4.1) outlines related HRA good practices, whereas NUREG-1842 summarizes the capabilities, limitations, and technical adequacy of the various cited HRA approaches. Relevant guidance can also be found in NUREG/CR-2300 (e.g., Section 4) and NUREG/CR-4550 (e.g., Section 7). Guidance specific to the treatment of dependencies between HFEs as well as the use of minimum values for joint HEPs can be found in EPRI 3002003150 and NUREG-1792 as well as SHARP1 (EPRI 101711 and EPRI 3583), which provides a thorough discussion on the topic of dependency. |
Each identified quantification approach has its own strengths and limitations. The adequacy or appropriateness of each approach for a given HFE is a function of the characteristics of that HFE, the context in which the HFE is defined, the required level of detail or conservatism, etc. Note that IDHEAS (NUREG-2256 and NUREG-2198) is more current than the other approaches and is an attempt by the NRC staff to update and consolidate HRA techniques. As opposed to being an HRA approach or quantification technique per se, the EPRI HRA Calculator is a frequently used software tool that automates the key elements used to quantify HEPs, including the ASEP HRA procedure (NUREG/CR-4772), THERP (NUREG-1278 and NUREG/CR-2254), and SPAR-H (EPRI 101711 and EPRI 3583). NUREG-1792 and EPRI 1021081 both address the need to consider a minimum value for the joint probability of multiple HFEs within the same accident sequence or cutset. Table 2-1 of NUREG-1792 recommends that such joint HEPs should not be below 1E-5, whereas Table 4-4 of EPRI 102081 provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. As documented within RAIs regarding internal events PRAs that support applications (e.g., TSTF-505 and 10 CFR 50.69 LARs), the NRC staff has accepted general use of a 1E-6 floor value for joint HEPs without providing additional justification. EPRI 3002013018, as supplemented by EPRI KBA 2021-001, provides guidance specific to the pre-initiator HRA of plant mitigating strategies as discussed in NEI 16-06 and clarified by the NRC’s 2022 memorandum. |
EPRI 3002008094, Data and Modeling of Pre-Initiator Human Failure Events in Probabilistic Risk Assessment, March 2017. NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). NUREG-1842, Evaluation of Human Reliability Analysis Methods Against Good Practices, September 2006 (ML063200058). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NUREG-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, August 1983 (ML071210299). NUREG/CR-2254, A Procedure for Conducting a Human Reliability Analysis for Nuclear Power Plants, May 1983 (ML072430772). NUREG/CR-4772, Accident Sequence Evaluation Program Human Reliability Analysis Procedure, February 1987 (ML071630060). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. NUREG/CR-6883, SPAR-H Human Reliability Analysis Method, August 2005 (ML051950061). INL/EXT-10-18533, SPAR-H Step-by-Step Guidance, Revision 2, May 2011 (ML112060305). NUREG-2198, The General Methodology of An Integrated Human Event Analysis System, May 2021 (ML21127A272). NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment, October 2022 (ML22300A117). NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). EPRI 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 2018. EPRI KBA 2021-001, Guidance for Pre-Initiator HRA for FLEX and Portable Equipment, September 2021. EPRI 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, October 2010. |
NUREG-1278, NUREG-1792, NUREG-1842, NUREG-2198, NUREG-2256, NUREG/CR-2300, NUREG/CR-2254, NUREG/CR-4550, NUREG/CR-4772, and NUREG/CR-6883 (as supplemented by INL/EXT-10-18533) provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in EPRI 3002013018, EPRI KBA 2021-001, and NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. IDHEAS has not been used in any LAR submittal reviewed by the NRC to date. |
| Internal Events | HR-E | HR-E1 HR-E2 HR-E3 HR-E4 |
Post-Initiator HRA | Identification of operator responses required by modeled accident sequences | Guidance on identifying the set of operator responses required for modeled accident sequences can be found in NUREG-1792 (e.g., Section 5.1), NUREG/CR-2300 (e.g., Section 4) and NUREG/CR-4550 (e.g., Section 7) as well as SHARP1 (EPRI 101711 and EPRI 3583), which is a framework for performing HRA. Lastly, EPRI 3002001048 elaborates further on the use of simulator data to support HRA. | Some specific approaches to HRA, including THERP (NUREG-1278 and NUREG/CR-2254), ATHEANA (NUREG-1624 and NUREG-1880), and IDHEAS (NUREG-2256 and NUREG-2198), address this topic, providing a source of additional clarification. Note that IDHEAS (NUREG-2256 and NUREG-2198) is more current than the other approaches and is an attempt by the NRC staff to update and consolidate HRA techniques. EPRI 3002013018, as supplemented by EPRI KBA 2021‑007, provides guidance specific to the HRA of plant mitigating strategies as discussed in NEI 16-06 and clarified by the NRC’s 2022 memorandum. |
NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NUREG-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, August 1983 (ML071210299). NUREG/CR-2254, A Procedure for Conducting a Human Reliability Analysis for Nuclear Power Plants, May 1983 (ML072430772). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. NUREG-2198, The General Methodology of An Integrated Human Event Analysis System, May 2021 (ML21127A272). NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment, October 2022 (ML22300A117). NUREG-1624, Technical Basis and Implementation Guide for A Technique for Human Event Analysis (ATHEANA), Revision 1, May 2000 (ML003719212, ML003719239). NUREG-1880, ATHEANA User’s Guide, June 2007 (ML072130359). EPRI 3002001048, Use of Simulator Data to Support HRA: A Case Study, December 2013. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). EPRI 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 2018. EPRI KBA 2021‑007, Guidance for Modeling Refueling of FLEX and Portable Equipment, December 2021. |
NUREG-1278, NUREG-1624, NUREG-1792, NUREG-1880, NUREG-2198, NUREG-2256, NUREG/CR-2300, NUREG/CR-2254, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in EPRI 3002013018, EPRI KBA 2021-007, and NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. IDHEAS has not been used in any LAR submittal reviewed by the NRC to date. |
| Internal Events | HR-F | HR-F1 HR-F2 |
Post-Initiator HRA | Definition of post-initiator HFEs | HFEs should be defined and modeled within the context of the PRA as well as at the function, system, train, or component level. Guidance for doing so appears in NUREG-1792 (e.g., Section 5.2), NUREG/CR-2300 (e.g., Section 4) and NUREG/CR-4550 (e.g., Section 7) as well as SHARP1 (EPRI 101711 and EPRI 3583), which is a framework for performing HRA. | Some specific approaches to HRA, including THERP (NUREG-1278 and NUREG/CR-2254), ATHEANA (NUREG-1624 and NUREG-1880), and IDHEAS (NUREG-2256 and NUREG-2198), address this topic, providing a source of additional clarification. Note that IDHEAS (NUREG-2256 and NUREG-2198) is more current than the other approaches and is an attempt by the NRC staff to update and consolidate HRA techniques. EPRI 3002013018, as supplemented by EPRI KBA 2021‑007, provides guidance specific to the HRA of plant mitigating strategies as discussed in NEI 16-06 and clarified by the NRC’s 2022 memorandum. |
NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, August 1983 (ML071210299). NUREG/CR-2254, A Procedure for Conducting a Human Reliability Analysis for Nuclear Power Plants, May 1983 (ML072430772). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. NUREG-2198, The General Methodology of An Integrated Human Event Analysis System, May 2021 (ML21127A272). NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment, October 2022 (ML22300A117). NUREG-1624, Technical Basis and Implementation Guide for A Technique for Human Event Analysis (ATHEANA), Revision 1, May 2000 (ML003719212, ML003719239). NUREG-1880, ATHEANA User’s Guide, June 2007 (ML072130359). EPRI 3002001048, Use of Simulator Data to Support HRA: A Case Study, December 2013. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). EPRI 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 2018. EPRI KBA 2021‑007, Guidance for Modeling Refueling of FLEX and Portable Equipment, December 2021. |
NUREG-1278, NUREG-1624, NUREG-1792, NUREG-1880, NUREG-2198, NUREG-2256, NUREG/CR-2300, NUREG/CR-2254, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in EPRI 3002013018, EPRI KBA 2021-007, and NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. IDHEAS has not been used in any LAR submittal reviewed by the NRC to date. |
| Internal Events | HR-G | HR-G1 HR-G2 HR-G3 HR-G4 HR-G5 HR-G6 HR-G7 HR-G8 HR-G9 HR-G10 |
Post-Initiator HRA | Assessment of the probabilities of the post-initiator HFEs | The assessment of the probabilities of the post-initiator HFEs is performed by using a systematic process that addresses the plant-specific and activity-specific influences on human performance. The following HRA methods are available: THERP (NUREG-1278 and NUREG/CR-2254); the ASEP HRA procedure (NUREG/CR-4772); the HCR/ORE method (EPRI 100259, EPRI 6937, and EPRI 6560); the CBDT method (EPRI 100259); SPAR-H (EPRI 101711 and EPRI 3583); ATHEANA (NUREG-1624 and NUREG-1880); and IDHEAS (NUREG-2256 and NUREG-2198). Furthermore, NUREG-1792 (e.g., Section 5.3) outlines related HRA good practices, whereas NUREG-1842 summarizes the capabilities, limitations, and technical adequacy of the various cited HRA approaches. Relevant guidance can also be found in NUREG/CR-2300 (e.g., Section 4) and NUREG/CR-4550 (e.g., Section 7). Guidance specific to the treatment of dependencies between HFEs as well as the use of minimum values for joint HEPs can be found in EPRI 3002003150 and NUREG-1792 as well as SHARP1 (EPRI 101711 and EPRI 3583), which provides a thorough discussion on the topic of dependency. |
Each identified quantification approach has its own strengths and limitations. The adequacy or appropriateness of each approach for a given HFE is a function of the characteristics of that HFE, the context in which the HFE is defined, the required level of detail or conservatism, etc. Note that IDHEAS (NUREG-2256 and NUREG-2198) is more current than the other approaches and is an attempt by the NRC staff to update and consolidate HRA techniques. As opposed to being an HRA approach or quantification technique per se, the EPRI HRA Calculator is a frequently used software tool that automates the key elements used to quantify HEPs using the CBDT method (EPRI 100259), the HCR/ORE method (EPRI 100259, EPRI 6937, and EPRI 6560), the ASEP HRA procedure (NUREG/CR-4772), THERP (NUREG-1278 and NUREG/CR-2254), or SPAR-H (EPRI 101711 and EPRI 3583). NUREG-1792 and EPRI 1021081 both address the need to consider a minimum value for the joint probability of multiple HFEs within the same accident sequence or cutset. Table 2-1 of NUREG-1792 recommends that such joint HEPs should not be below 1E-5, whereas Table 4-4 of EPRI 102081 provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. As documented within RAIs regarding internal events PRAs that support applications (e.g., TSTF-505 and 10 CFR 50.69 LARs), the NRC staff has accepted general use of a 1E-6 floor value for joint HEPs without providing additional justification. EPRI 3002013018, as supplemented by EPRI KBA 2021‑007, provides guidance specific to the HRA of plant mitigating strategies as discussed in NEI 16-06 and clarified by the NRC’s 2022 memorandum. Note that estimating HEPs for such HFEs is more challenging because associated HRA approaches may not address all potential plant-specific considerations, and not all HFEs may lend themselves to ready evaluation with the commonly used HRA tools. |
EPRI 3002008094, Data and Modeling of Pre-Initiator Human Failure Events in Probabilistic Risk Assessment, March 2017. NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). NUREG-1842, Evaluation of Human Reliability Analysis Methods Against Good Practices, September 2006 (ML063200058). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, August 1983 (ML071210299). NUREG/CR-2254, A Procedure for Conducting a Human Reliability Analysis for Nuclear Power Plants, May 1983 (ML072430772). NUREG/CR-4772, Accident Sequence Evaluation Program Human Reliability Analysis Procedure, February 1987 (ML071630060). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. NUREG/CR-6883, SPAR-H Human Reliability Analysis Method, August 2005 (ML051950061). INL/EXT-10-18533, SPAR-H Step-by-Step Guidance, Revision 2, May 2011 (ML112060305). NUREG-2198, The General Methodology of An Integrated Human Event Analysis System, May 2021 (ML21127A272). NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment, October 2022 (ML22300A117). EPRI 100259, An Approach to the Analysis of Operator Actions in PRA, June 1992. EPRI 6937, Operator Reliability Experiments Using Power Plant Simulators, Volumes 1-3, August 1980. EPRI 6560, A Human Reliability Analysis Approach Using Measurements for Individual Plant Examination, January 1990. EPRI 100259, An Approach to the Analysis of Operator Actions in PRA, June 1992. NEI 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, August 2016 (ML16286A297). NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). EPRI 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 2018. EPRI KBA 2021‑007, Guidance for Modeling Refueling of FLEX and Portable Equipment, December 2021. EPRI 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, October 2010. |
NUREG-1278, NUREG-1792, NUREG-1842, NUREG-2198, NUREG-2256, NUREG/CR-2300, NUREG/CR-2254, NUREG/CR-4550, NUREG/CR-4772, and NUREG/CR-6883 (as supplemented by INL/EXT-10-18533) provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. The NRC staff found that licensees that reference and apply the guidance identified in EPRI 3002013018, EPRI KBA 2021-007, and NEI 16-06, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. IDHEAS has not been used in any LAR submittal reviewed by the NRC to date. |
| Internal Events | HR-H | HR-H1 HR-H2 HR-H3 HR-H4 |
Post-Initiator HRA | Modeling of recovery actions | Recovery actions are applied at the cutset or scenario level and are demonstrated to be both plausible and feasible. Guidance on such actions appears in NUREG-1792 (e.g., Section 5.4) as well as Section 3.3 of SHARP1 (EPRI 101711 and EPRI 3583), which is a framework for performing HRA. | Recovery is associated with operators performing actions to compensate for the failed automatic actions but does not include repair of the equipment. | NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. |
NUREG-1792 provides guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | DA-A | DA-A1 DA-A2 DA-A3 DA-A4 |
Data Analysis | Definition of parameters | NUREG/CR-6823 provides guidance on the evaluation of basic event probability models (Section 2) as well as component failure and boundary definitions (Section 3). Additional guidance, including on the identification of basic events requiring probabilities and parameters to be estimated, can be found in NUREG/CR-2300 (e.g., Sections 3 and 5), NUREG/CR-2815 (e.g., Section 5), and NUREG/CR-4550 (e.g., Section 8) as well as ASME/ANS RA-S1.1, through cited SRs and associated commentary (e.g., Table 2-A.2.6-2). | The definition of parameters should be done in a manner consistent with the system analysis in terms of, for instance, the logic model, basic event boundary, and failure mode. | NUREG/CR-6823, Handbook of Parameter Estimation for Probabilistic Risk Assessment, September 2003 (ML032900131). NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990 (5066882). |
NUREG/CR-2300, NUREG/CR-2815, NUREG/CR-4550, and NUREG/CR-6823 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | DA-B | DA-B1 DA-B2 |
Data Analysis | Grouping components into a homogeneous population for parameter estimation | ASME/ANS RA-S1.1, through cited SRs, discusses the grouping of components according to type and characteristics of their usage, as well as the exclusion of outliers. NUREG/CR-2300 (e.g., Section 5), NUREG/CR-2815 (e.g., Section 5), NUREG/CR-4550 (e.g., Section 8), and NUREG/CR-6823 (e.g., Section 5) provide detailed guidance on such grouping for the purposes of parameter estimation. | Additional consideration is given to the grouping of components for parameter estimation within NUREG/CR-6928 (as updated by INL/EXT-21-65055) and NUREG/CR-7037, which present analysis of industry-average performance of components at U.S. NPPs. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG/CR-6823, Handbook of Parameter Estimation for Probabilistic Risk Assessment, September 2003 (ML032900131). NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, September 2007 (ML070650650). INL/EXT-21-65055, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants: 2020 Update, November 2021. |
NUREG/CR-2300, NUREG/CR-2815, NUREG/CR-4550, NUREG/CR-6823, NUREG/CR-6928 (as updated by INL/EXT-21-65055), and NUREG/CR-6890 (as updated by INL/RPT-22-68809) provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | DA-C | DA-C1 DA-C2 DA-C3 DA-C4 DA-C5 DA-C6 DA-C7 DA-C8 DA-C9 DA-C10 DA-C11 DA-C12 DA-C13 DA-C14 DA-C15 DA-C16 |
Data Analysis | Choosing generic parameter estimates and a collection of plant-specific data | NUREG/CR-6823 provides guidance on data sources (Section 4) as well as plant-specific data collection and interpretation (Section 5), reflecting considerations within the cited SRs of ASME/ANS RA-S1.1. NUREG/CR-2300 (e.g., Section 5.4), NUREG/CR-2815 (e.g., Section 5), NUREG/CR-4550 (e.g., Section 8), and NUREG-1489 (e.g., Appendix C) provide additional detailed guidance. | Examples of generic parameter estimates, and associated sources include: NUREG/CR-6928 (as updated by INL/EXT-21-65055) for component failure rates and probabilities; NUREG/CR-5497 (as updated by INL/EXT-21-62940) for CCFs; and NUREG/CR-6890 (as updated by INL/RPT-22-68809) for LOOP and SBO events (including recovery). PWROG‑18042‑NP and PWROG-18043-P, as clarified by the NRC’s 2022 memorandum (ML22014A084), present an approach for collecting and analyzing operating experience data associated with portable equipment as well as provide generic failure probabilities for portable equipment for use in developing plant-specific failure probabilities. Note that estimating the failure probabilities associated with this equipment is challenging because of limited associated operating experience. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG/CR-5497, Common-Cause Failure Parameter Estimations, October 1998. INL/EXT-21-62940, CCF Parameter Estimations 2020 Update, Revision 1, August 2022. NUREG/CR-6268, Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding, Revision 1, September 2007 (ML072970404). NUREG/CR-6823, Handbook of Parameter Estimation for Probabilistic Risk Assessment, September 2003 (ML032900131). NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, September 2007 (ML070650650). INL/EXT-21-65055, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants: 2020 Update, November 2021. NUREG/CR-6890, Reevaluation of Station Blackout Risk at Nuclear Power Plants, December 2005 (ML060200477, ML060200479, ML060200510). INL/RPT-22-68809, Analysis of Loss-of-Offsite Power Events 2021 Update, August 2022. NUREG-1489, A Review of NRC Staff Uses of PRA, June 2007 (ML063540593). PWROG‑18042‑NP, FLEX Equipment Data Collection and Analysis, Revision 1, February 2022 (ML22123A259). PWROG-18043-P, FLEX Equipment Data Collection and Analysis, Revision 1, August 2021. NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). |
NUREG-1489, NUREG/CR-2300, NUREG/CR-2815, NUREG/CR-4550, NUREG/CR-6823, NUREG/CR-5497 (as updated by INL/EXT-21-62940), NUREG/CR-6268, NUREG/CR-6928 (as updated by INL/EXT-21-65055), and NUREG/CR-6890 (as updated by INL/RPT-22-68809) provide guidance specifically developed by the NRC to inform the development of an internal events PRA. The NRC staff found that licensees that reference and apply the guidance identified in PWROG‑18042‑NP and PWROG-18043-P, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. |
| Internal Events | DA-D | DA-D1 DA-D2 DA-D3 DA-D4 DA-D8 |
Data Analysis | Calculation of parameters based on relevant generic industry and plant-specific evidence | NUREG/CR-6823 provides guidance on parameter estimation and model validation, including Bayesian techniques (Section 6), trends and aging (Section 7), and parameter estimation using data from different sources (Section 8). NUREG/CR-2300 (e.g., Section 5), NUREG/CR-2815 (e.g., Section 5), NUREG/CR-4550 (e.g., Section 8), and NUREG-1489 (e.g., Appendix C) provide additional detailed guidance. | If modifications to plant design or operating practice lead to a condition where past data are no longer representative of current performance, the use of old data should be limited, consistent with ASME/ANS RA-S1.1 and cited guidance. PWROG‑18042‑NP and PWROG-18043-P present an approach for collecting and analyzing operating experience data associated with portable equipment as well as provide generic failure probabilities for portable equipment for use in developing plant-specific failure probabilities. |
NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NUREG/CR-6823, Handbook of Parameter Estimation for Probabilistic Risk Assessment, September 2003 (ML032900131). NUREG-1489, A Review of NRC Staff Uses of PRA, June 2007 (ML063540593). PWROG‑18042‑NP, FLEX Equipment Data Collection and Analysis, Revision 1, February 2022 (ML22123A259). PWROG-18043-P, FLEX Equipment Data Collection and Analysis, Revision 1, August 2021. NRC Memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, May 2022 (ML22014A084). |
NUREG-1489, NUREG/CR-2300, NUREG/CR-2815, NUREG/CR-4550, and NUREG/CR-6823 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. The NRC staff found that licensees that reference and apply the guidance identified in PWROG‑18042‑NP and PWROG-18043-P, as clarified by the NRC’s 2022 memorandum (ML22014A084), will likely minimize the need for requests for additional information in future license application submittals. |
| Internal Events | DA-D | DA-D5 DA-D6 DA-D7 |
Data Analysis | Estimation of CCF Parameters | Guidance on the estimation of CCF parameters consistent with plant experience and component boundaries is provided within NUREG/CR-5485 and NUREG/CR-6268, which address relevant CCF quantification models such as the beta factor, alpha factor, multiple Greek letter, and binomial failure rate. Specific guidance on inter-system CCFs is provided in EPRI 1015096. Lastly, further insights into CCF modeling and its impacts on risk-informed decision-making can be found in EPRI 3002020764. | Examples of generic parameter estimates, and associated sources include NUREG/CR-5497 (as updated by INL/EXT-21-62940). | NUREG/CR-5485, Guidelines on Modeling Common Cause Failures in Probabilistic Risk Assessment, November 1998. NUREG/CR-5497, Common-Cause Failure Parameter Estimations, October 1998. INL/EXT-21-62940, CCF Parameter Estimations 2020 Update, Revision 1, August 2022. NUREG/CR-6268, Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding, Revision 1, September 2007 (ML072970404). EPRI 1015096, Investigation of Inter-System Common-Cause Failures, December 2007. EPRI 3002020764, Consideration of Common Cause Failures in Risk-Informed Decision-Making, October 2021. |
NUREG/CR-5485, NUREG/CR-5497 (as updated by INL/EXT-21-62940), and NUREG/CR-6268 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | QU-A | QU-A1 QU-A2 QU-A3 QU-A4 QU-A5 |
Quantification | Integration of the Level 1 PRA model to support quantification of CDF and LERF | The individual parts of the Level 1 internal events PRA (e.g., accident sequences, system models, data, and HRA elements) are integrated to allow for quantification of individual accident sequences and the mean CDF as well as to support the quantification of LERF. Relevant guidance can be found in NUREG-2300 (e.g., Sections 6 and 12), NUREG-4550 (e.g., Sections 10 and 12), NUREG/CR-2815 (e.g., Section 6), and NUREG-1489 (e.g., Section C.6). | As indicated by ASME/ANS RA-S1.1, through cited SRs, the state-of-knowledge correlation must be considered unless it can be demonstrated that the effect of the state-of-knowledge is not risk significant. For example, it has been found that risk-significant cutsets contributing to ISLOCA frequency that involve rupture of multiple valves can exhibit a significant SOKC impact. Appendix 6-A to NUREG-1855 provides further clarification on this topic. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG-1489, A Review of NRC Staff Uses of Probabilistic Risk Assessment, June 2007 (ML063540593). NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017 (ML17062A466). |
NUREG-1489, NUREG-1855, NUREG/CR-2300, NUREG/CR-2815, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | QU-B | QU-B1 QU-B2 QU-B3 QU-B4 QU-B5 QU-B6 QU-B7 QU-B8 QU-B9 QU-B10 |
Quantification | Quantification of the Level 1 PRA model | The internal events PRA model is to be quantified using appropriate models and codes, addressing method-specific limitations and features. Traditional PRA quantification methods make use of various approximations (e.g., rare event, delete-term, and minimum cutset upper bound); however, alternative methods, such as BDD logic, can be applied to calculate more accurate results and importance measures. General guidance on quantification appears in NUREG-2300 (e.g., Section 6) and NUREG-4550 (e.g., Section 10). | Various software codes and tools are used to support quantification of the Level 1 PRA model. These include the EPRI Phoenix Architect, RiskSpectrum, WinNUPRA, RISKMAN, and SAPHIRE. Regarding the toolset with the EPRI Phoenix Architect, it is the most frequently used and contains several supporting analysis tools, e.g., CAFTA, PRAQuant, UNCERT, ACUBE, Direct Probability Calculator, Qrecover, SysImp, and FRANX. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. |
NUREG/CR-2300 and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | QU-C | QU-C1 QU-C2 QU-C3 |
Quantification | Treatment of operator-action dependencies | Guidance specific to the treatment of dependencies between HFEs as well as the use of minimum values for joint HEPs can be found in EPRI 3002003150 and NUREG-1792 as well as SHARP1 (EPRI 101711 and EPRI 3583), which provides a thorough discussion on the topic of dependency. | NUREG-1792 and EPRI 1021081 both address the need to consider a minimum value for the joint probability of multiple HFEs within the same accident sequence or cutset. Table 2-1 of NUREG-1792 recommends that such joint HEPs should not be below 1E-5, whereas Table 4-4 of EPRI 102081 provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. As documented within RAIs regarding internal events PRAs that support applications (e.g., TSTF-505 and 10 CFR 50.69 LARs), the NRC staff has accepted general use of a 1E-6 floor value for joint HEPs without providing additional justification. |
EPRI 3002008094, Data and Modeling of Pre-Initiator Human Failure Events in Probabilistic Risk Assessment, March 2017. NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). EPRI 101711, SHARP1—A Revised Systematic Human Action Reliability Procedure, March 1993. EPRI 3583, Systematic Human Action Reliability Procedure, June 1984. EPRI 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, October 2010. |
NUREG-1792 provides guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | QU-D | QU-D1 QU-D2 QU-D3 QU-D4 QU-D5 QU-D6 QU-D7 QU-D8 |
Quantification | Review of quantification results for correctness, completeness, and consistency | Review of the quantification results for the internal events PRA is informed by guidance in NUREG/CR-2300 (e.g., Section 6), NUREG/CR-4550 (e.g., Section 12), NUREG-1489 (e.g., Section C.6.5), and NUREG/CR-2815 (e.g., Section 6). Additional guidance on measures of risk importance can be found in NUREG/CR-3385 and EPRI 105396. | Various software codes and tools that are used to support quantification of the Level 1 PRA model under HLR-QU-B are also used to support the review of the quantification results as well as perform importance and sensitivity analyses. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency, Revision 1, Volume 1, January 1990. NUREG-1489, A Review of NRC Staff Uses of Probabilistic Events Methodology, June 2007 (ML063540593). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). NUREG/CR-3385, Measures of Risk Importance and Their Applications, July 1983 (ML071690031). EPRI 105396, PSA Applications Guide, August 1995. |
NUREG-1489, NUREG/CR-2815, NUREG/CR-2300, NUREG/CR-3385, and NUREG/CR-4550 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | LE-A | LE-A1 LE-A2 LE-A3 LE-A4 LE-A5 |
LERF Analysis | Grouping of core damage states into PDSs based on their accident progression attributes | The purpose of the LERF analysis is to address the physical characteristics at the time of core damage that can influence LERF as well as the preceding accident sequence characteristics. In doing so, core damage sequences are grouped into PDSs. Guidance for performing this analysis can be found in NUREG/CR-6595 as well as WCAP-16341-P, which is consistent with the approach outlined in NUREG/CR-6595. Additional guidance also appears in NUREG-2300 (e.g., Sections 7 and 8) and NUREG/CR-4550 (e.g., Section 11), as well as NSAC-159-V1 (e.g., Section 4), with NSAC-159-V2 and NSAC-159-V3 providing guidelines for PWRs and BWRs, respectively. |
As noted by ASME/ANS RA-S1.1, there are multiple approaches that can be used to explicitly account for dependencies between the Level 1 and LERF/Level 2 PRA models, including treatment in the LERF/Level 2 PRA, expansion of the Level 1 PRA, construction of a bridge tree, transfer of information via PDS, or a combination of these approaches. | NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Revision 1, October 2004 (ML043240040). WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, November 2005. NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NSAC-159-V1, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 1: Main Report, October 1991. NSAC-159-V2, Generic Framework for IPE Back-End (LEVEL 2) Analysis; Volume 2: PWR Implementation Guidelines, October 1991. NSAC-159-V3, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 3: BWR Implementation Guidelines, October 1991. |
NUREG/CR-2300, NUREG/CR-4550, and NUREG/CR-6595 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI/NSAC and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | LE-B | LE-B1 LE-B2 LE-B3 |
LERF Analysis | Evaluation of contributors to LERF | LERF contributors appear in Table 2-2.8-9 of ASME/ANS RA-S1.1, which indicates that NUREG/CR-6595 may be used to identify them for applicability and that lessons learned from Fukushima regarding scenarios with hydrogen leakage should be considered. Guidance on the evaluation of LERF contributors can also be found in WCAP-16341-P, which is consistent with the approach outlined in NUREG/CR-6595. Lastly, further insights are found in NUREG-2300 (e.g., Section 7) and NUREG-4550 (e.g., Section 11) as well as NSAC-159-V1 (e.g., Sections 4 and 5) with NSAC-159-V2 and NSAC-159-V3 providing guidelines for PWRs and BWRs, respectively. |
NUREG/CR-6595 and WCAP-16341-P represent simplified approaches to LERF analysis. More detailed, plant-specific analysis can be performed using calculation codes and tools (e.g., CONTAIN, MELCOR, MAAP, SCDAP/RELAP5/MOD3.3, and RETRAN). NUREG/CR-6995 provides one example of such an analysis. EPRI 3002015998 further describes the steps that can be taken to perform a plant-specific evaluation that can result in a reduction in the conditional probability of a large-early release. Section 1-4 of ASME/ANS RA-S1.1, NUREG-1489 (e.g., Section C.5), and NUREG-2255 provide guidance on the formal use of expert judgment. |
NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Revision 1, October 2004 (ML043240040). WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, November 2005. NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Revision 1, Volume 1, January 1990. NSAC-159-V1, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 1: Main Report, October 1991. NSAC-159-V2, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 2: PWR Implementation Guidelines, October 1991. NSAC-159-V3, Generic Framework for IPE Back-End (LEVEL 2) Analysis; Volume 3: BWR Implementation Guidelines, October 1991. NUREG/CR-6995, SCDAP/RELAP5 Thermal-Hydraulic Evaluations of the Potential for Containment Bypass during Extended Station Blackout Severe Accident Sequences in a Westinghouse Four-Loop PWR, March 2010 (ML101130544). EPRI 3002015998, Practical Examples for the Improvement of Realism in Large Early Release Estimates, November 2021. NUREG-2255, Guidance for Conducting Expert Elicitation in Risk-Informed Decisionmaking Activities, September 2022 (ML22242A277). |
NUREG/CR-2300, NUREG/CR-4550, NUREG/CR-6595, and NUREG/CR-6995 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI/NSAC and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | LE-C | LE-C1 LE-C2 LE-C3 LE-C4 LE-C5 LE-C6 LE-C7 LE-C8 LE-C9 LE-C10 LE-C11 LE-C12 |
LERF Analysis | Performance of accident progression analysis to identify those progressions that have potential for a large early release | Development of accident sequences to a level of detail to account for the potential contributors to LERF is addressed by NUREG/CR-6595 (e.g., Appendix A) and NUREG-1935, which provide discussion and examples of LERF source terms. Additional guidance appears in WCAP-16341-P, which is consistent with the approach outlined in NUREG/CR-6595. Lastly, further insights are found in NUREG-2300 (e.g., Section 7), NUREG-1489 (e.g., Appendix C.4.4), and NUREG-4551 (e.g., Section 6 and Appendix C) as well as NSAC-159-V1 (e.g., Sections 4 and 5) with NSAC-159-V2 and NSAC-159-V3 providing guidelines for PWRs and BWRs, respectively. |
Accident sequence, data, human reliability, and system analysis in support of the LERF analysis should be performed in a manner consistent with other applicable SRs (e.g., AS, DA, HR, SY). | NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Revision 1, October 2004 (ML043240040). NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report, November 2012 (ML12332A057, ML12332A058). WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, November 2005. NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4551, Evaluation of Severe Accident Risks, Revision 1, Volume 1, December 1993 (ML072710062). NSAC-159-V1, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 1: Main Report, October 1991. NSAC-159-V2, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 2: PWR Implementation Guidelines, October 1991. NSAC-159-V3, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 3: BWR Implementation Guidelines, October 1991. |
NUREG-1935, NUREG/CR-2300, NUREG/CR-4551, and NUREG/CR-6595 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI/NSAC and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | LE-D | LE-D1 LE-D2 LE-D3 LE-D4 LE-D5 LE-D6 LE-D7 |
LERF Analysis | Evaluation of the containment’s ability to prevent a large early release | The LERF analysis should include an evaluation of the containment’s ability to prevent a large early release, including the impact of the accident sequence on the structural capability of the containment, the ability of the containment isolation system to contain the release, the potential for a containment bypass to occur (e.g., ISLOCA), and the potential for pressure-induced or thermally induced SGTRs to occur. Related guidance can be found in NUREG/CR-6595 and WCAP-16341-P, which is consistent with the approach outlined in NUREG/CR-6595. Further insights are found in NUREG-2300 (e.g., Section 7) and NUREG-4551 (e.g., Section 6 and Appendix C) as well as NSAC-159-V1 (e.g., Sections 4 and 5) with NSAC-159-V2 and NSAC-159-V3 providing guidelines for PWRs and BWRs, respectively. |
The analysis of ISLOCA events is informed by WCAP-17154-P, NUREG/CR-5928, and NSAC-154. Regarding the analysis of SGTRs, an acceptable approach per ASME/ANS RA-S1.1 is one that arrives at plant-specific split fractions by selecting the steam generator tube conditional failure probabilities based on current industry guidance for induced steam generator failure of similarly designed steam generators and loop piping. One example of such an analysis is found in NUREG-2195. |
NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Revision 1, October 2004 (ML043240040). WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, November 2005. NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4551, Evaluation of Severe Accident Risks, Revision 1, Volume 1, December 1993 (ML072710062). NSAC-159-V1, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 1: Main Report, October 1991. NSAC-159-V2, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 2: PWR Implementation Guidelines, October 1991. NSAC-159-V3, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 3: BWR Implementation Guidelines, October 1991. WCAP-17154-P, ISLOCA Risk Model, Revision 1, August 2013. NUREG/CR-5928, ISLOCA Research Program, July 1993 (ML072430731). NSAC-154, ISLOCA Evaluation Guidelines, September 1991. NUREG-2195, Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes, May 2018 (ML18122A012). |
NUREG-2195, NUREG/CR-2300, NUREG/CR-4551, NUREG/CR-5928, and NUREG/CR-6595 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI/NSAC and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | LE-E | LE-E1 LE-E2 LE-E3 LE-E4 LE-F1 LE-F2 |
LERF Analysis | Quantitative evaluation of the LERF contributors | NUREG/CR-6595 and WCAP-16341-P, which is consistent with the approach outlined in NUREG/CR-6595, provides guidance on parameter estimation to support quantification of LERF contributors. Additional guidance appears in NUREG-2300 (e.g., Section 7 and 12) and NUREG-4551 (e.g., Section 9) as well as NSAC-159-V1 (e.g., Sections 6 and 7) with NSAC-159-V2 and NSAC-159-V3 providing guidelines for PWRs and BWRs, respectively. | Parameter estimates and quantification in support of the LERF analysis should be performed in a manner consistent with other applicable SRs (e.g., DA, HR, QU). | NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Revision 1, October 2004 (ML043240040). WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, November 2005. NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4551, Evaluation of Severe Accident Risks, Revision 1, Volume 1, December 1993 (ML072710062). NSAC-159-V1, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 1: Main Report, October 1991. NSAC-159-V2, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 2: PWR Implementation Guidelines, October 1991. NSAC-159-V3, Generic Framework for IPE Back-End (Level 2) Analysis; Volume 3: BWR Implementation Guidelines, October 1991. |
NUREG/CR-2300, NUREG/CR-4551, and NUREG/CR-6595 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. In general, industry-developed methods, including those within EPRI/NSAC and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Events | Multiple | IE-A11 IE-D2 AS-A12 AS-C2 SC-A8 SC-B6 SC-C2 SY-A25 SY-B14 SY-C2 HR-A4 HR-E5 HR-I1 DA-A5 DA-E1 QU-A6 QU-F1 QU-F2 QU-F4 LE-G1 LE-G2 LE-G4 |
Documentation | Documentation of the internal events PRA and its results | ASME/ANS RA-S1.1 provides guidance for each relevant SR to allow for the review and written basis of the internal events PRA. NUREG-2300 and NUREG/CR-2815 (e.g., Sections 1 and 7) also provide general guidance. | Note that the appropriateness of any given documentation practice is ultimately dependent on the given application to which the internal events PRA is applied. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-2815, Probabilistic Safety Analysis Procedures Guide, January 1984 (ML063550253). |
NUREG/CR-2815 and NUREG/CR-2300 provide guidance specifically developed by the NRC to inform the development of an internal events PRA. |
| Internal Events | Multiple | QU-E1 QU-E2 QU-F3 LE-F3 LE-G3 IE-D1 AS-C1 SC-C1 SY-C1 HR-I2 DA-E2 |
Uncertainty | Identification, characterization, and documentation of uncertainties in the internal events PRA | Detailed guidance on treating different sources of epistemic uncertainty (i.e., parameter, model, and completeness uncertainties) are provided in NUREG-1855 as well as EPRI 1016737 and EPRI 1026511. Consistent with relevant SRs in ASME/ANS RA-S1.1, methods to assess the effects of model uncertainties and related assumptions are discussed. | Note that EPRI 3002020762 provides guidance on the analysis of uncertainty related to severe accident analysis using MAAP. Also, additional insights and guidance on uncertainty can be found in the sources of technical basis cited elsewhere for the internal events PRA, including NUREG-1489, (e.g., Section C.6), NUREG-2300 (e.g., Section 12), NUREG-4550 (e.g., Section 12), NUREG-4551 (e.g., Section 9), NSAC-159-V1 (e.g., Section 7), NUREG/CR-6595 (e.g., Section 7), and WCAP-16341-P (e.g., Section 7). Note that EPRI 1016737 provides generic industry lists of internal events (including flooding) sources of modeling uncertainty and that EPRI 1026511 (which has fire and seismic lists) presents equivalent sources for fire, seismic, and Level 2 PRA. |
NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017 (ML17062A466). EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012. EPRI 3002020762, Severe Accident Uncertainty Quantification and Analysis Using the Modular Accident Analysis Program (MAAP), September 2021. |
NUREG-1855 provide guidance specifically developed by the NRC to address matters of uncertainty, including related modeling uncertainty. Guidance in EPRI 1016737 and EPRI 1026511, as stated within NUREG-1855, complements that of the NUREG given that they expand upon specific methods to treat uncertainty and their use in applications. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFPP-A IFPP-B |
IFPP-A1 IFPP-B1 IFPP-B2 IFPP-B3 IFPP-B4 |
Plant Analysis Boundary and Partitioning | Definition of the global plant analysis boundary and its partitioning into flood areas for the internal flood PRA | The plant partitioning analysis ensures that all areas or locations within the licensee-controlled area where an internal flood could adversely affect any equipment are included in the internal flood PRA. Guidance on performing this analysis and defining flood areas for treatment within the internal flood PRA is provided in Section 3.1 of EPRI 1019194. | As stated in Table 3-A.2.1-3 of ASME/ANS RA-S1.1, physical characteristics that are used to define flood areas may include walls (watertight or nonwatertight), partial-height walls, doors (watertight or nonwatertight), hatches, berms, dikes, or curbs. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFSO-A | IFSO-A1 IFSO-A2 IFSO-A3 IFSO-A4 IFSO-A5 IFSO-A6 IFSO-A7 |
Flood Source Identification | Identification of flood sources and associated failure mechanisms | ASME/ANS RA-S1.1 states that flood sources include equipment (e.g., piping, valves, pumps) that are connected to fluid systems; other plant internal sources (e.g., tanks, pools); plant external sources of water (i.e., ultimate heat sinks such as reservoirs or rivers); in-leakage from other flood areas (e.g., back flow through drains, doorways); and for multi-unit sites with shared SSCs, potential sources with multi-unit or cross-unit impact. These sources could be impacted by different modes of component failure, including human-induced mechanisms, inadvertent actuation, and other events that result in a release of water, steam, or other liquids into a flood area. Section 3.2 of EPRI 1019194 provides guidance on the systematic identification of flood sources, including their associated failure mechanisms that can lead to various flood hazards, within each defined flood area. |
As clarified by Table 3-A.2.2-2 of ASME/ANS RA-S1.1, flood sources include water, steam, and other liquids (e.g., lubricating oil, fuel oil), but the flooding hazard considered in the scope of the internal flood PRA for oil sources is only the wetting hazard, which is, typically, qualitatively screened out in the internal flood PRA. As clarified by EPRI 1019194, the scope of the internal flood PRA includes all fluid sources outside of the containment structure that have a potential to cause flooding and/or HELB impacts. For example, systems within the RCS pressure boundary whose failure would represent a LOCA are excluded from the internal flood PRA as these events are addressed as part of the internal events PRA. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFSO-A | IFSO-A6 | Flood Source Characterization | Characterization of releases | Section 6.1 of EPRI 1019194 provides guidance on the characterization of releases for flood sources and their respective failure mechanisms. Characteristics addressed include the source capacity, temperature, and pressure as well as breach type (e.g., spray, flood, major flood) and flow rates. | Additional guidance on flood rate characterization, including the establishment of flood rate intervals and maximum break size, is provided within EPRI 3002024904. EPRI 3002024904 consolidates and updates guidance provided in previous EPRI reports, including EPRI 3002000079, EPRI 3002002787, EPRI 3002009993, EPRI 3002002787, and EPRI 3002009993. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). EPRI 3002024904, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments, Revision 5, August 2023. |
Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFSN-A | IFSN-A1 IFSN-A2 IFSN-A3 IFSN-A4 IFSN-A5 IFSN-A6 IFSN-A7 IFSN-A8 IFSN-A9 IFSN-A10 IFSN-A11 IFSN-A12 IFSN-A13 IFSN-A14 IFSN-A15 IFSN-A16 IFSN-A17 |
Flood Scenario Development and Characteri-zation | Development of flood scenarios and characterization of their consequences for the internal flood PRA | Flood scenarios are defined and characterized by identifying the flood source and the corresponding hazard and flooding effects, release rate, propagation pathways, flood areas impacted, SSCs impacted, and potential operator mitigation action. Guidance on the overall development and characterization of flood scenarios can be found in Section 4 of EPRI 1019194. This approach employs decision event trees, the branches of which define plausible flood scenarios and associated damage states to be addressed by the internal flood PRA. Guidance on performing flood scenario consequence analysis for the purposes of defining event tree decision branches and identified damage states can be found in Section 6 of EPRI 1019194. Such guidance includes the analysis of potential propagation pathways (e.g., Section 6.2.1) as well as the identification and evaluation of plant design features (e.g., flood alarms, water-tight doors, sump pumps), automatic actuations, and operator actions that may contain or terminate flooding (e.g., Sections 6.2.2 and 6.2.3, Appendix D). |
As indicated in Table 3-A.2.3-2 of ASME/ANS RA-S1.1, the development and characterization of flood scenarios is an iterative process that requires consideration of propagation pathways, plant design features, automatic actuations, and/or operator actions to be done in parallel. Additionally, per EPRI 1019194 and ASME/ANS RA-S1.1, analysis of propagation pathways should address not only interarea connections (e.g., drain lines, blowout panels, gaps below doors, HVAC ducts) but also their failure, whether structural (e.g., doors, walls, seals) or otherwise (e.g., backflow through drain lines involving failed check valves). As multiple factors must be considered and given that the dynamics of the analysis can rapidly change (e.g., when flood barriers fail, sump tanks overflow, etc.), calculation of flood timing impacts (e.g., times to damage SSCs) can be a resource-intensive exercise. However, as indicated by EPRI 1019194, conservative analysis (e.g., assuming the maximum expected flood associated within a flood area, not considering drainage and other protective features, etc.) can be used to identify risk-significant flood scenarios that require further refinement and use of more resource-intensive modeling approaches. As clarified by EPRI 3002010673, simplified, one-dimensional spreadsheet-based approaches to hydraulic modeling represent the state of practice and are currently used to refine flood scenario impacts, including associated timing, further. However, as the dynamic effects of flood scenarios are not well captured by such simplified approaches, caution is needed for scenarios in which the results of the internal flood PRA model are sensitive to the timing aspect of the SSC failures and other key sources of modeling uncertainty. EPRI 3002010673 further clarifies that more advanced, state-of-the-art three-dimensional fluid modeling techniques, if used, can provide valuable insights into flood scenario timing and dynamic effects but that such techniques have characteristics that restrict their practicality and range of applications. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). EPRI 3002010673, Investigation into the Use of Three-Dimensional Modeling Techniques to Assess Internal Flooding Scenarios, September 2017. |
Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. EPRI 3002010673 is an exploratory study into the use of three-dimensional modeling techniques for the purposes of understanding whether such techniques improve analysis of water propagation and yield different risk insights compared to simplified, one-dimensional analyses that represent the state of practice. However, within risk-informed applications, there is no evidence of use of these state-of-the-art techniques investigated by EPRI 3002010673. |
| Internal Flood | IFSN-A |
IFSN-A5 IFSN-A6 IFSN-A7 |
SSC Susceptibility | Identification of affected SSCs and evaluation of their susceptibility to internal-flood-related failure mechanisms | Guidance on the identification of SSCs affected by flood sources and their corresponding hazard and flooding effects can be found in Section 3.2.3 of EPRI 1019194, whereas guidance on the evaluation of SSC susceptibility to internal-flood-related failure mechanisms (e.g., submergence, spray set impingement, pipe whip, humidity, condensation, temperature, and pressure concerns) appears in Sections 6.2.4 and 6.2.5 of EPRI 1019194. | EPRI 1019194 and ASME/ANS RA-S1.1 clarify that the evaluation of SSC susceptibility may credit flooding mitigative features (e.g., spray shielding, equipment enclosure ratings for flood or spray proofing) but that test or operational data, engineering analysis, and/or expert judgment must be used to justify the conclusion that any SSC is not susceptible to damage by flooding effects. Note that since the evaluation of SSCs is highly dependent on the propagation path of the flood, the process of identifying and evaluating SSCs should be performed in concert with flood scenario development and characterization. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFEV-A | IFEV-A1 IFEV-A2 IFEV-A3 |
Initiating Event Analysis | Identification and grouping of flood-inducing initiating events | The initiating events analysis used to support the internal flood PRA includes floods that cause an initiator (via equipment failure and/or procedural shutdown) as well as those that result from an initiator (e.g., water hammer, rapid pressurization, maintenance, etc.). Guidance on the identification of internal-flood-induced initiating events is provided by Section 6.2.6 of EPRI 1019194. Additionally, once identified, flood scenarios can be combined into initiating-event groups based on guidance within EPRI 1019194 (i.e., Section 6.2.6) and ASME/ANS RA-S1.1 (i.e., Tables 3-2.4-2 and 3-A.2.4-2). |
Grouping of flood scenarios with dissimilar plant response impacts that are associated with different success criteria should be avoided. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFEV-B | IFEV-B1 IFEV-B2 IFEV-B3 IFEV-B4 IFEV-B5 |
Initiating Event Analysis | Quantification of the annual frequencies of scenarios resulting in flood-inducing initiating events | Section 5 of EPRI 1019194 provides high-level guidance on flood-induced initiating event frequency quantification. EPRI 3002024904, which is designed to be used in conjunction with the guidance in of EPRI 1019194, documents the most current approach for developing and applying pipe rupture frequencies, including related uncertainty distributions, for use in an internal flooding PRA. Additionally, the methodology presented in EPRI 3002024904 includes characterization of plant-level flooding from various causes, including pressure boundary failures, spurious fire protection system actuations, design deficiencies, and human-/maintenance-induced floods. Lastly, system-specific characteristics, such as corrosion-resistant piping within service water systems and susceptibility to flow accelerated corrosion in high energy piping systems, are addressed. |
ASME/ANS RA-S1.1 (i.e., Tables 3-2.4-3 and 3-A.2.4-3) indicates that frequencies applied should reflect plant-specific information (e.g., design, operating experience, conditions). For instance, when accounting for any aging effects, the appropriate service time for any new or replaced piping systems should be considered. EPRI 3002024904 provides guidance on considering plant-specific information and adjusting generic frequencies to address, for example, operating experience, plant aging, and reliability and integrity management programs. EPRI 3002024904 consolidates and updates guidance provided in previous EPRI reports, including EPRI 3002000079, EPRI 3002002787, EPRI 3002009993, EPRI 3002002787, and EPRI 3002009993. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). EPRI 3002024904, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments, Revision 5, August 2023. |
Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFPR-A IFPR-B |
IFPR-A1 IFPR-A2 IFPR-A3 IFPR-B1 IFPR-B2 IFPR-B3 IFPR-B4 IFPR-B5 IFPR-B6 IFPR-B7 IFPR-B8 IFPR-B9 IFPR-B10 |
Plant Response Modeling | Construction of the internal flood response model from the internal events PRA accident sequences plus sequences unique to internal flood and incorporation of internal-flood-induced initiators and equipment and operator failures | The internal events PRA accident sequences are the primary source for developing an internal-flood-specific PRA model. However, accident sequences screened out of the internal events PRA or those that represent unique challenges need to be identified. This guidance, which includes an example of how perform an integrated evaluation of flood scenarios, is provided in Section 8 of EPRI 1019194, with guidance specific to LERF in Section 8.2.2. | Internal flood and HELB events may simultaneously impact multiple structures, redundant systems, and components at a plant. Mitigation of the event may therefore require a combination of plant system responses and manual interventions not considered by accident sequence models within the internal events PRA. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFHR-A IFHR-B IFHR-C |
IFHR-A1 IFHR-A2 IFHR-B1 IFHR-B2 IFHR-B3 IFHR-C1 |
Internal Flood HRA | Adaptation of internal events HRA methods to address the impact of internal flood effects for the internal flood PRA | HFEs related to actions previously modeled in the internal events PRA may have to be modified because an internal flood may change the scenario characteristics, such as timing, cues, or specific actions that would have to be taken (e.g., due to internal-flooded pathways that affect the operator action transit routes). Additionally, new HFEs may need to be identified owing to the unique challenges posed by an internal flood. Section 7 of EPRI 1019194 provides guidance for doing so, adapting the quantification processes used for the internal events PRA. | As the definition of flood scenarios may include operator actions to isolate a flood, the model integration needs to account for possible dependencies between these internal-flood-specific recovery actions and the other operator actions modeled in the PRA. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFHR-D | IFHR-D1 | Internal Flood HRA | Identification and quantification of internal-flood-specific recovery actions | Recovery actions are new post-initiating event operator actions taken in response to an internal flood; that is, they are specific to internal floods, must address internal-flood-related effects, and are not already included as internal events HFEs. Section 7 of EPRI 1019194 provides guidance on the methodology to identify and quantify such actions within the internal flood PRA. This guidance includes the evaluation of potential errors in detection, diagnosis, and implementation of flood mitigation actions. | The likelihood of successful flood isolation is strongly correlated with the flow rates that result from a pressure boundary failure. Section 8.3.1 of EPRI 1019194 provides additional guidance for addressing this consideration given that pipe failure frequencies (e.g., in EPRI 3002010673) are either interval or threshold values. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | IFQU-A IFQU-B IFQU-C IFQU-D IFQU-E IFQU-F |
IFQU-A1 IFQU-A2 IFQU-A3 IFQU-A4 IFQU-A5 IFQU-A6 IFQU-B1 IFQU-C1 IFQU-D1 IFQU-D2 IFQU-E1 IFQU-F1 IFQU-F2 |
Internal Flood PRA Quantification | Internal flood-induced CDF and LERF quantification | Section 8 of EPRI 1019194 provides a procedure for performing internal flood risk quantification and thus integrating insights from the various analyses used to support internal flood PRA development, e.g., internal flood frequency analysis, scenario analysis, HRA, etc. Section 8.2.2 has guidance specific to LERF. | Appendix F of EPRI 1019194 provides guidance on the application of software and tools (e.g., EPRI’s FRANX) used to support quantification of the internal flood PRA. Similar guidance appears in EPRI 3002024904 regarding integration of flood frequencies. EPRI 3002024904 consolidates and updates guidance provided in previous EPRI reports, including EPRI 3002000079, EPRI 3002002787, EPRI 3002009993, EPRI 3002002787, and EPRI 3002009993. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). EPRI 3002024904, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments, Revision 5, August 2023. |
Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | Multiple | IFSO-A4 IFSN-A15 IFSN-A16 IFEV-B5 |
Screening | Qualitative and quantitative screening of flood areas, flood scenarios, and internal-flood-induced initiating events | ASME/ANS RA-S1.1, through cited SRs and associated commentary (i.e., Tables 3-A.2.2-2, 3-A.2.3-3, and 3-A.2.4-3), provides detailed criteria for performing qualitative and quantitative screening within the internal flood PRA. Guidance on qualitative screening is also provided in Section 3.4 of EPRI 1019194. Section 5.6 provides examples of qualitative and quantitative screening criteria for consideration of maintenance- induced floods. Additional guidance, including on the use of quantitative screening, is provided throughout the remainder of EPRI 1019194 (e.g., Sections 6 and 10). |
Note that the appropriateness of any given quantification or screening practice is ultimately dependent on the given application for which the internal flood PRA is quantified. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | Multiple | IFPP-B4 IFSO-A7 IFSN-A17 IFQU-A6 |
Walkdowns | Performance and documentation of plant walkdowns for the internal flood PRA | Walkdowns are typically performed to confirm the accuracy of information obtained from plant information sources and collect additional information that cannot be easily obtained from such sources. EPRI 1019194 (e.g., Section 3.3, Appendix E) provides guidance on conducting and documenting plant walkdowns performed in support of the internal flood PRA. Additional guidance is also provided by ASME/ANS RA-S1.1, through cited SRs and associated commentary (i.e., Tables 3-A.2.1-3, 3-A.2.2-2, 3-A.2.3-2, and 3-A.2.7-2). | No limitations noted | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | Multiple | IFPP-C1 IFSO-B1 IFSN-B1 IFEV-C1 IFPR-C1 IFHR-E1 IFQU-G1 IFQU-G2 IFQU-G4 |
Internal Flood PRA Documentation | Documentation of the internal flood PRA and its results | EPRI 1019194 (e.g., Section 9) provides the general practice for documenting the internal flood PRA to allow for its review and written basis. ASME/ANS RA-S1.1 also provides guidance for each relevant SR. | Note that the appropriateness of any given documentation practice is ultimately dependent on the given application to which the internal flood PRA is applied. | EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). | Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. |
| Internal Flood | Multiple | IFPP-B5 IFPP-C2 IFSO-A8 IFSO-B2 IFSN-A18 IFSN-B2 IFEV-B6 IFEV-C2 IFPR-B11 IFPR-C2 IFHR-D2 IFHR-E2 IFQU-A7 IFQU-F1 IFQU-F2 IFQU-G3 |
Uncertainty | Identification, characterization, and documentation of uncertainties in the internal flood PRA | Detailed guidance, inclusive of internal flood PRA, on treating different sources of epistemic uncertainty (i.e., parameter, model, and completeness uncertainties) are provided in NUREG-1855 as well as EPRI 1016737 and EPRI 1026511. Consistent with relevant SRs in ASME/ANS RA-S1.1, methods to assess the effects of model uncertainties and related assumptions are discussed. EPRI 1019194 (e.g., Section 8.2.3, Section 5.5) also outlines guidance for identifying uncertainties in the internal flood PRA as well as performing uncertainty and sensitivity analyses. EPRI 3002024904 addresses uncertainties associated with pipe system failure rate and flood frequency estimates. |
Note that the impact of uncertainties on the internal flood PRA and its results is ultimately dependent on the given application to which the internal flood PRA is applied. EPRI 3002024904 consolidates and updates guidance provided in previous EPRI reports, including EPRI 3002000079, EPRI 3002002787, EPRI 3002009993, EPRI 3002002787, and EPRI 3002009993. |
EPRI 1019194, Guidelines for Performance of Internal Flooding Probabilistic Risk Assessment, December 2009 (including the update and erratum dated August 5, 2020). EPRI 3002024904, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments, Revision 5, August 2023. NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017 (ML17062A466). EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012. |
Many internal flood PRAs have been developed and have undergone peer review to date. It is common practice to integrate them directly into the internal events PRA. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. NUREG-1855 provide guidance specifically developed by the NRC to address matters of uncertainty, including related modeling uncertainty. Guidance in EPRI 1016737 and EPRI 1026511, as stated within NUREG-1855, complements that of the NUREG given that they expand upon specific methods to treat uncertainty and their use in applications. |
| Internal Fires | PP-A | PP-A1 | Plant Partitioning | Selection of global plant analysis boundary for the internal fire PRA | The global plant analysis boundary for fire PRA is to be selected so that plant partitioning is performed on areas of the plant that contribute to fire risk. This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 1.5.1, Step 1. | No limitations noted | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology, September 2005 (ML15167A411). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | PP-B | PP-B1 PP-B2 PP-B3 PP-B4 PP-B5 PP-B6 PP-B7 |
Plant Partitioning | Partitioning of the plant into PAUs for the internal fire PRA | Partitioning of the plant into PAUs for the fire PRA involves gathering and characterizing fire compartment boundary information. This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 1.5.2 and 1.5.3, Step 2 and 3, NFPA 805, and PRA standard ASME/ANS RA-S1.1. | Justification of (1) non-fire-rated fire barriers, (2) active features such like water curtains, or (3) spatial separation requires assessment that considers the nature of fire sources in that location. ASME/ANS RA-S1.1, Table 4-A.2.1-3 states NUREG/CR-6850 provides criteria that may be applied in justifying decisions related to spatial separation, active fire barrier elements, and partitioning features that lack a fire-resistance rating. ASME/ANS RA-S1.1, SR PP-B3 identifies certain features that should not be credited. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NFPA 805, Performance-Based Standard for Fire Protection for LWR Reactor Electric Generating Plants, 2001. (Newer version available, but not endorsed for 10 CFR 50.48(c) transition). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. 10 CFR 50.48(c) establishes the requirements for using NFPA 805 as an alternative to the requirements associated with 10 CFR 50.48(b) (Fire Protection covered by Appendix R). |
| Internal Fires | ES-A | ES-A1 ES-A2 ES-A3 ES-A4 ES-A5 ES-A6 ES-A7 |
Equipment Selection | Identification of fire-induced initiating events for the internal fire PRA | Identification of fire-induced initiating events and the equipment whose failure, including spurious operation, would cause an automatic or manual trip, or plant shutdown. This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 2.5 (particularly, Chapter 2.5.3, Step 3) and NEI 00-01, Section 4. | While internal event accident sequences should be considered for applicability to the fire PRA, including ISLOCAs, fires do not normally initiate certain kinds of accidents, such as pipe breaks. Scenarios and equipment associated with spurious operations, including MSOs, should be included. ASME/ANS RA-S1.1, Tables 4-A.2.2-2 and 4-A.2.2-3 cite NEI 00-01 (see Section 4) as a source of an acceptable method for identifying MSO combinations. NEI 04-02, Section B.2.1, also provides guidance for identifying and screening MSOs. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NEI 00-01, Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 4, December 2019 (ML19351D276). NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 3, November 2019 (ML19351D277). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. In addition, RG 1.205 endorses portions of NEI 00-01, and ASME/ANS RA-S1.1 cites NEI 00-01 as an acceptable method for identifying MSO combinations. RG 1.205 also endorses NEI 04-02. |
| Internal Fires | ES-B | ES-B1 ES-B2 ES-B3 |
Equipment Selection | Identification of fire-induced mitigating system failure events for the internal fire PRA | Identification of equipment whose failure, including spurious operation, would compromise mitigating systems that are included in the internal fire PRA. This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 2.5 (particularly, Chapter 2.5.3, Step 4) and NEI 00-01, Section 4. | Scenarios and equipment associated with spurious operations, including MSOs, should be included. ASME/ANS RA-S1.1, Table 4-A.2.2-3 cites NEI 00-01 (see Section 4) as a source of an acceptable method for identifying MSO combinations. NEI 04-02, Section B.2.1, also provides guidance for identifying and screening MSOs. For fire scenarios leading to MCRA, NUREG-1921, Supplement 1, provides guidance related to equipment selection, including with regard to the capability of RSDPs, if available. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NEI 00-01, Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 4, December 2019 (ML19351D276). NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 3, November 2019 (ML19351D277). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines – Qualitative Analysis for Main Control Room Abandonment Scenarios, Supplement 1, January 2020 (ML20035E043). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. In addition, RG 1.205 endorses portions of NEI 00-01, and ASME/ANS RA-S1.1 cites NEI 00-01 as an acceptable method for identifying MSO combinations. RG 1.205 also endorses NEI 04-02. NUREG-1921, Supplement 1, provides guidance specifically developed by the NRC and EPRI for MCRA scenarios in fire PRA. |
| Internal Fires | ES-C | ES-C1 ES-C2 |
Equipment Selection | Identification of fire-induced instrumentation failures that impact operator actions credited in the internal fire PRA | Identification of instrumentation whose failure, including spurious operation, would impact the reliability of operator actions credited in the fire PRA. This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 2.5 (particularly, Chapter 2.5.3, Step 5) and NEI 00-01, Section 4. | Instruments that support operator actions that are credited in the fire PRA (or actions, that if taken, would fail a credited plant function) could be explicitly modeled in the fire PRA but may not be. For fire scenarios leading to MCRA, NUREG-1921, Supplements 1 and 2, provides guidance related to instrumentation, including concerning the capability of RSDPs, if available. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NEI 00-01, Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 4, December 2019 (ML19351D276). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines – Qualitative Analysis for Main Control Room Abandonment Scenarios, Supplement 1, January 2020 (ML20035E043). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines—Quantification Guidance for Main Control Room Abandonment Scenarios, Supplement 2, June 2019 (ML19162A378). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. In addition, RG 1.205 endorses portions of NEI 00-01, and ASME/ANS RA-S1.1 cites NEI 00-01 as an acceptable method for identifying MSO combinations. NUREG-1921, Supplements 1 and 2, provides guidance specifically developed by the NRC and EPRI for MCRA scenarios in fire PRA. |
| Internal Fires | CS-A | CS-A1 CS-A2 CS-A3 CS-A4 |
Cable Selection | Identification and location of plant cables whose failure can affect equipment or functions included in the internal fire PRA | Identification and location of plant cables whose fire-induced failure can affect equipment or functions included in the fire PRA. Fundamental guidance is provided in NUREG/CR-6850, Volume 2, Chapter 3.5, Steps 1, 2, and 3. Updated guidance is provided in NUREG/CR-7150, Volumes 1, 2, and 3, and NEI 00-01, Chapter 3. |
Cables associated with possible MSOs should be included. To meet CS-A1 CC-II, cables must be identified and for risk-significant contributors must be associated with equipment failure modes, therefore requiring simple failure mode analysis. ASME/ANS RA-S1.1, Table 4-A.2.3-2 states that Chapter 3 of NEI-00-01 provides an acceptable method for circuit failure analysis for the internal fire PRA. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-7150, JACQUE-FIRE – Volume 1: PIRT Exercise for NPP Fire-Induced Electrical Circuit Failure, October 2012 (ML12313A105). NUREG/CR-7150, JACQUE-FIRE – Volume 2: Expert Elicitation Exercise for NPP Fire-Induced Electrical Circuit Failure, May 2014 (ML14141A129). NUREG/CR-7150, JACQUE-FIRE – Volume 3: Technical Resolution to Open Issues on NPP Fire-Induced Electrical Circuit Failure, November 2017 (ML17331B098). NEI 00-01, Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 4, December 2019 (ML19351D276). |
RG 1.205 (ML21048A448) references the methodologies documented in NUREG/CR-6850 and NUREG/CR-7150, which are widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. In addition, RG 1.205 endorses portions of NEI 00-01, and ASME/ANS RA-S1.1 cites NEI 00-01 as an acceptable method for performing circuit failure analysis. |
| Internal Fires | CS-A | CS-A1 | Cable Selection | Circuit failure analysis for internal fire PRA | To understand which and how equipment functions are affected by fire-induced failure of associated cables, circuit failure analysis is required for risk-significant contributors. Fundamental guidance is provided in NUREG/CR-6850, Volume 2, Chapter 3.5, Chapter 3.5 Steps 1, 2, and 3. Updated guidance is provided in NUREG/CR-7150, Volumes 1, 2, and 3, and NEI 00-01, Chapter 3. |
To meet CS-A1 CC-II, cables must be identified for risk-significant contributors and associated with equipment failure modes, therefore requiring simple failure mode analysis. ASME/ANS RA-S1.1, Table 4-A.2.3-2 states that Chapter 3 of NEI-00-01 provides an acceptable method for circuit failure analysis for the internal fire PRA. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-7150, JACQUE-FIRE – Volume 1: PIRT Exercise for NPP Fire-Induced Electrical Circuit Failure, October 2012 (ML12313A105). NUREG/CR-7150, JACQUE-FIRE – Volume 2: Expert Elicitation Exercise for NPP Fire-Induced Electrical Circuit Failure, May 2014 (ML14141A129). NUREG/CR-7150, JACQUE-FIRE – Volume 3: Technical Resolution to Open Issues on NPP Fire-Induced Electrical Circuit Failure, November 2017 (ML17331B098). NEI 00-01, Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 4, December 2019 (ML19351D276). |
RG 1.205 (ML21048A448) references the methodologies documented in NUREG/CR-6850 and NUREG/CR-7150, which are widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. In addition, RG 1.205 endorses portions of NEI 00-01 and ASME/ANS RA-S1.1 cites NEI 00-01 as an acceptable method for performing circuit failure analysis. |
| Internal Fires | CS-B | CS-B1 CS-B2 |
Cable Selection | Electrical overcurrent protective device coordination for distribution circuits in the internal fire PRA | Regarding fire-induced electrical overcurrent protective device coordination for distribution circuits in the fire PRA, guidance is provided in NUREG/CR-6850, Volume 2, Chapters 2.5, 3.5.3, and 3.5.4 (Steps 4 and 5). | Circuit coordination for circuits in the PRA not already evaluated as part of Appendix R has to be performed. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | QLS-A | QLS-A1 QLS-A2 QLS-A3 QLS-A4 |
Qualitative Screening | Screen PAUs from the fire PRA without quantitative analysis | Qualitatively screen PAUs from the fire PRA without quantitative analysis by deterministic screening based on no impact to the plant or impact that is subsumed into a more frequent or more impactful event. This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 4. | The screening is non-quantitative. ASME/ANS RA-S1.1, SR QLS-A3 refers to SCR-3 of the generic screening criteria within Table 1-1.8-1, which refers to a deterministic screening based on no Impact to the plant or impact that is subsumed into a more frequent or more impactful event. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | PRM-A PRB-B |
PRM-A1 PRM-A2 PRM-A3 PRM-B1 PRM-B2 PRM-B3 PRM-B4 PRM-B5 PRM-B6 PRM-B7 PRM-B8 PRM-B9 PRM-B10 PRM-B11 PRM-B12 PRM-B13 PRM-B14 PRM-B15 |
Plant Response Modeling | Construction of the fire plant response model from the internal events PRA accident sequences plus sequences unique to fire and incorporation of fire-induced initiators and equipment and operator failures | The internal events PRA accident sequences are the primary source for developing a fire-specific PRA model. However, accident sequences screened out of the internal events PRA or those that represent unique challenges need to be identified (e.g., such as those involving spurious operations). This guidance is provided in NUREG/CR-6850, Volume 2, Chapter 5.5. | Significant additional modeling is required to account for accident scenarios that involve abandonment from the control room because of fire and shutdown from an alternate shutdown location. NUREG-1921, Supplements 1 and 2, provides guidance on how to model abandonment from the MCR necessitated by loss of habitability or control due to fire leading to shutdown from an alternate shutdown location. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines – Qualitative Analysis for Main Control Room Abandonment Scenarios, Supplement 1, January 2020 (ML20035E043). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines – Quantification Guidance for Main Control Room Abandonment Scenarios, Supplement 2, June 2019 (ML19162A378). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1921, Supplements 1 and 2, provides guidance specifically developed by the NRC and EPRI for MCRA scenarios in fire PRA. |
| Internal Fires | FSS-A | FSS-A1 FSS-A2 FSS-A3 FSS-A4 |
Ignition Source and Damage Target Set Selection | Selection of sufficient combinations of ignition sources (or groups thereof) and damage target sets that represent the fire scenarios for each PAU that has not been screened out and upon which the estimates of the risk contribution are based | The number of individual fire scenarios and the level of detail included in the analysis of each scenario is commensurate with the relative risk importance of the PAU under analysis with the more risk-significant PAUs likely characterized by detailed analysis of multiple and/or more specific fire scenarios. Guidance related to the identification of and the level of analysis for ignition sources that may impact the fire risk is provided in NUREG/CR-6850, Volume 2, Chapters 8 and 11. This guidance also covers the identification and characterization of damage target sets, including secondary combustibles. NUREG-1934 provides additional guidance on the selection and definition of fire scenarios within the fire modeling process. |
The selection and characterization of the fire scenarios that support the fire PRA is a highly iterative process, and any determination regarding the sufficiency of selected combinations and resulting level of analysis is a function of the application that the fire PRA supports. This includes the appropriateness of any conservative or bounding analyses. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1934 provides guidance specifically developed by the NRC and EPRI to perform fire modeling analysis in support of fire PRA and related applications. |
| Internal Fires | FSS-B | FSS-B1 FSS-B2 |
MCRA Fire Scenarios | Fire scenarios requiring a transfer of primary command and control outside of the MCR | The decision to abandon the MCR are based on specific conditions, including MCR habitability issues and loss of MCR control functions. Guidance on the selection and characterization of MCRA fire scenarios, including abandonment criteria, is provided in Chapter 11 of NUREG/CR-6850, Volume 2. Guidance for analyzing the human and plant response to such scenarios, including detailed consideration of the decision to abandon, is greatly expanded by NUREG-1921, Supplement 1. | The success of performing shutdown from outside of the MCR is dependent on several factors, including the plant strategy and supporting procedures, capabilities of the RSDP, and the number of local operator actions. Proper treatment of MCRA scenarios thus requires even closer cooperation between multiple fire PRA disciplines, including fire scenario development, plant response modeling, and HRA. Guidance on task interfaces and interactions to address the additional needs of MCRA scenarios is provided by NUREG-1921, Supplement 1. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines – Qualitative Analysis for Main Control Room Abandonment Scenarios, Supplement 1, January 2020 (ML20035E043). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1921, Supplement 1 provides guidance specifically developed by the NRC and EPRI for MCRA scenarios in fire PRA. |
| Internal Fires | FSS-C | FSS-C1 FSS-C2 FSS-C3 FSS-C4 FSS-G1 |
Timing and Extent of Fire Damage | Characterization of factors that influence the timing and extent of fire damage of selected fire scenarios | Ignition sources and secondary combustibles, whether treated probabilistically or not, are characterized by several factors, including source type, location, intensity, and duration. Section 8 of NUREG/CR-6850, Volume 2, provides guidance on characterizing ignition sources and their threat to targets via ZOIs based on the conservative or bounding conditions (e.g., peak HRRs) as well as through the application of severity factors, which are outlined in Appendix E to NUREG/CR-6850, Volume 2. Section 11 of NUREG/CR-6850, Volume 2, expands this guidance further by presenting more detailed and refined treatments for fire scenario analysis, including addressing fire growth and spread. ASME/ANS RA-S1.1, Table 4-A.2.6-8 further clarifies that additional phenomena associated with multicompartment fire scenarios, beyond those associated with fire scenarios of single PAUs, must also be considered (e.g., hot gas flow through openings and ducts from the PAU of fire origin). Such guidance appears in Section 11.5.4 of NUREG/CR-6850, Volume 2. In addition to the guidance within NUREG/CR-6850, factors that influence the timing and extent of fire damage, particularly HRR distributions, have been subsequently informed by significant industry research efforts, the outcomes of which are clarified, by ignition source, within the adjacent column. Lastly, note that NUREG-1934 provides additional detailed guidance and examples on the implementation of the fire modeling process, addressing those factors that influence the timing and extent of fire damage. |
Specific guidance on those factors affecting the timing and extent of fire damage for the different ignition source types is documented not only in NUREG/CR-6850, Volume 2, but also throughout various technical reports, supplements, and updates. Appendix G to NUREG/CR-6850, Volume 2, provides guidance on defining the HRR profiles for electrical cabinets, transient combustible materials, and flammable/combustible liquids. Regarding the treatment of transient fires, NUREG-2233 provides the most current and complete guidance, including on HRR distributions and other fire characteristics as well as definition and treatment of transient combustible control locations. In particular, this technical report documents an updated transient fire distribution, intended as an improved realism replacement for the more conservative one in Appendix G of NUREG/CR-6850, Volume 2, and defines an additional distribution for transient combustible control locations. Moreover, NUREG-2233 provides a comprehensive set of fire growth and decay guidance that improves upon the transient fire growth methodology in FAQ 08-0052 (NUREG/CR-6850 Supplement 1, Section 17). Lastly, note that the EPRI-led Fire PRA Methods Review Panel clarified specific guidance in Section G.5 of NUREG/CR-6850 regarding the technical basis for transient fire HRRs applied within the fire PRA, but the latest guidance on this topic appears in NUREG-2233. NUREG-2178, Volume 2, provides revised HRR distributions for electric motors and dry transformers. Appendix R to NUREG/CR-6850, Volume 2, provides guidance on calculating the flame spread and HRR from cable fires. FAQ 08-0049 (NUREG/CR-6850 Supplement 1, Section 11) provides further guidance on cable tray fires, including applicability of the cable fire parameters in Appendix R of NUREG/CR-6850, Volume 2. Guidance on HRRs, ignition, and spread rates is further supplemented by the information in NUREG/CR-7010, Volume 1. Supplemental guidance particular to the treatment of self-ignited cable fires and cable fires caused by welding and cutting is provided by FAQ 13-0005, which provides an alternate but more realistic approach to that within NUREG/CR-6850, Volume 2, for the detailed modeling of these types of fire events. For electrical enclosures, Appendix E to NUREG/CR-6850, Volume 2, provides distributions of peak HRR values as a function of the types (qualified vs. unqualified) and amounts (single vs. multiple bundles) of cables inside vertical electrical cabinets. NUREG-2178, Volume 1, provides more refined electrical enclosure classifications and corresponding peak HRR distributions; though, the existing HRR distributions in Appendix E and G to NUREG/CR-6850, Volume 2, are considered bounding by comparison. Also, NUREG-2230 supplements guidance in Appendix G to NUREG/CR-6850, Volume 2, regarding fire growth of electrical cabinet fires by providing HRR timing profiles for interruptible fires. Lastly, note that FAQ 08-0043 (NUREG/CR-6850 Supplement 1, Section 12) clarifies assumptions regarding the location of fires within electrical cabinets and that guidance particular to the treatment of junction box scenarios is provided by FAQ 13-0006. Appendix S to NUREG/CR-6850, Volume 2, provides guidance on addressing fire propagation between electrical cabinets. NUREG-2178, Volume 2, expands upon this guidance and develops a detailed method for determining the thermal radiation impact from fires inside electrical cabinets. FAQ 08-0042 (NUREG/CR-6850 Supplement 1, Section 8) provides further guidance on fire propagation from well-sealed, robustly secured electrical cabinets, whereas FAQ 14-0009 clarifies the treatment of fire propagation specific to well-sealed, robustly secured MCCs. Appendix L to NUREG/CR-6850, Volume 2, as supplemented by FAQ 06-0018 (NUREG/CR-6850 Supplement 1, Section 5) and FAQ 14-0008, addresses the treatment of MCB fire scenarios. NUREG-2178, Volume 2, provides supplemental guidance on MCB fire scenarios and presents an alternate approach to Appendix L to NUREG/CR-6850. Appendix M to NUREG/CR-6850, Volume 2, as supplemented by FAQ 06-0017 (NUREG/CR-6850 Supplement 1, Section 4), FAQ 07-0035 (NUREG/CR-6850 Supplement 1, Section 7), and NUREG-2262, provides guidance on the treatment of HEAFs, including their ZOI. Guidance on hydrogen fires and turbine generator fires is provided in Appendix N and O, respectively, to NUREG/CR-6850, Volume 2. FAQ 08-0044 (NUREG/CR-6850 Supplement 1, Section 9) revises guidance in NUREG/CR-6850, Volume 2, on the characterization of MFW pump oil spill fires. Additional guidance on the treatment of other types of pump oil fires was also developed by the EPRI-led Fire PRA Methods Review Panel. Lastly, for fixed and transient ignition sources, NUREG-2178, Volume 2, provides guidance on location factors when fires are postulated in a corner or along a wall. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). NUREG-2178, Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume, April 2016 (ML16110A140). NUREG-2178, Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE), Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric motors, Indoor Dry Transformers, and the Main Control Board, June 2020 (ML20168A655). NUREG-2230, Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants, June 2020 (ML20157A148). NUREG-2233, Methodology for Modeling Transient Fires in Nuclear Power Plant Fire Probabilistic Risk Assessment, October 2020 (ML20289A568). NUREG-2262, High Energy Arcing Fault Frequency and Consequence Modeling, April 2023 (ML23108A113). NUREG-7010, Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE), Phase 1: Horizontal Trays (Volume 1), July 2012 (ML12213A056). FAQ 13-0005, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0005, on Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting, December 2013 (ML13319B181). FAQ 13-0006, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0006 on Modeling Junction Box Scenarios in a Fire PRA, December 2013 (ML13331B213). FAQ 14-0008, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0008, on Main Control Board Treatment, July 2014 (ML14190B307). FAQ 14-0009, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0009, on Treatment of Well-Sealed MCC Electrical Panels Greater than 440V, April 2015 (ML15114A441). NRC-to-NEI Letter, Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, “Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires,” June 2012 (ML12171A583). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850 (including Supplement 1), which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. The guidance provided in the closeout memoranda to FAQs 13-0005, 13-0006, 14-0008, and 14-0009 is acceptable for use by licensees. NUREG-1934, NUREG-2178 (Volumes 1 and 2), NUREG-2230, NUREG-2233, NUREG-2262, and NUREG-7010 provide guidance specifically developed by the NRC and EPRI to expand upon that found in NUREG/CR-6850 and closeout memoranda to FAQs, with either supplemental guidance and/or alternate approaches. Guidance on referenced methods developed by the EPRI-led Fire PRA Methods Review Panel was endorsed by the NRC staff with clarification. |
| Internal Fires | FSS-C | FSS-C5 FSS-C6 |
Target Damage Criteria and Thermal Response | Characterization of target damage and calculation of time to damage | Damage and/or ignition criteria, including approaches for time to damage for targets typically considered in NPP fire scenarios, are reviewed in Section 8 (including Section 8.5.1.2 and Table 8-2) as well as Appendix H to NUREG/CR-6850, Volume 2. For cable fires, additional guidance is provided in Appendix R to NUREG/CR-6850, Volume 2, as clarified by FAQ 08-0049 (NUREG/CR-6850 Supplement 1, Section 11). | In addition to those methods within NUREG/CR-6850, Volume 2, NUREG/CR-6931 and NUREG-1805, Supplement 1, document the THIEF Model, which provides an alternate approach to assess the damage to electrical cables. FAQ 16-0011 clarifies the guidance in NUREG/CR-6850 associated with damage and ignition of cables subjected to fire generated conditions in order to provide a more realistic characterization of fire propagation in stacks of cable trays. NUREG-2178, Volume 2, provides expanded guidance for evaluating the time to damage for generic cables exposed to a time-dependent temperature or heat flux (i.e., the Heat Soak Method). FAQ 08-0053 and NUREG/CR-7102 provide additional guidance, including damage criteria, for Kerite FR cables. FAQ 13-0004 provides supplemental guidance for application of the damage criteria provided in NUREG/CR-6850, Volume 2, for sensitive electronics. Note that methods discussed elsewhere and applied to assess the timing and extent of fire damage in fire scenarios (e.g., those used to address propagation between electrical cabinets or across the MCB) inherently reflect some treatment of and assumptions related to target damage. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). NUREG-1805, Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program – Supplement 1, Volumes 1 and 2, July 2013 (ML13211A097 and ML13211A098). NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire, Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board, June 2020 (ML20168A655). NUREG/CR-6931, Cable Response to Live Fire (CAROLFIRE) Volume 3 : Thermally Induced Electrical Failure (THIEF) Mode, April 2008 (ML081190261). NUREG/CR-7102, Kerite Analysis in Thermal Environment of FIRE (KATE-Fire): Test Results – Final Report, December 2011 (ML11333A033). FAQ 08-0053, Close-Out of NFPA 805 FAQ 08-0053, Revision 1 on Kerite FR Cable Failure Thresholds, June 2012 (ML121440155). FAQ 13-0004, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0004, Revision 1, on Clarifications regarding Treatment of Sensitive Electronics, December 2013 (ML13322A085). FAQ 16-0011, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 16-0011, on Alternative Methodology to NUREG/CR-6850 for Bulk Cable Tray Ignition Criteria, March 2018 (ML18074A020). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850 (including Supplement 1), which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. The guidance provided in the closeout memoranda to FAQs 13-0004 and 16-0011 is acceptable for use by licensees. NUREG-1805 (Supplement 1), NUREG-2178 (Volume 2), NUREG/CR-6931, and NUREG/CR-7102 provide guidance specifically developed by the NRC and EPRI to expand upon that found in NUREG/CR-6850 and closeout memoranda to FAQs, with either supplemental guidance and/or alternate approaches. |
| Internal Fires | FSS-C | FSS-C7 | Fire Barrier Elements within PAUs High-Hazard Ignition Sources | Crediting fire barrier elements in the analysis of fire scenarios within a single PAU to, e.g., limit fire damage or delay the spread of fire or the onset of fire damage | ASME/ANS RA-S1.1, Table 4-A.2.6-4 states that fire barriers may include passive barriers (e.g., non-rated partition walls, cable wraps, or radiant energy shields) or active barriers (e.g., normally open fire doors or water curtains), provided there is a justifiable basis for doing so and failure of credited elements is contemplated. General guidance on barriers and partitions appears in Section 1 of NUREG/CR-6850, Volume 2. Guidance on the treatment of passive barriers appears within Appendix Q to NUREG/CR-6850, Volume 2, whereas guidance on the treatment of active features, such as water curtains, as well as dependency between multiple features, is in Appendix P. Sections 8 and 11 to NUREG/CR-6850, Volume 2, as well as NUREG-1934 provide guidance on crediting fire barrier elements within the context of fire scenario analysis to limit fire damage or delay the spread of fire or the onset of fire damage. |
ASME/ANS RA-S1.1, Table 4-A.2.6-4 distinguishes between the analysis of fire scenarios impacting adjacent PAUs (the multicompartment fire analysis), which is addressed by HLR-FSS-G and its SRs, and the analysis addressed here by FSS-C7, i.e., for those cases where barriers exist within a single PAU (i.e., the barriers exist but were not credited during plant partitioning). Also, ASME/ANS RA-S1.1, Table 4-A.2.6-4 provides guidance that credited barriers should not be subjected to damage from a high-hazard ignition source unless it has been subject to qualification or other proof of performance by analysis or testing under such conditions. Such sources would include HEAFs as well as large quantities of flammable liquid or hydrogen gas. Note that for fire PRAs supporting applications, such as NFPA 805, there is much technical analysis and guidance (inclusive of applicable codes, standards, research, and testing) used to support compliance with the fundamental fire protection program and design elements established by NFPA 805, which includes deterministic assessment of fire barrier elements and could form, in part, the basis of a more probabilistic treatment within a fire PRA. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001. (Newer version available, but not endorsed for 10 CFR 50.48(c) transition) |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850 (including Supplement 1), which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1934 provides guidance specifically developed by the NRC and EPRI to update guidance provided in NUREG/CR-6850. |
| Internal Fires | FSS-D | FSS-D1 FSS-D2 FSS-D3 FSS-D4 FSS-D5 FSS-D6 FSS-D7 FSS-D8 FSS-D9 FSS-D10 |
Fire Analysis Tools | Use of and basis for analysis tools applied to fire scenarios | Fire analysis tools, which may be empirical, analytical, or statistical, are only to be used within known limits of applicability and to address conditions for which the tools have sufficient capability to model. Sections 8 and 11 of NUREG/CR-6850, Volume 2, provide guidance on the scoping and detailed fire modeling of fire scenarios, respectively, and address the selection and use of fire analysis tools. NUREG-1934 expands upon this guidance, detailing further the proper application of fire models to fire scenarios and outlining a step-by-step process for using fire modeling in NPP applications. This includes guidance on how to define fire modeling goals, characterize fire scenarios, select fire analysis tools, calculate fire-generated conditions, conduct sensitivity and uncertainty analyses, and document the results. NUREG-1824, Supplement 1, provides the most current guidance on the V&V of fire analysis tools and specifically evaluates five fire models that have been used for NPP applications. These include: the empirical correlations within the NRC’s Fire Dynamics Tools (as documented within NUREG-1805 and its supplement) and EPRI’s Fire-Induced Vulnerability Evaluation (FIVE); the zone models Consolidated Model of Fire Growth and Smoke Transport (CFAST) and MAGIC; and the CFD model Fire Dynamics Simulator (FDS). While the original NUREG-1824 remains valid for the versions of the fire models for which the V&V documented in the report was conducted, NUREG-1824, Supplement 1, expands on this earlier effort by evaluating newer versions of the fire models, including additional test data to support validation, and providing other enhancements (e.g., more quantifiable assessment of model accuracy, additional model output quantities, etc.). Lastly, the supplement was designed to complement NUREG-1934, which also contains technical V&V guidance, including how to use the results of the validation study in typical NPP fire modeling analyses. |
There are several additional empirical and statistical rule sets used to model fire behavior (including intensity and spread) within a fire PRA beyond those documented in NUREG-1805 and its supplement. For cable fires, Appendix R to NUREG/CR-6850, Volume 2, as supplemented by the information and the FLASH-CAT Model documented in NUREG/CR-7010, Volume 1, addresses the growth and spread of fire within vertical stacks of cable trays. Additionally, the THIEF model, which is used to estimate temperatures within an electrical cable, is documented within NUREG/CR-6931, Volume 3, and Supplement 1 to NUREG-1805. Note that treatment of self-ignited cable fires and cable fires caused by welding and cutting is addressed by FAQ 13-0005. For both generic transient fires as well as those in transient combustible control locations, NUREG-2233 develops probabilistic distributions of ZOI that enable the screening of targets without the need to separately calculate a ZOI through fire modeling. Values are included for vertical, vertical-in-a-corner, and horizontal ZOIs for a variety of cases, including exposed sensitive electronics, thermoplastic cables, Kerite-FR cables, thermoset cables, and bulk cable/tray ignition. Appendix S to NUREG/CR-6850, Volume 2, provides guidance on addressing fire propagation between electrical cabinets. NUREG-2178, Volume 2, expands upon this guidance. FAQ 08-0042 (NUREG/CR-6850 Supplement 1, Section 8) provides further guidance on fire propagation from well-sealed, robustly secured electrical cabinets, whereas FAQ 14-0009 clarifies the treatment of fire propagation from well-sealed, robustly secured MCCs. Treatment of junction box scenarios is provided by FAQ 13-0006. Appendix L to NUREG/CR-6850, Volume 2, as supplemented by FAQ 06-0018 (NUREG/CR-6850 Supplement 1, Section 5) and FAQ 14-0008, addresses treatment of MCB fire scenarios. NUREG-2178, Volume 2, provides supplemental guidance on MCB fire scenarios and presents an alternate approach to Appendix L to NUREG/CR-6850. Appendix M to NUREG/CR-6850, Volume 2, as supplemented by FAQ 06-0017 (NUREG/CR-6850 Supplement 1, Section 4), FAQ 07-0035 (NUREG/CR-6850 Supplement 1, Section 7), and NUREG-2262, provides guidance on the treatment of HEAFs. Guidance on hydrogen fires and turbine generator fires is provided in Appendix N and O, respectively, to NUREG/CR-6850, Volume 2. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). NUREG-1805, Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program – Final Report, December 2004 (ML043290075). NUREG-1805, Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program – Supplement 1, Volumes 1 and 2, July 2013 (ML13211A097 and ML13211A098). NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, May 2007 (ML071730543). Supplement 1 to NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Supplement 1, Final Report, November 2016 (ML16309A011). NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire, Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board, June 2020 (ML20168A655). NUREG-2233, Methodology for Modeling Transient Fires in Nuclear Power Plant Fire Probabilistic Risk Assessment, October 2020 (ML20289A568). NUREG-2262, High Energy Arcing Fault Frequency and Consequence Modeling, April 2023 (ML23108A113). NUREG/CR-6931, Cable Response to Live Fire (CAROLFIRE), Volume 3: Thermally Induced Electrical Failure (THIEF) Model, April 2008. (ML081190261). NUREG/CR-7010, Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE), Phase 1: Horizontal Trays (Volume 1), July 2012 (ML12213A056). FAQ 13-0005, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0005, Revision 5 on Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting, December 2013 (ML13319B181). FAQ 13-0006, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0006 on Modeling Junction Box Scenarios in a Fire PRA, December 2013 (ML13331B213). FAQ 14-0008, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0008, on Main Control Board Treatment, July 2014 (ML14190B307). FAQ 14-0009, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0009, on Treatment of Well-Sealed MCC Electrical Panels Greater than 440V, April 2015 (ML15114A441). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850 (including Supplement 1), which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. The guidance provided in the closeout memoranda to FAQs 13-0005, 13-0006, 14-0008, and 14-0009 is acceptable for use by licensees. NUREG-1805 (and its supplement), NUREG-1824 (and its supplement), NUREG-1934, NUREG-2178 (Volume 2), NUREG-2233, NUREG-2262, NUREG-6931, and NUREG/CR-7010 provide guidance specifically developed by the NRC and EPRI to expand upon that found in NUREG/CR-6850 and closeout memoranda to FAQs, with either supplemental guidance and/or alternate approaches. |
| Internal Fires | FSS-D | FSS-D8 | Smoke Damage | Evaluate the potential for smoke damage to fire PRA equipment on a qualitative basis | Appendix T to NUREG/CR-6850, Volume 2, provides guidance on which equipment may be vulnerable to smoke damage and how to address it within the fire PRA. | The current state of knowledge cannot support detailed quantitative assessments of smoke damage as a part of a fire PRA. Fire scenarios that assume widespread damage (e.g., damage across an entire PAU) will generally include potential smoke damage within the limits of the assumed fire damage. For other scenarios, qualitative assessments can be factored into a quantitative fire PRA analysis as component failures are assumed in addition to thermally induced damage. Given this approach, the fire PRA results will retain considerable uncertainty relative to this topic. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | FSS-E | FSS-E1 FSS-E2 FSS-E3 FSS-E4 |
Fire Detection and Suppression | Application of conditional probabilities of target damage given fire ignition | ASME/ANS RA-S1.1, Table 4-A.2.6-6, states that current PRA practice involves the application of NSPs, i.e., the probability that suppression efforts, manual or otherwise, fail to suppress the fire before the onset of the postulated damage. They include an assessment of both effectiveness as well as an overall assessment of detection and suppression system unavailability. The latter is intended to represent functional performance of the system, including unreliability. Appendix P to NUREG-6850, Volume 2, as supplemented by FAQ 08-0050 (NUREG/CR-6850 Supplement 1, Section 14), provides guidance on the analysis of detection and suppression. This includes both the overall development of NSPs, which makes use of a detection-suppression event tree approach and associated aspects, e.g., manual fire suppression curves, generic unreliability values for fixed detection and suppression systems, etc. Using the revised guidance within FAQ 08-0050 (NUREG/CR-6850 Supplement 1, Section 14) for developing manual fire suppression curves, NUREG-2169 updates the curves, documented in FAQ 08-0050 with more recent fire event data. These curves have since been further updated by NUREG-2178, Volume 2, NUREG-2230, and NUREG-2262, as clarified in the adjacent column, for select ignition sources. Regarding VEWFD and conventional spot-type smoke detection, NUREG-2180 provides guidance on the evaluation of system performance, operating experience, and fire PRA quantification for applications in NPPs where these systems are expected to detect fires in their incipient (pre-flaming) stage. Treatment of detection and suppression within the fire scenario analysis is addressed by guidance in Chapter 11 to NUREG-6850, Volume 2, and NUREG-1934. |
ASME/ANS RA-S1.1, Table 4-A.2.6-6, clarifies that the statistical treatment of manual fire suppression is typically complementary to the events included when fire frequency is estimated; as a result, the two factors are typically highly dependent; thus, use of manual NSPs, including supporting data, must be consistent with the corresponding fire-ignition frequency values applied. Additional guidance on maintaining this consistency can be found in the updated close-out of FAQ 08-0048. NUREG-2230 revises the detection-suppression event tree described in Appendix P of NUREG/CR-6850 for electrical cabinet fires to include paths for interruptible and growing fire split fractions, the probability of no automatic smoke detection, early intervention by plant personnel, success of MCR indication, MCR operator response HEP, etc. Additionally, this guidance develops new quantitative parameters to facilitate these revisions. Moreover, NUREG-2230 provides updates to the manual fire suppression curves in NUREG-2169 for: electrical cabinets (Bin 15) applicable to interruptible and growing electrical cabinet fire scenarios; other non-cabinet electrical ignition sources (e.g., motors, pumps, and transformers); and for MCR fires. For MCR fires, a revised NSP floor value is also provided. Detailed information on the suppression rate for MCR fires can be found in NUREG-2178, Volume 2. Consistent with the guidance in FAQ 17-0013 regarding the definition of non-suppression time, NUREG-2262 provides an update to the manual fire suppression curve for HEAFs in FAQ 17-0013, which itself updated the overly conservative curve in NUREG-2169. FAQ 08-0052 (NUREG/CR-6850 Supplement 1, Section 17) provides further guidance on treatment of transient fires regarding manual suppression within the MCR. Relative to MCB fire scenarios, Appendix L to NUREG/CR-6850, Volume 2, as supplemented by FAQ 06-0018 (NUREG/CR-6850 Supplement 1, Section 5) and FAQ 14-0008, provides guidance on the treatment of such scenarios, including detection and suppression. NUREG-2178, Volume 2, provides supplemental guidance on MCB fire scenarios and presents an alternate approach to Appendix L to NUREG/CR-6850, including a new detection-suppression event tree. Note that interim guidance on incipient fire detection systems within FAQ 08-0046 (NUREG/CR-6850 Supplement 1, Section 13), was retired considering the release of NUREG-2180. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database, January 2015 (ML15016A069). NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire, Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board, June 2020 (ML20168A655). NUREG-2180, Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE), December 2016 (ML16343A058). NUREG-2230, Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants, June 2020 (ML20157A148). NUREG-2262, High Energy Arcing Fault Frequency and Consequence Modeling, April 2023 (ML23108A113). FAQ 08-0046, Response to July 28, 2016, Letter regarding Retirement of NFPA 805 FAQ 08-0046 on “Incipient Fire Detection Systems,” November 2018 (ML16253A111). FAQ 08-0048, Update of Close-Out of National Fire Protection Association Frequently Asked Question 08-0048, Fire Ignition Frequencies, May 2015 (ML15134A046). FAQ 17-0013, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 17-0013, on High Energy Arcing Fault Non-Suppression Probability, March 2018 (ML18075A071). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850 (including Supplement 1), which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. The guidance provided in the closeout memoranda to FAQs 08-0046 and 17-0013 is acceptable for use by licensees. NUREG-1934, NUREG-2169, NUREG-2178 (Volume 2), NUREG-2180, NUREG-2230, and NUREG-2262 provide guidance specifically developed by the NRC and EPRI to expand upon that found in NUREG/CR-6850 and closeout memoranda to FAQs, with either supplemental guidance and/or alternate approaches. |
| Internal Fires | FSS-F | FSS-F1 FSS-F2 |
Exposed Structural Steel |
Analysis of risk-relevant ignition sources with the potential for causing fire-induced failure of exposed structural steel | The fire-induced failure is of concern for any locations in which there are both exposed structural steel and high-hazard fire sources present. ASME/ANS RA-S1.1, Table 4-A.2.6-7 states that such fire sources would include a lube-oil fire ensuing from a catastrophic failure of the turbine as well as oil storage tanks, hydrogen storage tanks and piping, and mineral oil-filled transformers. Appendices N and O to NUREG/CR-6850, Volume 2, provides additional guidance particular to hydrogen and turbine generator fires, respectively. NUREG-1934 provides general guidance relative to fire scenario analysis as well as an example of how to treat a scenario involving exposed structural steel (i.e., Appendix F). |
ASME/ANS RA-S1.1, Table 4-A.2.6-7, clarifies that for lube-oil fires, the possibility of the effects of pooling, the flaming oil traversing multiple levels, and spraying from continued lube-oil pump operation should be considered. Analogous considerations apply to other ignition sources as well (e.g., oil storage tanks and hydrogen gas). | NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1934 provides guidance specifically developed by the NRC and EPRI to expand upon guidance provided in NUREG/CR-6850 and closeout memoranda to FAQs. |
| Internal Fires | FSS-G | FSS-G1 FSS-G2 FSS-G3 FSS-G4 FSS-G5 FSS-G6 FSS-G7 |
Multicompart-ment Fire Scenarios | Identification and assessment of multicompartment fire scenarios | General guidance on plant partitioning and its impact on the identification of multicompartment fire scenarios appears in Section 1 of NUREG/CR-6850, Volume 2, whereas Section 11.5.4 of NUREG/CR-6850, Volume 2, outlines a method for the identification and assessment of multicompartment fire scenarios, including treatment of fire barriers. ASME/ANS RA-S1.1, Table 4-A.2.6-8 states that fire barriers may include passive barriers (e.g., walls, normally closed fire doors, penetration seals, and other similar features that require no action, manual or automatic, to perform their intended function) or active barriers (e.g., normally open fire doors, dampers, water curtains, and other similar items that require that some action to occur for the element to perform its intended function), provided there is a justifiable basis for doing so and failure of credited elements is contemplated. Additional guidance on the treatment of passive barriers appears within Appendix Q to NUREG/CR-6850, Volume 2, whereas guidance on the treatment of active features, such as water curtains, as well as dependency between multiple features, is in Appendix P. |
As elaborated in Section 11.5.4 of NUREG/CR-6850, Volume 2, data on barrier failure probability is sparse and subject to many limitations; consequently, and unless otherwise justified, a screening probability (e.g., 0.1) is initially recommended. Note that for fire PRAs supporting applications, such as NFPA 805, there is much technical analysis and guidance (inclusive of applicable codes, standards, research, and testing) used to support compliance with the fundamental fire protection program and design elements established by NFPA 805, which includes deterministic assessment of fire barrier elements and could, in part, be the basis of a more probabilistic treatment within a fire PRA. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | IGN-A | IGN-A1 IGN-A2 IGN-A3 IGN-A4 IGN-A5 IGN-A6 IGN-A7 IGN-A8 IGN-A9 |
Ignition Frequency Estimation | General methodology estimating fire ignition frequencies for specific plant PAUs and/or fire scenarios for the internal fire PRA | A general methodology for estimating fire ignition frequencies for specific PAUs and/or fire scenarios that includes combining evidence from generic and plant specific data, apportioning fixed and transient ignition sources, and counting methods is provided in NUREG/CR-6850, Volume 2, Chapter 6 (particularly Chapter 6.3.1, Table 6-1, and Chapter 6.5), Chapter 8 (e.g., Chapter 8.5.5, and Chapter 11 (e.g., Chapter 11.3.1). As clarified in the adjacent column, additional guidance for select ignition sources has since been provided by FAQs 12-0064, 13-0005, 13-0006, 14-0007, and 14-0008, and 14-0009 as well as NUREG-2178 and NUREG-2262. |
The guidance on the general methodology for estimating fire ignition frequencies has been augmented and clarified by various technical reports, supplements, and updates. FAQs 12-0064 and 14-0007 provide supplemental guidance on the use of influence factors on transient fire frequencies. FAQ 06-0016 (NUREG/CR-6850 Supplement 1, Section 3) provides clarification of guidance on counting rules and criteria for electrical cabinets (i.e., Bin 15). FAQ 08-0042 (NUREG/CR-6850 Supplement 1, Section 8) further clarifies guidance on well-sealed and robustly secured cabinets given that cabinets meeting certain criteria should not be part of the Bin 15 count because doing so would dilute the count and thereby reduce the frequency apportioned to other cabinets. Note that regarding the treatment of well-sealed and robustly secured MCC electrical cabinets with voltage levels at 440V or greater, FAQ 14-0009 reiterates and reinforces the counting guidance within NUREG/CR-6850, Volume 2. Supplemental guidance particular to the treatment of self-ignited cable fires and cable fires caused by welding and cutting is provided by FAQ 13-0005, which provides an alternate but more realistic approach to that within NUREG/CR-6850, Volume 2, for the detailed modeling of these types of fire events and thereby the apportionment of frequency to modeled fire scenarios. FAQ 13-0006 augmented the guidance within NUREG/CR-6850, Volume 2, Chapters 6 and 11, to apportion the fire ignition frequency associated with junction boxes and related scenarios. For instance, for areas in which the number of junction boxes may be difficult to determine, FAQ 13-0006 recommends apportionment based on the ratio of cables in the area to the total number of cables. FAQ 07-0031 (NUREG/CR-6850 Supplement 1, Section 6) provides additional clarification on electric motors, pumps, transformers, and ventilation subsystems. Chapter 5 of NUREG-2718, Volume 2, expands this guidance further for motors and transformers. NUREG/CR-6850 Appendix L , provides guidance on the apportionment of the MCB fire ignition frequency. FAQs 06-0018 (NUREG/CR-6850 Supplement 1, Section 5) and 14-0008 provide supplemental guidance with regard to defining the MCB as well as treating MCB partitions and certain MCB configurations (e.g., when MCB controls exist on the back side of the MCB). NUREG-2178, Volume 2, provides additional guidance on MCB fire scenarios and presents an alternate approach to Appendix L to NUREG/CR-6850 for evaluating the risk of fire events originating in the MCB, including frequency apportionment. For HEAF ignition sources (i.e., switchgear, load centers and bus ducts), NUREG-2262 (e.g., Chapter 5) provides an update to counting guidance in NUREG/CR-6850, Volume 2, as supplemented by FAQ 06-0017 (NUREG/CR-6850 Supplement 1, Section 4) and FAQ 07-0035 (NUREG/CR-6850 Supplement 1, Section 7). Note that for certain ignition sources (e.g., Bins 16.a and 16.b), NUREG-2262 concludes that updated guidance was needed to reflect operating experience more accurately. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). FAQ 12-0064, Close-Out of National Fire Protection Association 805 Frequently Asked Question 12-0064on Hot Work/Transient Fire Frequency Influence Factors, January 2013 (ML12346A488). FAQ 13-0005, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0005, on Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting, December 2013 (ML13319B181). FAQ 13-0006, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0006 on Modeling Junction Box Scenarios in a Fire PRA, December 2013 (ML13331B213). FAQ 14-0007, Close-Out of Fire 14-0007, Revision 1 on Transient Fire Frequency Likelihood, March 2018 (ML18088B096). FAQ 14-0008, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0008, on Main Control Board Treatment, July 2014 (ML14190B307). FAQ 14-0009, Close-Out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0009 on Treatment of Well-Sealed MCC Electrical Panels Greater than 440V, April 2015 (ML15114A441). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire, Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric Motors, Indoor Dry Transformers, and the Main Control Board, June 2020 (ML20168A655). NUREG-2262, High Energy Arcing Fault Frequency and Consequence Modeling, April 2023 (ML23108A113). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850 (including Supplement 1), which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. The guidance provided in the closeout memoranda to FAQs 12-0064, 13-0005, 13-0006, 14-0007, 14-0008, and 14-0009 is acceptable for use by licensees. NUREG-2178 and NUREG-2262 provide guidance specifically developed by the NRC and EPRI to expand upon that found in NUREG/CR-6850 and closeout memoranda to FAQs, with either supplemental guidance and/or alternate approaches. |
| Internal Fires | IGN-A | IGN-A1 IGN-A2 IGN-A3 |
Ignition Frequency Estimation | Generic fire ignition frequencies and probability distribution data for the internal fire PRA | NUREG-2169 provides updated generic fire ignition frequencies from the frequencies provided in NUREG/CR-6850, Volume 2, Chapter 6, and probability distribution data. Select frequencies within NUREG-2169 are updated by NUREG-2178, Volume 2, NUREG-2230, and NUREG-2262, as clarified in the adjacent column. |
Selecting a combination of the lowest frequencies from the two main sources of generic fire ignition frequencies has not been allowed in NFPA 805 or TSTF-505 reviews. Note that the generic fire ignition frequencies promogulated by FAQ 08-0048 (NUREG/CR-6850 Supplement 1, Section 10) have been superseded, in their entirety, by NUREG-2169. By analogy, this would also apply to frequencies for cable fires initiated by welding and cutting that were developed by the EPRI-led Fire PRA Methods Review Panel. Following the methods outlined in NUREG-2169, Chapter 7 of NUREG-2230 provides an updated generic fire ignition frequency for non-HEAF electrical cabinets (Bin 15). For HEAF ignition sources (i.e., switchgear, load centers and bus ducts), updated frequencies appear in Chapter 5 of NUREG-2262 (Bins 16.a, 16.b, 16.1, and 16.2). For the MCB (Bin 4), updated frequencies appear in Chapter 7 of NUREG-2718, Volume 2 (Bin 4). |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database, January 2015 (ML15016A069). FAQ 08-0048, Update of Close-Out of National Fire Protection Association Frequently Asked Questions 08-0048, Fire Ignition Frequencies, May 2015 (ML15134A046). FAQ 12-0064, Close-Out of National Fire Protection Association Frequently Asked Questions 12-0064, Revision 1 on Hot Work/Transient Fire Frequency Influence Factors, January 2013 (ML12346A488). NRC-to-NEI Letter, Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, “Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires,” June 2012 (ML12171A583). NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire, Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric motors, Indoor Dry Transformers, and the Main Control Board, June 2020 (ML20168A655). NUREG-2230, Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinet Fires in Nuclear Power Plants, June 2020 (ML20157A148). NUREG-2262, High Energy Arcing Fault Frequency and Consequence Modeling, April 2023 (ML23108A113). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-2169, as supplemented by NUREG-2178, Volume 2, NUREG-2230, and NUREG-2262, provide guidance specifically developed by the NRC and EPRI to update the frequencies provided in NUREG/CR-6850, Volume 2. The guidance provided in the closeout memorandum to FAQs, including 12-0064, is acceptable for use by licensees. |
| Internal Fires | IGN-A | IGN-A7 | Ignition Frequency Estimation | Applying influence factors to fire ignition frequencies for the internal fire PRA | Clarification and augmentation of guidance in NUREG/CR-6850, Volume 2, Chapter 6 was needed to produce the appropriate range of fire ignition frequencies, namely through refined guidance on use of influence factors to apportion transient fire frequencies to account for preventive and corrective maintenance, occupancy, and stored combustible location conditions. FAQ 12-0064 provides this guidance. | The clarification in FAQ 12-0064 places specific limitations for when an influencing factor of less than "1.0" or "0" can be used. Guidance in FAQ 12-0064 was enhanced by FAQ 14-0007 to account for variations associated with transient ignition frequencies within a PAU. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). FAQ 12-0064, Close-Out of National Fire Protection Association 805 Frequently Asked Question 12-0064 on Hot Work/Transient Fire Frequency Influence Factors, January 2013 (ML12346A488). FAQ 14-0007, Close-Out of NFPA 805 FAQ 14-0007, Revision 1 on Transient Fire Frequency Likelihood, March 2018 (ML18088B096). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. The guidance provided in the closeout memorandum to FAQs 12-0064 and 14-0007 is acceptable for use by licensees. |
| Internal Fires | CF-A | CF-A1 CF-A2 CF-A3 |
Circuit Failure Mode Likelihood Analysis | Determination of fire-induced conditional probability of circuit and cable failure modes for the internal fire PRA | Guidance regarding the determination of fire-induced conditional probability of circuit and cable failure modes, including guidance on the probability of hot shorts and their durations for different situations, is provided NUREG/CR-6850, Volume 2, Chapter 10; NUREG/CR-7150, Volumes 2 and 3; and FAQ 08-0051 (NUREG/CR-6850 Supplement 1, Section 16). | NUREG/CR-7150, Volume 2, provides updated hot short probabilities and associated durations from those provided in NUREG/CR-6850, Volume 2, and guidance about addressing issues such as how to treat panel wiring, truck cables and instrument cables. Note that NUREG/CR-7150, Volume 2, eliminated the use of Option #2 (Computational Probability Estimates) documented within Chapter 10 of NUREG/CR-6850, Volume 2, as a technical approach for calculating spurious operation likelihood. NUREG/CR-7150, Volume 3, resolves the issues surrounding the treatment of a number of special cases, such as for shorting switches and control power transformers. FAQ 08-0047 (NUREG/CR-6850 Supplement 1, Section 15) provides clarifications to the guidance in NUREG/CR-6850 concerning the quantification of spurious probabilities that explicitly differentiate between independent and dependent combinations of other cables affected by the fire. Damage to the first cable may result in no further outcome for damage to a second cable. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE) – Volume 2: Expert Elicitation Exercise for Nuclear Power Plant Fire-Induced Electrical Circuit Failure, May 2014 (ML14141A129). NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE) – Volume 3: Technical Resolution to Open Issues on Nuclear Power Plant Fire-Induced Electrical Circuit Failure, November 2017 (ML17331B098). NUREG/CR-6850, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1, September 2010 (ML103090242). |
RG 1.205 (ML21048A448) references the methodologies documented in NUREG/CR-6850 (including Supplement 1) and NUREG/CR-7150, which are widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | FHR-A FHR-B FHR-C |
FHR-A1 FHR-A2 FHR-B1 FHR-B2 FHR-C1 |
Fire HRA | Adaptation of internal events HRA methods to address the impact of fire effects for the internal fire PRA | Adaptation of internal events HRA methods is needed to address the impact of fire effects. Fire impacts can require different kinds of operator response compared to internal events accident sequences (e.g., response to spurious operations). NUREG-1921 provides detailed guidance on the methodology described above. NUREG-1921, Supplement 1, provides guidance on how to model abandonment from the MCR necessitated by loss of habitability or control due to fire leading to shutdown from an alternate shutdown location, with NUREG-1921, Supplement 2, providing guidance for quantifying the probabilities of associated HFEs. NUREG/CR-6850, Volume 2, Chapter 12, provides high-level guidance on HRA for fire PRA. |
Specific HRA methods are not identified as HRA methods for fire PRA, but rather internal events HRAs methods that must be adapted to fire scenarios (e.g., CBDT and HCR/ORE methods for the cognitive contribution to error and the ASEP/THERP for the execution contribution as well as the ATHEANA approach, which relies on an expert elicitation quantification process). As summarized by NUREG-1921, NUREG-1792, and EPRI 1021081 all address the need to consider a minimum value for the joint probability of multiple HFEs within the same accident sequence or cutset. Table 2-1 of NUREG-1792 recommends that such joint HEPs should not be below 1E-5, whereas Table 4-4 of EPRI 102081 provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. For the fire HRA, specifically, NUREG-1921 recommends that the application of a lower bound follow the same guidance as was applied to the internal events PRA. As documented within RAIs regarding fire PRAs that support applications (e.g., NFPA 805 and TSTF-505 LARs), the NRC staff has accepted general use of a 1E-5 floor value for fire-specific joint HEPs and 1E-6 floor value for joint HEPs from the internal events that are credited in the fire PRA without providing additional justification. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1921, EPRI/NRC-RES Fire HRA Guidelines, July 2012 (ML12216A104). NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines – Qualitative Analysis for Main Control Room Abandonment Scenarios, Supplement 1, January 2020 (ML20035E043). NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines—Quantification Guidance for Main Control Room Abandonment Scenarios, Supplement 2, June 2019 (ML19162A378). NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), April 2005 (ML051160213). EPRI 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, October 2010. |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1921 provides guidance specifically developed by the NRC and EPRI for HRA in fire scenarios. ASME/ANS RA-S1.1 cites NUREG-1921 as an acceptable method for calculating the fire-related effects of all HFEs. Supplements 1 and 2 to NUREG-1921 provide guidance specifically developed by the NRC and EPRI for MCRA scenarios in fire PRA. |
| Internal Fires | FHR-D | FHR-D1 | Fire HRA | Identification and quantification of fire-specific recovery actions | Fire response operator actions, or recovery actions, are new post-initiating event operator actions required in response to a fire and are typically directed by the fire procedures; that is, they are specific to fire, must address fire-related effects, and are not already included as internal events HFEs. NUREG-1921 provides detailed guidance on the methodology to identify and quantify such actions within the fire PRA. NUREG-1921, Supplement 1, provides guidance on how to identify and model abandonment from the MCR necessitated by loss of habitability or control due to fire leading to shutdown from an alternate shutdown location, with NUREG-1921, Supplement 2, providing guidance for quantifying the probabilities of associated HFEs. NUREG/CR-6850, Volume 2, Chapter 12, provides high-level guidance on HRA for fire PRA. |
Additional guidance on the treatment of and risk insights associated with recovery actions, as evaluated under NFPA 805, is provided in RG 1.205 and NEI 04-02. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, July 2012 (ML12216A104). NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines – Qualitative Analysis for Main Control Room Abandonment Scenarios, Supplement 1, January 2020 (ML20035E043). NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines—Quantification Guidance for Main Control Room Abandonment Scenarios, Supplement 2, June 2019 (ML19162A378). NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 3, November 2019 (ML19351D277). NFPA 805, Performance-Based Standard for Fire Protection for LWR Reactor Electric Generating Plants, 2001. (Newer version available, but not endorsed for 10 CFR 50.48(c) transition) |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. RG 1.205 also endorses NEI 04-02. NUREG-1921 provides guidance specifically developed by the NRC and EPRI for HRA in fire scenarios. ASME/ANS RA-S1.1 cites NUREG-1921 as an acceptable method for calculating the fire-related effects of all HFEs. Supplements 1 and 2 to NUREG-1921 provide guidance specifically developed by the NRC and EPRI for MCRA scenarios in fire PRA. |
| Internal Fires | FQ-A FQ-B FQ-C FQ-D FQ-E |
FQ-A1 FQ-A2 FQ-A3 FQ-A4 FQ-A5 FQ-B1 FQ-C1 FQ-D1 FQ-D2 |
Fire PRA Quantification | Internal fire-induced CDF and LERF quantification | Section 14 of NUREG/CR-6850, Volume 2, provides a procedure for performing fire risk quantification and thus integrating insights from the various analyses used to support fire PRA development, e.g., circuit failure mode likelihood analysis, fire scenario analysis, fire HRA, etc. | Quantitative screening guidance and criteria are reviewed in ASME/ANS RA-S1.1 and Table 1-1.8-1, Chapter 7 to NUREG/CR-6850, Volume 2, as well as NUREG-1855. Additional guidance also appears in EPRI 1016737 and EPRI 1026511. Note that the appropriateness of any given quantification or screening practice is ultimately dependent on the given application for which the fire PRA is quantified. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017 (ML17062A466). EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012. |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1855 provides guidance specifically developed by the NRC to address matters of uncertainty, including related quantitative screening. Guidance in EPRI 1016737 and EPRI 1026511, as stated within NUREG-1855, complements that of the NUREG given that they expand upon specific methods to treat uncertainty and their use in applications. |
| Internal Fires | Multiple | PP-C1 ES-D1 CS-C1 QLS-B1 PRM-C1 PRM-C2 FSS-H1 FSS-H2 IGN-B1 CF-B1 FHR-E1 FQ-G1 FQ-G3 |
Fire PRA Documentation | Documentation of the fire PRA and its results | Section 16 to NUREG/CR-6850, Volume 2, provides the general practice for documenting the fire PRA to allow for its review and written basis. ASME/ANS RA-S1.1 also provides guidance for each relevant SR. | Note that the appropriateness of any given documentation practice is ultimately dependent on the given application to which the fire PRA is applied. | NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). | RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. |
| Internal Fires | Multiple | FQ-F1 FQ-F2 FQ-G2 PP-B7 PP-C2 ES-C2 ES-D2 QLS-A5 QLS-B2 PRM-B15 PRM-C3 FSS-G8 FSS-H3 IGN-A10 IGN-B2 CF-A3 CF-B2 FHR-D2 FHR-E2 |
Uncertainty | Identification, characterization, and documentation of uncertainties in the fire PRA | Detailed guidance, inclusive of fire PRA, on treating different sources of epistemic uncertainty (i.e., parameter, model, and completeness uncertainties) are provided in NUREG-1855 as well as EPRI 1016737 and EPRI 1026511. Consistent with relevant SRs in ASME/ANS RA-S1.1, methods to assess the effects of model uncertainties and related assumptions are discussed. Appendix U to NUREG/CR-6850, Volume 2, also outlines guidance for identifying uncertainties in the fire PRA and performing uncertainty and sensitivity analyses. |
Note that the impact of uncertainties on the fire PRA and its results is ultimately dependent on the given application to which the fire PRA is applied. Specific to fire modeling performed as part of the fire scenario analysis, NUREG-1934 provides methods to assess uncertainty, including modeling uncertainty. Supplement 1 to NUREG-1824 is designed to complement NUREG-1934 and provides a quantitative assessment of fire model uncertainty that supersedes the qualitative approach used in the original NUREG-1824. |
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities – Volume 2: Detailed Methodology, September 2005 (ML15167A411). NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, May 2007 (ML071730543). Supplement 1 to NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, November 2016 (ML16309A011). NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017 (ML17062A466). EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012 NUREG-1934, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG), November 2012 (ML12314A165). |
RG 1.205 (ML21048A448) references the fire PRA methodology documented in NUREG/CR-6850, which is widely used in fire PRAs supporting applications, such as NFPA 805, TSTF-505, and 10 CFR 50.69 LARs, that have been accepted by the NRC. NUREG-1855 provides guidance specifically developed by the NRC to address matters of uncertainty, including related modeling uncertainty. Guidance in EPRI 1016737 and EPRI 1026511, as stated within NUREG-1855, complements that of the NUREG given that they expand upon specific methods to treat uncertainty and their use in applications. NUREG-1824 and its supplement as well as NUREG-1934 provides guidance specifically developed by the NRC and EPRI to update guidance provided in NUREG/CR-6850 and closeout memoranda to FAQs. |
| Seismic | SHA-A | SHA-A1 SHA-A2 SHA-A3 SHA-A4 SHA-A5 SHA-A6 |
Hazard Analysis Approach and Study Level | Basis for the calculation of the frequencies for exceeding different levels of seismic horizontal vibratory ground motion | The state of practice for PSHA model development follows the NUREG/CR-6372 guidelines, which were developed further in NUREG-2213 and NUREG-2117. As stated in Section 5-A.2 of ASME/ANS RA-S1.1, these SSHAC guidelines describe a structured approach for producing a model that represents the center, body, and range of the technically defensible interpretations of the available Earth science information, as required by HLR SHA-A. Both NUREG-2115 and PEER 2018/08, as evaluated by the NRC staff within RIL 2020-11, demonstrate an implementation of the SSHAC guidelines. |
Section 2 of EPRI 1025287, as endorsed, also provides related but high-level guidance. The specified lower-bound magnitude for use in the SHA should be consistent with current practice, and RG 1.208 provides one acceptable approach to establishing this magnitude. The RG also provides an acceptable approach to establishing the number of standard deviations to be included in the analysis of GMPEs. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance," February 2013 (ML12319A074). NUREG-2213, Updated Implementation Guidelines for SSHAC Hazard Studies, October 2018 (ML18282A082). NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, Revision 1, April 2012 (ML12118A445). NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, Volumes 1 and 2, April 1997 (ML080090003, ML080090004). NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Volumes 1 through 6, January 2012 (ML12048A804, ML12048A833, ML12048A851, ML12048A858, ML12048A859, ML12048A860). RG 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, Revision 0, March 2007 (ML070310619). PEER 2018/08, Central and Eastern North America Ground-Motion Characterization – NGA-East Final Report, December 2018. RIL 2020-11, NRC Staff Evaluation of the Next Generation Attenuation for Central and Eastern North America Project (NGA-EAST) Ground Motion Model Characterization, September 2020 (ML20255A115). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). RG 1.208 as well as NUREG-1488, NUREG-2213, NUREG-2115, NUREG- 2117, and NUREG/CR-6372 provide guidance specifically developed by the NRC staff to inform the assessment of seismic risk, including related aspects of seismic PRA. PEER 2018/08 was evaluated by the NRC staff for use in regulatory activities within RIL 2020-11, and the NGA-East GMC model documented by PEER 2018/08 has supported the NRC staff assessments of updated seismic hazards for multiple NPP sites (e.g., ML23006A091, ML23192A447). |
| Seismic | SHA-B | SHA-B1 SHA-B2 SHA-B3 SHA-B4 SHA-B5 |
Input Data and Information | Identification and characterization of inputs to the PSHA | Inputs to the PSHA are to include characterization of uncertainty and be based on current geological, seismological, and geophysical data; local site topography; and surficial geologic and geotechnical site properties. Guidance on identifying, collecting, and evaluating all available data, models, and methods for PSHA model development is provided by the SSHAC guidelines (NUREG/CR-6372, NUREG-2213, and NUREG-2117). Additional guidance on data collection to support a site-specific analysis is provided in RGs 1.138 and 1.132. | Section 2 of EPRI 1025287, as endorsed, also provides related but high-level guidance. Depending on factors including regional characteristics, the geographical region around the site that is addressed in the PSHA can extend up to a radius of 1000 km. Section 1.1 of RG 1.208 provides related guidance. Note that when a regional model is changed (as opposed to augmented, such as with additional local faults), it is no longer considered to be the original model and no longer carries the SSHAC pedigree. The SSHAC guidelines provide guidance on when an existing regional study should be refined or replaced. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance", February 2013 (ML12319A074). NUREG-2213, Updated Implementation Guidelines for SSHAC Hazard Studies, October 2018 (ML18282A082). NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, Revision 1, April 2012 (ML12118A445). NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, Volumes 1 and 2, April 1997 (ML080090003, ML080090004). NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Volumes 1 through 6, January 2012 (ML12048A804, ML12048A833, ML12048A851, ML12048A858, ML12048A859, ML12048A860). RG 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, Revision 0, March 2007 (ML070310619). RG 1.132, Geologic and Geotechnical Site Characterization Investigations for Nuclear Power Plants, Revision 3, December 2021 (ML21298A054). RG 1.138, Laboratory Investigations of Soils and Rocks for Engineering Analysis and Design of Nuclear Power Plants, Revision 3, December 2014 (ML14289A600). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). RGs 1.208, 1.132, and 1.138 as well as NUREG-2213, NUREG-2115, NUREG-2117, and NUREG/CR-6372 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SHA-C | SHA-C1 SHA-C2 SHA-C3 SHA-C4 SHA-C5 |
Seismic Source Characterization | Identification and characterization of seismic sources as a means of assessing the frequency of exceedance of seismic ground motion levels | The SSHAC guidelines (NUREG/CR-6372, NUREG-2213, and NUREG-2117) provide guidance on seismic source characterization modeling for input to PSHA evaluations. An example of a regional study conducted using the SSHAC guidelines is documented in NUREG-2115, and site-specific examples (e.g., ML19273A907, ML18120A201) also are available. |
Section 2 of EPRI 1025287, as endorsed by the NRC staff, also provides related but high-level guidance. Note that when a regional model is changed (as opposed to augmented, such as with additional local faults), it is no longer considered to be the original model and no longer carries the SSHAC pedigree. The SSHAC guidelines provide guidance on when an existing regional study should be refined or replaced. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). NUREG-2213, Updated Implementation Guidelines for SSHAC Hazard Studies, October 2018 (ML18282A082). NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, Revision 1, April 2012 (ML12118A445). NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, Volumes 1 and 2, April 1997 (ML080090003, ML080090004). NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Volumes 1 through 6, January 2012 (ML12048A804, ML12048A833, ML12048A851, ML12048A858, ML12048A859, ML12048A860). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). NUREG-1488, NUREG-2213, NUREG-2115, NUREG-2117, and NUREG/CR-6372 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SHA-D | SHA-D1 SHA-D2 SHA-D3 SHA-D4 |
Ground Motion Characterization | Ground motion characterization to determine the range of seismic horizontal vibratory ground motion | The SSHAC guidelines (NUREG/CR-6372, NUREG-2213, and NUREG-2117) provide guidance on GMC modeling for input to PSHA evaluations. EPRI 3002000717 and the more recent PEER 2018/08, as evaluated by the NRC staff within RIL 2020-11, demonstrate an implementation of these guidelines, the latter of which represents the latest GMC model (known as the NGA-East GMC model) for the Central and Eastern United States. Site-specific examples (e.g., ML19273A907, ML18120A201) also are available. |
Section 2 of EPRI 1025287, as endorsed by the NRC staff, also provides related but high-level guidance. Note that when a regional model is changed (as opposed to augmented, such as with additional local faults), it is no longer considered to be the original model and no longer carries the SSHAC pedigree. The SSHAC guidelines provide guidance on when an existing regional study should be refined or replaced. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). NUREG-2213, Updated Implementation Guidelines for SSHAC Hazard Studies, October 2018 (ML18282A082). NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, Revision 1, April 2012 (ML12118A445). NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, Volumes 1 and 2, April 1997 (ML080090003, ML080090004). NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Volumes 1 through 6, January 2012 (ML12048A804, ML12048A833, ML12048A851, ML12048A858, ML12048A859, ML12048A860). EPRI 3002000717, EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, June 2013 (ML13170A385). NRC Letter, Approval of Electric Power Research Institute Ground Motion Model Review Project Final Report for Use by Central and Eastern United States Nuclear Power Plants, August 2013 (ML13233A102). PEER 2018/08, Central and Eastern North America Ground-Motion Characterization – NGA-East Final Report, December 2018. RIL 2020-11, NRC Staff Evaluation of the Next Generation Attenuation for Central and Eastern North America Project (NGA-EAST) Ground Motion Model Characterization, September 2020 (ML20255A115). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). NUREG-1488, NUREG-2213, NUREG-2115, NUREG-2117, and NUREG/CR-6372 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. The updated EPRI ground motion model was endorsed by the NRC staff (ML13233A102). PEER 2018/08 was evaluated by the NRC staff for use in regulatory activities within RIL 2020-11, and the NGA-East GMC model documented by PEER 2018/08 has supported the NRC staff assessments of updated seismic hazards for multiple NPP sites (e.g., ML23006A091, ML23192A447). |
| Seismic | SHA-E | SHA-E1 SHA-E2 SHA-E3 |
Site Response | Determination of the effects of local site response | Site response analyses are performed to quantify how near-surface geologic materials and their dynamic properties modify seismic vibratory motions entering the site from the underlying rock. Section 2.4 and Appendix B of EPRI 1025287, as endorsed by the NRC staff, provides detailed guidance on the site response analysis and development of site-specific amplification factors. As stated in RG 1.208, approaches to developing hazard-consistent site-specific soil motions, incorporating profile uncertainties, are provided in NUREG-6728. These approaches vary in complexity from simple deterministic amplification of probabilistically derived rock response spectra (Approaches 1, 2A, and 2B) to rigorous treatment of soil amplification within the PSHA (Approaches 3, 3A, 3B, and 4). RIL 2021-15, which demonstrates application of the SSHAC guidelines (NUREG/CR-6372, NUREG-2213, and NUREG-2117) to the conduct of local site response analyses, provides guidance on the treatment of uncertainties in the local site response analysis. |
As clarified in its endorsement of EPRI 1025287, the NRC staff noted that the SPID does not provide guidance on the development of FIRS used for performing soil-structure interaction analyses. Such is provided in the NRC staff’s endorsement letter to EPRI 1025287, DC/COL-ISG-017, and Section 3 of EPRI 3002004396. Also, the NRC staff’s endorsement letter to EPRI 1025287 provides additional guidance on the development of the site response in cases where limited site data exists. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance.” February 2013 (ML12319A074). NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk- Consistent Ground Motion Spectra Guidelines, November 2001 (ML013100232). DC/COL-ISG-017, Ensuring Hazard-Consistent Seismic Input for Site Response and Soil Structure Interaction Analyses, March 2010 (ML092230543). RG 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, Revision 0, March 2007 (ML070310619). EPRI 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, July 2015. NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 3002004396, “High Frequency Program: Application Guidance for Functional Confirmation and Fragility,” September 2015 (ML15218A569). RIL 2021-15, Documentation Report for SSHAC level 2: Site Response, November 2021 (ML21323A056). |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). RG 1.208, DC/COL-ISG-017, and NUREG/CR-6728 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. RIL 2021-15 documents work sponsored by the NRC that applied the SSHAC process for systematically identifying and propagating epistemic uncertainties in the local site response analysis. The site response approach and lessons learned documented by RIL 2021-15 have been ap plied to the NRC staff assessments of updated seismic hazards for multiple NPP sites (e.g., ML23006A091, ML23192A447). Note that such assessments update the site response analyses previously performed in NUREG/KM-0017 (ML21344A126). |
| Seismic | SHA-F | SHA-F1 SHA-F2 SHA-F3 SHA-F4 SHA-I2 |
Evaluation, Propagation of Hazard Uncertainties | Evaluation and propagation of uncertainties in the final quantification of hazard estimates | ASME/ANS RA-S1.1 indicates that the SSHAC guidelines (NUREG/CR-6372, NUREG-2213, and NUREG-2117) provide a structured approach for conducting a PSHA, inclusive of the hazard quantification process and propagation of uncertainties. Both NUREG-2115 and PEER 2018/08, as evaluated by the NRC staff within RIL 2020-11, demonstrate an implementation of these guidelines to the SSC and GMC components of PSHAs, respectively. Additionally, RIL 2021-15 applies the SSHAC process for systematically identifying and propagating epistemic uncertainties in the local site response analysis. |
Evaluation of uncertainties that fall within the scope of the PSHA study and SSHAC methodology is thoroughly addressed such that beyond the family of resulting hazard curves, further evaluation using the PRA quantification model is deemed unproductive and inappropriate by ASME/ANS RA-S1.1. In contrast, other SHA aspects that are treated outside the SSHAC process should be identified and evaluated through PRA model quantification to assess their effect on risk insights. |
NUREG-2213, Updated Implementation Guidelines for SSHAC Hazard Studies, October 2018 (ML18282A082). NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, Revision 1, April 2012 (ML12118A445). NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, Volumes 1 and 2, April 1997 (ML080090003, ML080090004). NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Volumes 1 through 6, January 2012 (ML12048A804, ML12048A833, ML12048A851, ML12048A858, ML12048A859, ML12048A860). PEER 2018/08, Central and Eastern North America Ground-Motion Characterization – NGA-East Final Report, December 2018. RIL 2020-11, NRC Staff Evaluation of the Next Generation Attenuation for Central and Eastern North America Project (NGA-EAST) Ground Motion Model Characterization, September 2020 (ML20255A115). RIL 2021-15, Documentation Report for SSHAC level 2: Site Response, November 2021 (ML21323A056). |
NUREG-1488, NUREG-2213, NUREG-2115, NUREG-2117, and NUREG/CR-6372 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. RIL 2021-15 documents work sponsored by the NRC that applied the SSHAC process for systematically identifying and propagating epistemic uncertainties in the local site response analysis. The site response approach and lessons learned documented by RIL 2021-15 have been applied to the NRC staff assessments of updated seismic hazards for multiple NPP sites (e.g., ML23006A091, ML23192A447). |
| Seismic | SHA-G | SHA-G1 SHA-G2 |
Spectral Shape and Vertical Motions | Determination of spectral shapes and vertical motions | The spectral shape determined in the SHA should be based on site-specific analysis. The guidance in Section 2 and Appendix B to EPRI 1025287 result in site spectral shapes that are based on the site amplification calculations, which consider site-specific soil properties. RG 1.208, which references guidance in NUREG/CR-6728 and NUREG/CR-6769, provides one approach to establishing vertical-horizontal spectral ratios, which can be combined with the appropriate horizontal spectra to derive vertical spectra. EPRI 3002012994 also provides guidance for developing mean vertical-horizontal ratios for a range of site conditions (rock and soil) and levels of ground motion. |
Guidance specific to the estimation of high frequency vertical spectra appears in Section 3.2 and Appendix A to EPRI 3002004396. | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk- Consistent Ground Motion Spectra Guideline, November 2001 (ML013100232). NUREG/CR-6769, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Development of Hazard- & Risk-Consistent Seismic Spectra for Two Sites, May 2002 (ML021440275, ML021440282). RG 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, Revision 0, March 2007 (ML070310619). EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessments, September 2018. EPRI 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, July 2015. NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 3002004396, “High Frequency Program: Application Guidance for Functional Confirmation and Fragility,” September 2015 (ML15218A569). |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002012994 consolidates and supersedes guidance from three reports: EPRI 103959, Methodology for Developing Seismic Fragilities, June 1994; EPRI 1002988, Seismic Fragility Applications Guide, December 2002; and EPRI 1019200, Seismic Fragility Application Guide Update, December 2009. Each of which is referenced by EPRI 1025287 as recommended guidance. NUREG/CR-6728 and NUREG/CR-6372 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SHA-H | SHA-H1 SHA-H2 SHA-H3 SHA-H4 |
Secondary Hazards | Evaluation of other seismic hazards in addition to vibratory ground motion | ASME/ANS RA-S1.1 provides general guidance on and outlines a process (see Figure 5-A.2-9) for the identification, screening, and valuation of other direct seismic hazards (e.g., fault displacement) or secondary hazards caused by vibratory ground motions (e.g., landslide, soil liquefaction, soil settlement, and earthquake-induced external flooding). Additional detailed guidance on the identification and treatment of seismically induced consequential hazards can be found in EPRI 3002000709 (e.g., Section 5.3). |
Specific guidance on consequential internal fire and flood events appears in Appendix G of EPRI 3002000709. Additional guidance on liquefaction is found in RG 1.198, NUREG/CR-5741, and EPRI NP-6041-SL. Earthquake-induced external flooding hazards are to be developed in concert with Part 8 of ASME/ANS RA-S1.1. |
RG 1.198, Procedures and Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites, November 2003 (ML033280143). NUREG/CR-5741, Technical Bases for Regulatory Guide for Soil Liquefaction, March 2000 (ML003701612). EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. |
EPRI 3002000709 provides implementation guidance to support the evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRA supporting applications. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). RG 1.198 and NUREG/CR-5741 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SFR-A | SFR-A1 SFR-A2 |
Seismic Fragility Analysis | Identification of SSCs, including associated failure modes and information relevant to modeling fragility correlation, within the scope of the fragility analysis | The scope of the fragility analysis is typically defined in the form of a SEL and includes identification of the SSCs that are credited in a seismic PRA. Table 5-A.2.2-2 of ASME/ANS RA-S1.1 provides guidance on information relevant to the fragility analysis and correlation. This includes the relationship between failure modes and mechanisms. The latter of which could result from acceleration demands due to seismic shaking, displacement demands in the case of seismic spatial interactions, etc., leading to a possible correlation between the failures of affected SSCs. EPRI 1025287 (e.g., Section 6.4.3) provides capacity-based criteria to determine those SSCs for which fragility analyses should be conducted. Additional guidance on correlation can be found in NUREG/CR-7237 as well as in EPRI 3002000709 (e.g., Section 5.6, Appendix D). |
The development of the SEL must ensure consistency between the failure modes defined by the systems analysis in the SPRA and the failure mechanisms defined by the fragility analysis. This ensures that the fragilities developed are applied to and modeled within the plant response model appropriately. | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance," February 2013 (ML12319A074). NUREG/CR-7237, Correlation of Seismic Performance in Similar SSCs (Structures, Systems, and Components), December 2017 (ML17348A155). EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRA supporting applications. NUREG/CR-7237 provides guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SFR-B | SFR-B1 SFR-B2 SFR-B3 SFR-B4 SFR-B5 SFR-B6 |
Seismic Fragility Analysis | Seismic response that the SSCs experience at failure | Seismic response parameters, such as displacements and in-structure accelerations, that represent a realistic estimate of the failure level of SSCs are required. Table 5-A.2.2-3 of ASME/ANS RA-S1.1 summarizes several different response-analysis approaches to potentially meet this intent. Additional guidance for seismic-response analyses is provided in Section 6.3.2 of EPRI 1025287. With regard to scaling an existing response analysis, Table 5-A.2.2-3 of ASME/ANS RA-S1.1 states that the scaling procedures given in EPRI NP-6041-SL and EPRI 3002012994 may be used. Guidance on structural models for seismic response calculations appears not only within Table 5-A.2.2-3 of ASME/ANS RA-S1.1 but also with EPRI 1025287 (e.g., Sections 6.3.1 and 6.3.3, Appendix C). For determining median-centered seismic response and its variability due to randomness and uncertainty in the various parameters affecting seismic response, EPRI 3002012994 provides guidance. Regarding SSI, Table 5-A.2.2-3 of ASME/ANS RA-S1.1 states that ASCE 4-16 and EPRI 3002012994 provide guidelines for performing such analysis, including treatment of variabilities. Section 6.3.3 of EPRI 1025287 further clarifies conditions where SSI effects may not need to be considered, such as for rock sites. If probabilistic response analysis is performed to calculate structural loads and floor response spectra, Table 5-A.2.2-3 of ASME/ANS RA-S1.1 and ASCE 4-16 speak to the number of simulations required to ensure stable responses. |
Note that EPRI 1025287 does not provide specific guidance on performing new response analysis for use in developing ISRS and fragilities. The new response analysis is generally conducted when the criteria for use of existing models are not met or if more realistic estimates are deemed necessary. | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance," February 2013 (ML12319A074). EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991 (NP-6041-SLR1). EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013 (3002000709). EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessments, September 2018. ASCE 4-16, Seismic Analysis of Safety-Related Nuclear Structures, April 2017. RIL 2021-05, Evaluation of ASCE 4-16 and ASCE 43-18 (Draft) for Use in the Risk-Informed, Performance-Based Seismic Design of Nuclear Power Plant Structures, Systems, and Components, July 2021 (ML21194A062). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. EPRI 3002012994 consolidates and supersedes guidance from three reports: EPRI 103959, Methodology for Developing Seismic Fragilities, June 1994; EPRI 1002988, Seismic Fragility Applications Guide, December 2002; and EPRI 1019200, Seismic Fragility Application Guide Update, December 2009. Each of which is referenced by EPRI 1025287 as recommended guidance. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). ASCE 4-16 was evaluated for adequacy of use by the NRC (ML21194A062). |
| Seismic | SFR-C | SFR-C1 SFR-C2 |
Seismic Fragility Analysis | Fragility thresholds | The term “inherently rugged” refers to seismic capacities well beyond the capacities of SSCs that normally govern seismic risk. Inherently rugged. As indicated by Table 5-A.2.2-4 of ASME/ANS RA-S1.1, EPRI 1025287 (e.g., 6.4.3) and EPRI 3002012994 include extensive discussions on the meaning of “inherently rugged” as well as a list of types of SSCs that are typically considered inherently rugged. Guidance that can be used for establishing the basis for a fragility threshold is provided in EPRI 1025287 (e.g., Section 6.4.3), EPRI 3002012994, and EPRI NP-6041-SL. They provide generic fragility screening-level seismic capacities as well as guidance on how to justify that SSCs meet the fragility screening levels. |
Table 5-A.2.2-4 of ASME/ANS RA-S1.1 clarifies that the fragility threshold in high-seismicity sites may be higher than the generic screening-level capacities provided in the cited technical basis documents. For such cases, alternate criteria should be developed and justified to establish seismic capacities for comparison with the higher fragility threshold. The capacities could be based on a combination of the use of site seismic design criteria, site-specific test data, and bounding analyses. | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessments, September 2018. |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002012994 consolidates and supersedes guidance from three reports: EPRI 103959, Methodology for Developing Seismic Fragilities, June 1994; EPRI 1002988, Seismic Fragility Applications Guide, December 2002; and EPRI 1019200, Seismic Fragility Application Guide Update, December 2009. Each of which is referenced by EPRI 1025287 as recommended guidance. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). |
| Seismic | SFR-D | SFR-D1 SFR-D2 SFR-D3 SFR-D4 SFR-D5 SFR-D6 SFR-D7 |
Seismic Fragility Analysis | Walkdowns for establishing or confirming as-built, as-operated conditions | Fragilities of SSCs in an SPRA should represent as-built and as-operated conditions. The accepted practice to achieve this requirement is through plant walkdowns. EPRI 1025286 provides guidance on the performance of such walkdowns. | Additional historical guidance developed as part of the IPEEE and USI A-46 efforts is also relevant. Such guidance includes NUREG/CR-4334, EPRI NP-6041-SL, and the SQUG GIP. The latter two documents address seismically induced interactions, including fire and flooding. | EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 1025286, Seismic Walkdown Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, June 2012. NRC Letter, Endorsement of Electric Power Research Institute (EPRI) Draft Report 1025286, “Seismic Walkdown Guidance,” May 2012 (ML12145A529). Seismic Qualification Utility Group (SQUG) Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 3A, December 2001 (ML040560263). NUREG/CR-4334, An Approach to the Quantification of Seismic Margins in Nuclear Power Plants, August 1985 (ML090500182). |
EPRI 1025286 was endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). NUREG/CR-4334 provides guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. The SQUG GIP has been reviewed for use by the NRC; this latest version referenced has been updated with clarifications, interpretations, positions, and exceptions identified by the NRC staff. |
| Seismic | SFR-E | SFR-E1 SFR-E2 SFR-E3 SFR-E4 |
Seismic Fragility Analysis | Analysis of failure mechanisms affecting failure modes modeled in the plant-response analysis using the CDFM (or hybrid) approach | The CDFM (or hybrid) fragility approach is a simplified method of estimating fragilities. It first calculates the HCLPF capacity and then estimates the family of fragility curves based on conservatively biased, generic values of logarithmic standard deviations (see Table 6-2 in EPRI 1025287). Detailed guidance on and examples of the CDFM (or hybrid) method can be found in EPRI 1025287 (e.g., Section 6.4.1), EPRI 3002012994, and EPRI NP-6041-SL. | As summarized in EPRI 1025287 and EPRI 3002012994, CDFM-based fragilities, although more approximate, are sufficiently accurate for first-order estimates of plant seismic risk measures and relative significance for the purposes of identifying risk-significant SSCs, for which more realistic fragilities must be developed using the SOV approach. As summarized in Table 5-A.2.2-6 of ASME/ANS RA-S1.1, the use of generic or conservative fragilities can affect risk metrics and risk insights through distortion and masking. Consequently, justification for their use is required (e.g., sensitivity studies), and in doing so, consideration should be given to the combined effect of multiple generic or conservative fragilities due to SSC dependency in the seismic PRA. Note that in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident, some applications made use of an iterative approach in the application of generic or conservative fragilities such that each iteration further refined the resulting fragility estimates with greater realism and site-specificity based on SSC risk measures and relative significance. This is appropriate and consistent with EPRI 1025287, EPRI 3002012994, and ASME/ANS RA-S1.1, provided that the SOV approach is applied to risk-significant SSCs. Specific guidance on the definition of risk significance in the context of fragility analysis can be found in Table 5-A.2.2-6 of ASME/ANS RA-S1.1 and Section 6.4.3 of EPRI 1025287. For high frequency sensitive components, guidance on fragility analysis can be found in Section 6.4.2 and Figure 6-7 of EPRI 1025287. Supplemental guidance is provided in EPRI 3002004396 for applying the seismic fragility methods documented within EPRI 3002012994 to account for high-frequency input motions. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessments, September 2018. EPRI 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, July 2015. NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 3002004396, “High Frequency Program: Application Guidance for Functional Confirmation and Fragility,” September 2015 (ML15218A569). |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002012994 consolidates and supersedes guidance from three reports: EPRI 103959, Methodology for Developing Seismic Fragilities, June 1994; EPRI 1002988, Seismic Fragility Applications Guide, December 2002; and EPRI 1019200, Seismic Fragility Application Guide Update, December 2009. Each of which is referenced by EPRI 1025287 as recommended guidance. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). |
| Seismic | SFR-E | SFR-E1 SFR-E2 SFR-E3 SFR-E4 |
Seismic Fragility Analysis | Analysis of failure mechanisms affecting failure modes modeled in the plant-response analysis using the SOV approach | Although universally applicable, the SOV approach entails individually estimating the median values and logarithmic standard deviations of several different variables affecting capacity and demand; thus, it is more time and resource intensive relative to more simplified methods (e.g., the CDFM/hybrid approach). Consequently, as indicated in EPRI 1025287 and EPRI 3002012994, this method is reserved but required for SSCs deemed to be risk significant. Detailed guidance on and examples of the SOV method can be found in EPRI 3002012994. | As indicated in Table 5-A.2.2-6 of ASME/ANS RA-S1.1, realistic and site-specific fragilities are required for risk-significant SSCs in the seismic PRA unless conservative or generic fragility curves estimated using more simplified methods (e.g., the CDFM/hybrid approach) can be appropriately justified, that is, their use significantly changes neither risk metrics nor risk insights. In all other cases, the SOV approach is to be used. Specific guidance on the definition of risk significance in the context of fragility analysis can be found in Table 5-A.2.2-6 of ASME/ANS RA-S1.1 and Section 6.4.3 of EPRI 1025287. For high frequency sensitive components, guidance on fragility analysis can be found in Section 6.4.2 and Figure 6-7 of EPRI 1025287. Supplemental guidance is provided in EPRI 3002004396 for applying the seismic fragility methods documented within EPRI 3002012994 to account for high-frequency input motions. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk Assessments, September 2018. EPRI 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, July 2015. NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 3002004396, “High Frequency Program: Application Guidance for Functional Confirmation and Fragility,” September 2015 (ML15218A569). |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002012994 consolidates and supersedes guidance from three reports: EPRI 103959, Methodology for Developing Seismic Fragilities, June 1994; EPRI 1002988, Seismic Fragility Applications Guide, December 2002; and EPRI 1019200, Seismic Fragility Application Guide Update, December 2009. Each of which is referenced by EPRI 1025287 as recommended guidance. |
| Seismic | SPR-A | SPR-A1 SPR-A2 SPR-A3 SPR-A4 |
Seismically Induced Initiating Events | Identification of seismically induced initiating events for the seismic PRA | The seismic plant response model is to consider the entire spectrum of seismically induced initiating events that could cause risk-significant accident sequences and/or accident-progression sequences. This includes those caused directly by the seismic event (e.g., LOOP, LOCA, etc.) as well as seismically induced hazard events resulting from secondary hazards (e.g., seismically induced internal flooding, external flooding, fire ignition sources, etc.). EPRI 3002000709 (e.g., Sections 5 and Appendix C) provides related guidance, and as stated in Table 5-A.2.3-2 of ASME/ANS RA-S1.1, a reasonably complete list of seismically induced external hazards to be addressed for the possibility of seismically induced events. Note also that Appendix G to the guide provides specific guidance on seismically induced internal fire and flooding events. Lastly, NUREG-1407 provides guidance relevant to seismically induced initiating events as well. |
While it does not provide explicit guidance on the identification of other seismically induced initiating events, EPRI 1025287 indicates that the seismic PRA should be built upon the existing models from the internal events PRA. Guidance for the identification of other nonvibratory hazards generated by seismic events (e.g., soil liquefaction, fault displacement) is consistent with sources cited under HLR SHA-H. This includes the guidance on soil liquefaction within RG 1.198, NUREG/CR-5741, and EPRI NP-6041-SL. Earthquake-induced external flooding hazards are to be developed in concert with Part 8 of ASME/ANS RA-S1.1. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. NUREG/CR-5741, Technical Bases for Regulatory Guide for Soil Liquefaction, March 2000 (ML003701612). NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). RG 1.198, Procedures and Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Site, November 2003 (ML033280143). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support the evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRA supporting applications. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). RG 1.198, NUREG-1407, and NUREG/CR-5741 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SPR-B | SPR-B1 SPR-B2 SPR-B3 SPR-B4 SPR-B5 SPR-B6 SPR-B7 SPR-B8 SPR-B9 SPR-B10 SPR-B11 SPR-B12 SPR-B13 |
Plant Response Modeling | Construction of the seismic plant response model from the internal events PRA accident sequences and systems logic model, addressing sequences unique to seismic and incorporation of seismic-induced impacts | The internal events PRA accident sequences are the primary source for developing a seismic-specific PRA model. However, accident sequences screened out of the internal events PRA or those that represent unique challenges need to be identified. Consequently, significant new logic and additional modeling are required to account for accident scenarios that involve seismically induced initiating events and failures. Guidance for doing so is provided in NUREG-1407 and NUREG/CR-4840 as well as in Section 5 and Appendix C to EPRI 3002000709. Additional guidance on the correlation of seismically induced SSC failures can be found in NUREG/CR-7237 as well as in EPRI 3002000709 (e.g., Section 5.6, Appendix D). EPRI 1025287 (e.g., Section 6.4.3) and Table 5-A.2.3-3 of ASME/ANS RA-S1.1 provide guidance on determining those SSCs that may be omitted from explicit modeling in the SPRA via relative and cumulative screening. For high frequency sensitive components, guidance can be found in EPRI 1025287 (e.g., Section 6.4.2 and Figure 6-7), and this includes guidance for cases in which either circuit analysis or operator-action analysis may be used as part of the seismic PRA to understand a given relay’s role in plant safety. Additional guidance on the treatment of low-fragility relays or relay chatter for use in a seismic PRA can be found in Section 5.7 of EPRI 3002000709. Specific guidance on the modeling of seismically induced consequential hazards, including internal fire and flood events, appears in EPRI 3002000709 (e.g., Section 5.3 and Appendix G) and NUREG-1407. External flooding hazards retained in the seismic PRA should be addressed in concert with Part 8 of ASME/ANS RA-S1.1. Guidance on multi-unit impacts is provided in Table 5-A.2.3-3 of ASME/ANS RA-S1.1. |
While it does not provide explicit guidance on the identification of other seismically induced initiating events, EPRI 1025287 indicates that the seismic PRA should be built upon the existing models from the internal events and other hazard PRAs that are relevant to the results of the seismic PRA. Per SR SPR-B2 of ASME/ANS RA-S1.1 significant deficiencies identified by peer review for these models should be resolved and incorporated into the seismic PRA plant response model. Guidance for doing so is provided by NEI 17-07, which superseded NEI 12-13 and was endorsed by RG 1.200, Revision 3. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). NUREG/CR-4840, Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150, November 1990 (ML063460465). NUREG/CR-7237, Correlation of Seismic Performance in Similar SSCs (Structures, Systems, and Components), December 2017 (ML17348A155). EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. NEI 17-07, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, August 2019 (ML19231A182). NEI 12-13, External Hazards PRA Peer Review Process Guidelines, August 2012 (ML12240A027). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. NUREG-1407, NUREG/CR-4840, and NUREG/CR-7237 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. NEI 17-07 was endorsed by RG 1.200, Revision 3, and contains a consolidated update of PRA peer review guidance from NEI 00-02, NEI 05-04, NEI 07-12, and NEI 12-13. |
| Seismic | SPR-C | SPR-C1 SPR-C2 SPR-C3 SPR-C4 SPR-C5 SPR-C6 |
Seismic Equipment List | Identification of SSCs that contribute to accidents included in the seismic plant response model | SSCs that may induce and/or be affected by seismic-related hazards are to be identified for evaluation in the seismic PRA. These SSCs are included in the SEL, the development of which is addressed by guidance in NUREG-1407, EPRI NP-6041-SL, and EPRI 3002000709 (e.g., Sections 5.1 and 5.2, Appendix E). Specific guidance on SEL SSCs that induce or are affected by seismically induced consequential hazards, including internal fire and flood sources, also appears in EPRI 3002000709 (e.g., Sections 5.3 and Appendix G) and NUREG-1407. Those SSCs that are affected by external flooding hazards should be addressed in concert with Part 8 of ASME/ANS RA-S1.1. Specific guidance on failure modes of interest in a seismic PRA can be found in Table 5-A.2.3-4 of ASME/ANS RA-S1.1 and EPRI 3002000709 (e.g., Sections 4.1.2). For high-frequency sensitive components, EPRI 1025287 (e.g., Section 2.4) and EPRI 3002004396 provide guidance. Section 6.5 of EPRI 1025287 provides LERF considerations for the SEL. |
The development of the SEL must ensure consistency between the failure modes defined by the systems analysis in the SPRA and the failure mechanisms defined by the fragility analysis. | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, July 2015. NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 3002004396, “High Frequency Program: Application Guidance for Functional Confirmation and Fragility,” September 2015 (ML15218A569). |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). NUREG-1407 provides guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| Seismic | SPR-D | SPR-D1 SPR-D2 SPR-D3 SPR-D4 SPR-D5 |
Seismic HRA | Adaptation of internal events HRA methods to address the impact of seismic effects for the seismic PRA | Adaptation of internal events HRA methods are needed to address the impact of seismic-related effects. Seismic impacts can require different kinds of operator response compared to internal event accident sequences (e.g., response to spurious operations). Related guidance can be found in EPRI 3002000709 (e.g., Section 5.5, Appendix C) as well as EPRI 3002008093. | Specific HRA methods are not identified as HRA methods for seismic PRA, but rather internal events HRA methods that must be adapted to seismic scenarios (e.g., CBDT and HCR/ORE methods for the cognitive contribution to error and the ASEP/THERP for the execution contribution as well as the ATHEANA approach which relies on an expert elicitation quantification process). | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, “Seismic Evaluation Guidance,” February 2013 (ML12319A074). EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. EPRI 3002008093, An Approach to Human Reliability Analysis for External Events with a Focus on Seismic, December 2016. |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. EPRI 3002008093 provides a framework for HRA, a general screening approach, and a detailed quantification approach, which adapts, for external events, the EPRI HRA methodology that has been widely used to support internal events and fire PRA applications. Consequently, this report has been cited in a number of seismic PRAs supporting applications. |
| Seismic | SPR-E | SPR-E1 SPR-E2 SPR-E3 SPR-E4 SPR-E5 SPR-E6 SPR-E8 |
Seismic PRA Quantification | Seismic-induced CDF and LERF quantification | The analysis to quantify CDF and LERF represents an integration of the seismic hazard, fragility, and plant response analyses. Guidance on the quantification and convergence analysis for a seismic PRA is provided in EPRI 3002000709, specifically Section 5.9 and Appendix C. Regarding LERF, Table 5-A.2.3-6 of ASME/ANS RA-S1.1 indicates that a LERF model for internal events is used as the basis for the seismic PRA. Specific LERF-related guidance for a seismic PRA appears in Section 5.8 of EPRI 3002000709 as well as Section 6.5.1 of EPRI 1025287. |
ASME/ANS RA-S1.1 does not recommend any specific seismic PRA methodology, including quantification methods; however, quantification of a seismic PRA presents additional challenges relative to the internal events PRA model. For instance, internal events PRA quantification methods that make use of various approximations (e.g., rare event, delete-term, and minimum cutset upper bound) tend to significantly overestimate results and thereby distort risk insights when applied to seismic PRAs, which contain high failure probability events. Consequently, to overcome this limitation, seismic PRAs supporting applications applied alternative methods, such as BDD logic employed by the EPRI Advanced Cutset Upper Bound Estimator software, to calculate more accurate results and importance measures. However, while these alternative methods, including BDD-based quantification, can yield more accurate (and even, in theory, exact) results, they require considerably more time and computing resources to do so. To overcome this limitation, the more resource-intensive methods were often applied to a specified number of high-ranking minimal cutsets, whereas the more traditional MCUB-based quantification methods were applied to the remaining low-ranking minimal cutsets. While doing so renders more accurate quantification possible, a sufficient portion of the risk must be addressed to alleviate distortive effects on results and risk insights. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance", February 2013 (ML12319A074). EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. |
| Seismic | Multiple | SPR-E8 SPR-E7 SPR-F3 SHA-F3 SHA-F4 SHA-I2 SFR-E5 SFR-F2 |
Uncertainty | Identification, characterization, and documentation of uncertainties in the seismic PRA | Detailed guidance, inclusive of seismic PRA, on treating different sources of epistemic uncertainty (i.e., parameter, model, and completeness uncertainties) is provided in NUREG-1855 as well as in EPRI 1016737 and EPRI 1026511. Consistent with relevant SRs in ASME/ANS RA-S1.1, methods to assess the effects of model uncertainties and related assumptions are discussed. Additional guidance on uncertainty and sensitivity analysis in a seismic PRA is provided in EPRI 3002000709, specifically Section 5.10 and Appendix C. |
Note that the impact of uncertainties on the seismic PRA and its results is ultimately dependent on the given application to which the seismic PRA is applied. Though EPRI 1025287 does speak to specific elements that need to be documented, it does not provide explicit guidance for assessing uncertainty in the results of a seismic PRA. As discussed herein under HLR SHA-F, the valuation of uncertainties that fall within the scope of the PSHA study and SSHAC methodology are thoroughly addressed such that beyond the family of resulting hazard curves, further evaluation using the PRA quantification model is deemed unproductive and inappropriate by ASME/ANS RA-S1.1. Tables 5-A.2.1-6 and 5-A.2.2-6 of ASME/ANS RA-S1.1 identify examples of potentially significant assumptions associated with the seismic hazard and fragility analyses, respectively. |
EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance,” February 2013 (ML12319A074). NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Revision 1, March 2017 (ML17062A466). EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012. EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. |
EPRI 1025287 and EPRI 3002004396 were endorsed by the NRC and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. NUREG-1855 provides guidance specifically developed by the NRC to address matters of uncertainty, including related quantitative screening. Guidance in EPRI 1016737 and EPRI 1026511, as stated within NUREG-1855, complements that of the NUREG given that they expand upon specific methods to treat uncertainty and their use in applications. |
| Seismic | Multiple | SHA-I1 SHA-I2 SFR-F1 SFR-F2 SPR-F1 SPR-F2 SPR-F3 SPR-F4 |
Documentation | Documentation of the seismic PRA and its results | EPRI 1025287 (e.g., Section 6.8) and EPRI 3002000709 (e.g., Appendix F) provides the general practice for documenting a seismic PRA to allow for its review and written basis. ASME/ANS RA-S1.1 also provides guidance for each relevant SR. For seismic PRA studies conducted as part of the IPEEE program, Appendix C to NUREG-1407 outlines detailed documentation and reporting guidelines associated with them. Similar guidance appears in EPRI NP-6041-SL. Lastly, the SSHAC guidelines (NUREG-2213, NUREG-2117, NUREG/CR-6372) have significant documentation requirements consistent with HLR SHA-I and related to the PSHA. |
Note that the level of effort for developing seismic PRA documentation depends on whether and to what extent existing information is being used. For example, sites that make use of an existing seismic source model (e.g., NUREG-2115) or ground motion model (e.g., EPRI 3002000717) can take advantage of the significant documentation available for those projects. For those sites where a new PSHA is performed, particularly for a SSHAC Level 3 PSHA, a significant effort may be necessary to develop adequate PSHA documentation. | EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012 (ML12333A170). NRC-2013-0038, Endorsement of Electric Power Research Institute Final Draft Report 1025287, "Seismic Evaluation Guidance," February 2013 (ML12319A074). NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). NUREG-2213, Updated Implementation Guidelines for SSHAC Hazard Studies, October 2018 (ML18282A082). NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, Revision 1, April 2012 (ML12118A445). NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, Volumes 1 and 2, April 1997 (ML080090003, ML080090004). NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Volumes 1 through 6, January 2012 (ML12048A804, ML12048A833, ML12048A851, ML12048A858, ML12048A859, ML12048A860). EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991. EPRI 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013. EPRI 3002000717, EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, June 2013 (ML13170A385). |
EPRI 1025287 was endorsed by the NRC staff and used in seismic PRAs supporting applications, such as TSTF-505, and 10 CFR 50.69 LARs, as well as in response to Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). EPRI 3002000709 provides implementation guidance to support evaluation of seismic risk evaluations, including those reflected by EPRI 1025287, and has thus been heavily used in seismic PRAs supporting applications. EPRI NP-6041-SL represents one of two primary sets of guidance and methods used to perform a seismic margins assessments reviewed by the NRC in support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20). NUREG-1407, NUREG-2213, NUREG-2115, NUREG-2117, and NUREG/CR-6372 provide guidance specifically developed by the NRC to inform the assessment of seismic risk, including related aspects of seismic PRA. |
| High Winds | All | All | High Winds PRA Framework | Wind Hazard, Fragility, and Plant Response Analyses | EPRI 3002003107 documents a graded approach for performing the major tasks of a high winds PRA. Guidelines and methods for the risk assessment of high winds hazards contained within this document include those related to conducting a wind hazard analysis, evaluating the strength of SSCs and their responses to wind and wind-driven missile loads (i.e., assessment of SSC fragilities), and developing and quantifying a plant response model that incorporates the results from the wind hazards analysis. Related guidance can also be found in the following: EPRI 3002008092 for developing a high-winds equipment list and performing walkdowns; EPRI 3002015994 for evaluating windborne missile fragilities; EPRI 3002018232 for characterizing loss of offsite power duration and recovery; and EPRI 3002020906 for consideration of wind-driven rain. |
As indicated in Section 7-A.1 of ASME/ANS RA-S1.1, the nature of wind hazards and effects result in a broad technical scope for development of a high winds PRA. While the state-of-the-art and practice has matured significantly over the past decade to address this breadth of scope, a probabilistic framework for developing a high winds PRA is still in its relative nascency. That is, while evidence of use demonstrates its guidance is sufficient to support the development of a high winds PRA, EPRI 3002003107, deemed as early research and development by EPRI, has had limited application and identifies potential technical gaps and areas for further research. However, this and other referenced EPRI guidance represents the state of practice. Section 7-A.4 of ASME/ANS RA-S1.1 provides many resources that may be used to inform the development of a high winds PRA but emphasizes that due to the current and evolving nature of the state of practice and experience related to high winds PRA, discretion must be applied in assessing the appropriateness and acceptability of associated methods. Examples of such resources include NUREG/CR-4461, NUREG-7004, and NUREG-7005. In support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20), vulnerabilities to severe accidents caused by external events, including high winds, were identified using a progressive screening approach that represented a series of analyses in increasing levels of detail, effort, and resolution. Guidance on this approach and perspectives gained from its implementation can be found in NUREG-1407 and NUREG-1742, respectively. As addressed separately in this table, probabilistic methodologies of limited scope and application have been documented, applied, and accepted within applications to address tornado missiles. |
EPRI 3002003107, High-Wind Risk Assessment Guidelines, June 2015. EPRI 3002008092, High Winds Equipment List and Walkdown Guidance, 2016. (This document is not available online; however, it appears in the list of available EPRI Risk and Safety Management publications.) EPRI 3002015994, Evaluation of Windborne Missile Fragilities for Piping, Vent Stacks, Liquid-Filled Tanks, and Concrete Panels, September 2019. EPRI 3002018232, High Wind Loss of Offsite Power Durations and Recovery, October 2020. EPRI 3002020906, Consideration of Wind-Driven Rain in High Winds PRA, 2021. (This document is not available online; however, it appears in the list of available EPRI Risk and Safety Management publications.) NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). NUREG-1742, Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, April 2002 (ML021270070, ML021270122, ML021270674). NUREG/CR-4461, Tornado Climatology of the Contiguous United States, Revision 2, February 2007 (ML070810400). NUREG-7004, Technical Basis for Regulatory Guidance on Design-Basis Hurricane-Borne Missile Speeds for Nuclear Power Plants, November 2011 (ML11341A102). NUREG-7005, Technical Basis for Regulatory Guidance on Design-Basis Hurricane Wind Speeds for Nuclear Power Plants, November 2011 (ML11335A031). |
Only a limited number of high winds PRAs have been developed to date, and even fewer have undergone peer review. In general, industry-developed methods, including those within EPRI and/or WCAP reports, have been used to support risk-informed applications. However, the NRC has neither commented on nor assessed these specific methods during their reviews of risk-informed applications. For example, the Callaway 10 CFR 50.59 application (ML21344A005) cites the use of EPRI 3002003107 and EPRI 3002008092. In short, some risk-informed applications have applied screening and/or conservative analyses; targeted probabilistic approaches (e.g., for tornado missiles); and/or PRAs for high winds to demonstrate that the risk being addressed was of no or insignificant impact on the application. Alternatively, others have incorporated the impact of hazard mechanisms within the internal events PRA. Lastly, at least one application (e.g., the Callaway 10 CFR 50.69 application, ML21344A005) has demonstrated that the high winds PRA was appropriately peer reviewed, consistent with RG 1.200, and that associated F&Os were closed using an approach, i.e., Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 (ML17086A431), approved by the NRC staff (ML17079A427). NUREG-1407 and NUREG-1742 provide guidance specifically developed by the NRC to inform the assessment of hazards and related risk. |
| High Winds | Multiple | Multiple | Tornado Missiles | TORMIS Risk Analysis | The TORMIS risk assessment methodology, as documented by EPRI NP-768, NP-769, and NP-2005 (Volumes 1 and 2), provides a probabilistic alternative for assessing the need for positive tornado missile protection for specific safety-related plant features in accordance with the criteria of NUREG-0800. | As noted in the NRC’s SER (ML080870291), the use of the EPRI methodology, or any tornado missile probabilistic study, is limited to the evaluation of specific plant features that involve additional costly tornado missile protective barriers or alternative systems. It provides licensees the option of revising the plant’s licensing basis for tornado missile protection from a purely deterministic methodology to one that includes limited use of a probabilistic approach. While approved for use at several plants, issues associated with the methodology’s implementation and limitations of use have been raised (RIS 2008-14 and RIS 2015-06). |
EPRI NP-768, Tornado Missile Risk Analysis, May 1978. EPRI NP-769, Tornado Missile Risk Analysis - Appendices, May 1978. EPRI NP-2005-V1, Tornado Missile Simulation and Design Methodology, Volume 1: Simulation Methodology, Design Applications, and TORMIS Computer Code, August 1981. EPRI NP-2005-V2, Tornado Missile Simulation and Design Methodology, Volume 2: Model Verification and Database Updates, August 1981. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, (ML070810350). Safety Evaluation Report, Electric Power Research Institute (EPRI) Topical Reports Concerning Tornado Missile Probabilistic Risk Assessment (PRA) Methodology, October 2023 (ML080870291). NRC Regulatory Issue Summary 2008-14, Use of TORMIS Computer Code for Assessment of Tornado Missile Protection, June 2008 (ML080230578). NRC Regulatory Issue Summary 2015-06, Tornado Missile Protection, June 2008 (ML15020A419). |
EPRI NP-768, NP-769, and NP-2005 (Volumes 1 and 2) have been approved for use by the associated SER; though, a plant’s current, site-specific licensing basis cannot be changed unless the NRC or licensee took action to specifically amend the operating license. Within some risk-informed applications (e.g., TSTF-505, 10 CFR 50.69), the results of the TORMIS methodology have been applied to assess the impact of tornado-generated missiles on SSCs included within the TORMIS analysis and thus demonstrate that the associated risk is of no or insignificant impact on the application. |
| High Winds | Multiple | Multiple | Tornado Missiles | TMRE | As documented in NEI 17-02, the TMRE assesses the risk posed by tornado missiles for the purpose of determining whether additional physical protection is warranted. It is a hybrid methodology comprised of two key elements: a deterministic element to establish the likelihood that an SSC will be struck and damaged by a tornado-generated missile, and a probabilistic element to assess the impact of the missile damage on CDF and LERF. | To date, the application of this method has been limited in that it may only be applied to discovered conditions where tornado-missile protection was required by the plant’s current licensing basis and not provided. Additionally, it must not be used either to remove existing tornado-missile protection, or to avoid providing tornado-missile protection during reviews done in support of the plant modification process. Lastly, as stated in license amendments approved to date (e.g., ML18347A385, ML18304A394), the NRC’s approval of the use of the methodology at individual plants does not signify that the NRC staff has generically approved NEI 17-02. | NEI 17-02, Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document, Revision 1 (ML17268A036). | The NRC staff has approved plant-specific application of the TMRE methodology, including any applicable deviations, for use at select plants via an approved license amendment. Within some risk-informed applications (e.g., TSTF-505, 10 CFR 50.69), the results of the TMRE methodology have been applied to assess the impact of tornado-generated missiles on SSCs included within the analysis and thus demonstrate that the associated risk is of no or insignificant impact on the application. |
| High Winds | Multiple | Multiple | Tornado Missiles | TMSC | The TMSC is a spreadsheet calculational tool that computes the conditional hit probability of tornado and high-wind missiles at power plant sites. The results from this tool can be used as inputs to quantify the contribution of risk from such missiles. EPRI 3002023809 and EPRI 3002023810, respectively, document the approach as well as its verification and validation | The TMSC was developed as an alternative to the TORMIS risk analysis approach as these tools determine SSC failure probabilities from wind-generated missiles differently. | EPRI 3002023809, Tornado Missile Strike Calculator (TMSC) Version 1.0, August 2022. EPRI 3002023810, Verification and Validation of the Tornado Missile Strike Calculator, October 2022. |
EPRI 3002023809 and EPRI 3002023810 have no evidence of use (e.g., in supporting a change to a site’s licensing basis, risk-informed application). |
| External Flood | All | All | External Flood PRA Framework | External Flood Hazard Analysis, Fragility Evaluation, and Plant Response Analysis | As indicated in Section 8-A.1.1 of ASME/ANS RA-S1.1, the probabilistic analysis of external floods has unique challenges, including the diverse nature of flooding phenomena, flood event duration, and complex severity measures. Owing to these technical challenges and despite the significant advances in the probabilistic modeling of hydrologic processes, a comprehensive probabilistic flood hazard assessment methodology has, as clarified by DG-1290, not yet been developed. Moreover, the current and evolving nature of the state of practice and experience related to external flooding PRA has yielded a range of diverse methods with varying degrees of maturity and acceptance, but at this time, there are no widely accepted probabilistic frameworks, methods, or toolsets for developing an external flood PRA. | Ongoing research and development efforts by the NRC staff, EPRI, national laboratories, and others within the industry have yielded insights into potential methods, including the frequency characterization of the various external flooding hazard mechanisms. Only recently has guidance been promulgated that informs the development of a comprehensive external flood PRA. This guidance appears in EPRI 3002023808, which addresses the derivation and characterization of the external flood hazards; the identification and fragility modeling of flood-related SSCs; and the integration of pre-flood and post-flood plant response actions. However, to date, there is no evidence of its use. Section 8-A.4 of ASME/ANS RA-S1.1 provides many resources that may be used to inform the development of an external flood PRA but emphasizes that due to the current and evolving nature of the state of practice and experience related to external flood PRA, discretion must be applied in assessing the appropriateness and acceptability of associated methods. In support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20), vulnerabilities to severe accidents caused by external events, including external floods, were identified using a progressive screening approach that represented a series of analyses in increasing levels of detail, effort, and resolution. Guidance on this approach and perspectives gained from its implementation can be found in NUREG-1407 and NUREG-1742, respectively. In response to Recommendations 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340), licensees were requested to reevaluate the flood-causing mechanisms at their sites against present-day regulatory guidance and methodologies being used for early site permits and combined license reviews. External flooding sources included the effects from local intense precipitation on the site, probable maximum flood on stream and rivers, storm surges, seiches, tsunami, and dam failures. Key guidance used to perform the reevaluations appears in RG 1.59 (and its proposed revision, DG-1290), ANSI/ANS 2.8-2019, NUREG-0800, NUREG/CR-6966, NUREG/CR-7046, NUREG/CR-7134, JLD-ISG-2012-06, and JLD-ISG-2013-01. Note that NUREG/CR-7046 provides a risk-informed screening approach for external flooding. This so-called hierarchical hazard assessment approach applies a progressively refined, stepwise screening process to the estimation of site-specific external flooding hazards necessary to evaluate whether SSCs are capable of performing their safety functions during a design basis flood. An alternate screening process can be found in ANSI/ANS 2.8-2019. For responding to reevaluated flooding hazards that exceed a facility’s design basis flood, integrated assessments were requested to demonstrate the adequacy of the existing plant design and mitigating strategies. NEI 16-05, as endorsed by JLD-ISG-2016-01, provides an acceptable methodology to perform focused evaluations and integrated assessments of flood mechanisms that exceed the design-basis flood parameters for a facility. Additional guidance appears in JLD-ISG-2012-05. In response to Recommendations 2.3 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340), licensees also performed verification walkdowns to assess external flood protection and mitigation capabilities as well as to verify the adequacy of and conformance with the current licensing basis. Guidance for doing so appears in NEI 12-07, as endorsed by the NRC staff (ML12144A142, ML12159A290). Additional guidance can be found in EPRI 3002015989. |
EPRI 3002023808, External Flooding Guidance for Probabilistic Risk Assessment, October 2022. RG 1.59, Design Basis Floods for Nuclear Power Plants, Revision 2 with errata, July 1980 (ML003740388). DG-1290, Proposed Revision 3 to Regulatory Guide 1.59, Design-Basis Floods for Nuclear Power Plants. February 2022 (ML19289E561). NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, (ML070810350). NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). NUREG-1742, Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, April 2002 (ML021270070, ML021270122, ML021270674). NUREG/CR-6966, Tsunami Hazard Assessment at Nuclear Power Plant Sites in the United States of America, March 2009 (ML091590193). NUREG/CR-7046, Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America, November 2011 (ML11321A195). NUREG/CR-7134, The Estimation of Very-Low Probability Hurricane Storm Surges for Design and Licensing of Nuclear Power Plants in Coastal Areas, October 2012 (ML12310A025). JLD-ISG-2012-05, Guidance for Performing the Integrated, Assessment for External Flooding, Revision 0, November 2012 (ML12311A214). JLD-ISG-2012-06, Interim Staff Guidance for Performing a Tsunami, Surge, or Seiche Hazard Assessment, Revision 0, January 2013 (ML12314A412). JLD-ISG-13-01, Interim Staff Guidance for Estimating Flooding Hazards due to Dam Failure, Revision 0, July 2013 (ML13151A153). JLD-ISG-2016-01, Guidance for Activities Related to Near-Term Task Force Recommendation 2.1, Flooding Hazard Reevaluation; Focused Evaluation and Integrated Assessment, Revision 0, April 2016 (ML16162A301). NEI 12-07, Guidelines for Performing Verification Walkdowns of Plant Flood Protection Features, May 2012 (ML12144A401). NEI 16-05, External Flooding Assessment Guidelines, Revision 1, June 2016 (ML16165A178). EPRI 3002015989, External Flooding Probabilistic Risk Assessment Walkdown Guidance, July 2019. ANSI/ANS 2.8-2019, Determining Design Basis Flooding at Power Reactor Sites, 2019. |
Given that well-established probabilistic frameworks and methods for assessing site-specific extreme precipitation and flooding events of interest are not widely available and/or used, the NRC staff, as clarified by DG-1290, will evaluate them on a case-by-case basis. However, such approaches have not been evaluated, and an external flood PRA has not been used in risk-informed applications that have been reviewed and approved by the NRC (e.g., TSTF-505, 10 CFR 50.69) to date. EPRI 3002023808 has no evidence of use. NEI 12-07 and NEI 16-05 were endorsed by the NRC staff and used in external flooding analyses in response to Recommendations 2.1 and 2.3 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). NUREG-0800, NUREG-1407, NUREG-1742, NUREG/CR-6966, NUREG/CR-7046, NUREG/CR-7134, JLD-ISG-2012-05, JLD-ISG-2012-06, JLD-ISG-2013-01, and JLD-ISG-2016-01 provide guidance specifically developed by the NRC to inform the assessment of external flooding hazards and related risk. ANSI/ANS 2.8-2019 is referenced as appropriate guidance by NEI 16-05, which is endorsed by the NRC staff, and DG-1290. EPRI 3002015989 provides additional guidance applied, in part, by licensees in response to Recommendation 2.3 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident (ML12053A340). |
| Other Hazards | HLR-XHA HLR-XFR HLR-XPR |
All | Other Hazards | PRA development for other hazard mechanisms | Table 6-B-1 of ASME/ANS RA-S1.1 provides a list of hazards for consideration. NUREG/CR-2300, particularly Section 10, outlines methods and procedures for selection and assessment of such hazards. Additional guidance and examples of related analyses can be found in NUREG/CR-4550, NUREG/CR-4832, NUREG/CR-4839, and NUREG/CR-5042. | Note that, as the scope of other hazards is so broad, the acceptability of any additional hazard-specific methods applied to address individual hazards is ultimately dependent on the given application to which the methods are used. | NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, January 1983 (ML063560439, ML063560440). NUREG/CR-4550, Analysis of Core Damage Frequency: Peach Bottom, Unit 2 External Events, Volume 4, Revision 1, Part 3 (ML20066C272). NUREG/CR-4832, Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods Integration and Evaluation Program (RMIEP): External Event Scoping Quantification, Volume 7, March 1993 (10166577). NUREG/CR-4839, Methods for External Event Screening Quantification: Risk Methods Integration and Evaluation Program (RMIEP) Methods Development, July 1992 (ML062260210). NUREG/CR-5042, Evaluation of External Hazards to Nuclear Power Plants in the United States, December 1987 (ML14196A083). |
NUREG/CR-2300, NUREG/CR-4550, NUREG/CR-4832, NUREG/CR-4839, and NUREG/CR-5042-01 provide guidance specifically developed by the NRC to inform and document the assessment of other hazards and related risk. |
| Other Hazards | HLR-EXT | All | Screening and Conservative Analysis | Screening and conservative analyses to exclude a hazard or hazard group from further evaluation | Specific hazards (or mechanisms) may be eliminated from a detailed PRA by using a qualitative or quantitative screening process. Section 5 of NUREG-1855 provides guidance on approaches to address the scope and level-of-detail items that are not modeled in a PRA and, ultimately, determine whether those missing scope and level-of-detail items are significant to the decision under consideration. Lastly, this guidance includes consideration of quantitative and qualitative screening criteria that can be applied. Table 1-1.8-1 of ASME/ANS RA-S1.1 also specifies the general criteria (both quantitative and qualitative) that can be used in considering whether a hazard or hazard group can be screened out from consideration in the construction of the PRA. |
Note that the appropriateness of any screening and/or conservative modeling practice is ultimately dependent on the given application to which the practice is used to support. In support of IPEEE for Severe Accident Vulnerabilities (Generic Letter 88-20), vulnerabilities to severe accidents caused by external events were identified and dispositioned using a progressive screening approach that represented a series of analyses in increasing levels of detail, effort, and resolution. Guidance on this approach and perspectives gained from its implementation can be found in NUREG-1407 and NUREG-1742, respectively. |
NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 (ML063550238). NUREG-1742, Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, April 2002 (ML021270070, ML021270122, ML021270674) NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017 (ML17062A466). EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012. |
The practice of using screening and conservative analyses to exclude a hazard or hazard group from further evaluation and thus to demonstrate that the risk being addressed is of no or insignificant impact on the application is frequently employed within risk-informed applications that have been reviewed and approved by the NRC (e.g., TSTF-505, 10 CFR 50.69). NUREG-1407 and NUREG-1742 provide guidance specifically developed by the NRC to inform the assessment of hazards and related risks. NUREG-1855 provides guidance specifically developed by the NRC to address matters of completeness uncertainty, including related qualitative and quantitative screening. Guidance in EPRI 1016737 and EPRI 1026511, as stated within NUREG-1855, complements that of the NUREG given that they expand upon specific methods to treat uncertainty and their use in applications. |
| Hazard(s) | HLR(s) | SR(s) | Category | Methods | Applicability | Limitations/Clarifications | Source of Technical Basis | Evidence of Use |