ML23142A022
| ML23142A022 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/22/2023 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Reynolds R Constellation Energy Generation |
| References | |
| EPID L-2022-LLA-0185, EPID L-2022-LLA-0186 | |
| Download: ML23142A022 (1) | |
Text
From: Richard Guzman To: Reynolds, Ronnie J:(Constellation Nuclear)
Subject:
Nine Mile Point Nuclear Station, Unit 1 - TSTF-505 and 10 CFR 50.69 PRA Audit Questions (EPIDs L-2022-LLA-0185, L-2022-LLA-0186)
Date: Monday, May 22, 2023 7:19:46 AM Attachments: NMP1 TSTF-505_50.69 PRA AQs.pdf
- Ron,
By letters dated December 15, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML22349A108 and ML22349A521), Constellation Energy Generation, LLC (CEG, the licensee) submitted two license amendment requests (LARs) for Nine Mile Point Nuclear Station, Unit 1 (NMP1). The proposed amendments would modify license DPR-63 and the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler 505 (TSTF-505), Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, and the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.
On January 31, 2023, the NRC staff issued an audit plan (ML23025A386) that conveyed intent to conduct a regulatory audit to support its review of the subject licensing actions.
Based on the commonalities between the LARs and subsequent overlap in technical content and review personnel, the staff is conducting a combined audit that addresses both LARs. In the audit plan, the NRC staff requested an electronic portal setup and provided a list of documents to be added to the online portal. The audit plan also indicated that the NRC may request information and audit meetings/interviews throughout the audit period.
The NRC staff has performed an initial review of the list of documents and is developing a list of audit questions.
The first set of audit questions were sent to you via email on April 28, 2023 (ML23118A388). The second set of questions are provided in the attachment. The staff is targeting June 13-15, 2023, to conduct the audit meetings via MS Teams to discuss the responses to the questions. To facilitate these discussions, please post the written response(s) to these questions on the online portal as they are completed. The planned agenda for the audit discussions will be provided in a separate message. Please contact me any time prior if a clarification discussion is needed.
Thank you,
Rich Guzman Sr. PM, Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Office: O-9C7 l Phone: (301) 415-1030 Richard.Guzman@nrc.gov
AUDIT QUESTIONS
LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO
ADOPT TSTF-505, REVISION 2 AND
LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED
CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS
CONSTELLATION ENERGY GENERATION, LLC
NINE MILE POINT NUCLEAR STATION, UNIT 1
DOCKET NO. 50-220
By letters dated December 15, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML22349A108 and ML22349A521), Constellation Energy Generation, LLC (CEG, the licensee) submitted two license amendment requests (LARs) for Nine Mile Point Nuclear Station, Unit 1 (NMP1). The proposed amendments would modify license DPR-63 and the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF)
Traveler 505 (TSTF-505), Provide Risk -informed Extended Completion Times, RITSTF Initiative 4b, and to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk -informed categorization and treatment of structures, systems and components for nuclear power reactors. On January 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan (ML23025A386) that conveyed intent to conduct a regulatory audit to support its review of the subject license amendments. Based on the commonalities between the LARs and subsequent overlap in technical content and review personnel, the NRC staff is conducting a combined audit that addresses both LARs.
The NRC staff has determined that the additional information is needed to support its review as shown in the following audit questions. The questions are ordered by number and identified by technical review branch/area and associated LAR (i.e., RICT for the TSTF-505 LAR and 50.69 for the LAR re: risk -informed categorization and treatment of SSCs).
Audit Questions - RICT
Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ML17317A256), states that the scope, level of detail, and technical adequacy of the probabilistic risk assessment (PRA) are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process. The NRCs SE for Nuclear Energy Institute (NEI) Topical Report NEI 06- 09, Revision 0-A, Risk -Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
Guidelines, Industry Guidance Document, dated November 6, 2006 (ML122860402) (hereafter NEI 06-09-A), and the NRCs Final Safety Evaluation for NEI 06- 09-A, dated May 17, 2007 (ML071200238), state that the PRA models should conform to the guidance in RG 1.200, Revision 1, An Approach for Determining the Technical Adequacy of Probabilistic Risk
Assessment Results for Risk-Informed Activities. The current version applicable to this LAR is RG 1.200, Revision 2 (ML090410014), which clarifies the current applicable American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard is ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA -S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. In RG 1.200, the quality of the PRA must be compatible with the safety implications of the proposed Technical Specification ( TS) change and the role the PRA plays in justifying the change.
RG 1.200 describes a peer review process using ASME/ANS RA-Sa-2009 as one acceptable approach for determining the technical acceptability of the PRA. The primary results of a peer review are the facts and observations (F&Os) recorded by the peer review team and the subsequent resolution of these F&Os. NEI 06 A states that the PRA shall meet Capability Category (CC)-II for the supporting requirements of the PRA standard, and any deviations from these capability categories relative to the RMTS program shall be justified.
Audit Question-06 (APLA - RICT) - Credit for FLEX Equipment and Actions
NRC memorandum dated May 6, 2022 1 provides the NRCs staff updated assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a probabilistic risk assessment (PRA) model in support of risk-informed decision-making in accordance with the guidance of RG 1.2002.
With regards to equipment failure probability, in the memorandum dated May 6, 2022, the NRC staff states in Conclusion 4:
Licensees that choose not to use the generic failure probabilities in PWROG-18042 to develop plant-specific failure probabilities for portable FLEX equipment modeled in PRAs used for risk-informed applications should submit a justification for the methods and probabilities used to the NRC for review and approval.
It appears that NUREG-6928 fixed equipment failure rates (with a 2x increase) were used as probabilities for FLEX portable equipment. It is unclear to the NRC staff how the NMP1 approach satisfies the concerns of Conclusion 4.
With regards to the uncertainty related to equipment failure probabilities, in the updated NRC memorandum, the NRC staff states in Conclusion 8:
PWROG-18043, Revision 1, notes that there was insufficient data to quantify the failure to load probabilities for portable diesel generators due to lack of detailed data. To account for the uncertainty in the testing activitieslicensees should ensure their preventive maintenance strategies include such testing and that the data reported provides this information. licensees should continue to assess the uncertainty in equipment failure rates and address or disposition it.
The NRC staff notes the results of a FLEX sensitivity study was provided by the licensee; however, it is unclear to the NRC staff if this sensitivity addresses the concerns of Conclusion 8.
1 U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, dated May 6, 2022 (ADAMS Accession No. ML22014A084).
2 U.S. Nuclear Regulatory Commission, Acceptability of Probabilistic Risk Assessment Results for Risk -Informed Activities, RG 1.200, Revision 3, December 2020 (ML20238B871).
For example, the licensees FLEX sensitivity study increased the FLEX maintenance unavailability factors by a factor of 10 (i.e., from 0.01 to 0.1). However, failure probability uncertainties associated with other FLEX failure modes may have a greater impact on risk (e.g.,
FLEX generator fails to run after first hour has a failure probability of 6.16E-2 in the licensees analysis and the uncertainty associated with this failure probability is likely to significantly increase the risk impact). The FLEX failure probabilities assumed in the licensees sensitivity study appears to be noticeably lower than that in PWROG-18042, which was approved by NRC (e.g., fail-to-start probability of the portable diesel-driven pump in PWROG-18042 is a factor of 17 higher than that assumed in the licensees sensitivity study). The results from the FLEX sensitivity study demonstrate a change of 28.8% in core damage frequency (CDF) value, which seems to indicate that FLEX failure probabilities are a key source of uncertainty. Also, the sensitivity study does not appear to address the impact of this uncertainty on the risk-informed completion time (RICT) calculations.
With regards to human reliability analysis (HRA), in the memorandum dated May 6, 2022, the NRC staff states, in part, in Conclusion 11:
EPRI [Electric Power Research Institute] 3002013018 provides updated detailed industry guidance for estimating the human error probabilities (HEPs) of the actions needed to implement mitigating strategies using portable equipment.
EPRI 3002013018 provides guidance that is acceptable to the NRC, with the clarifications below...
FLEX actions were developed by the licensee using the methodologies provided in EPRI 3002013018. However, it is unclear to the NRC staff how NMP1 analysis addressed the staff clarifications on the use of the EPRI guidance.
With regards to PRA u pgrade, the NRC staff states in the updated NRC memorandum in Conclusion 2:
Therefore, Conclusion 2 remains unchanged [that] for any new risk-informed application that has incorporated mitigating strategiesthe licensee should either perform a focused-scope peer review of the PRA model or demonstrate [that it does not meet the three criteria of an PRA u pgrade].
The NRC staff notes that the NMP1 PRA models appear to utilize updated industry guidance, and therefore, it is unclear whether the FLEX analysis is an PRA u pgrade for NMP1.
Given these observations, address the following:
a) Describe the methodology used to assess the failure probabilities of any modeled equipment credited in the licensee's mitigating strategies (i.e., FLEX). The discussion should include a justification of the rationale for parameter values, and how the uncertainties associated with the parameter values are considered in the RICT calculations in accordance with ASME/ANS RA-Sa-2009 3, as endorsed by RG 1.200 (e.g., supporting requirements for HLR-DA-D). This justification should
3 American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) PRA standard ASME/ANS RA-Sa -2009, "Addenda to ASME/ANS RA -S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Ap plications", February 2009, New York, NY (Copyright).
also disposition any significant differences between these FLEX parameter values and those generic failure probabilities in PWROG-18042.
-OR-
Alternatively, propose a mechanism to incorporate into the NMP1 PRA models used for RICT calculations updated FLEX parameter values prior to implementing the RICT program.
b) Provide a discussion detailing the methodology used to assess operator actions related to FLEX equipment and the licensee personnel that perform these actions. The discussion should include:
- i. A summary of how the licensee evaluated the impact of the NRC clarifications in memorandum dated May 6, 2022, with regards to using the EPRI 3002013018 FLEX HRA methodology.
ii. Provide updated FLEX HRA results, if required, to address the NRC clarifications.
iii. Provide justification that the use of the EPRI FLEX HRA methodology does not meet the definition of an PRA u pgrade as defined by RG 1.200.
Alternatively, if a justification is not provided, p ropose a mechanism to conduct a focused-scope peer review (FSPR) regarding incorporation of the EPRI FLEX HRA method for the NMP1 PRA models. Include in the mechanism to close out all F&Os that result from the FSPR prior to implementing the RMTS program.
c) Provide an updated assessment of the impact on risk values and uncertainty analysis provided in the LAR by FLEX equipment credited in NMP1s PRA models. This assessment should include, if required, any modifications to FLEX modeling based on the issues raised in this question. Include in this discussion:
(i) The impact of FLEX on any of the RICT values provided in Table E1-2 of the LAR and on the total baseline risk values provided in LAR Enclosure 5.
(ii) Discuss whether the uncertainty associated with FLEX modeling is a key source of uncertainty for the RMTS program.
If this uncertainty is "key," then describe and provide a basis for how this uncertainty will be addressed in the RMTS program (e.g., programmatic changes such as identification of additional RMAs, program restrictions, or the use of bounding analyses which address the impact of the uncertainty). If the programmatic changes include identification of additional RMAs, then (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in Section 2.3.4 of NEI 06 A; and (2) provide RMA examples that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty and explain how these RMAs are expected to reduce the risk associated with this uncertainty.
Audit Question-07 (APLA - RICT) - RG 1.200, Revision 1 Gap Assessment
NEI 06-09-A requires that a licensees PRA be of sufficient quality and level of detail to support the RMTS process, and that the PRA r eflect the current plant configuration and operating practices, and applicable plant and industry operational experience.
of t he LAR states that the internal events and internal flooding peer review was conducted against RG 1.200, Revision 1 that endorsed the 2005 PRA standard that preceded the 2009 PRA standard as endorsed by RG 1.200, Revisions 2 and 3. The NRC staff notes that the licensee provided a recent gap assessment between the two PRA standards. The assessment highlights several new requirements made in Revision 2 of RG 1.200. However, there appears to be no disposition to determine whether these new requirements constituted a PRA u pgrade to NMP1s PRA models.
The NRC staff notes that the 2009 ASME/ANS PRA standard defines a PRA upgrade as either: a new method applied to the NMP1 PRA models, significant changes to accident sequences, or significant changes to accident progression sequences. It is unclear to the NRC staff whether any PRA u pgrades subsequent to the 2008 peer review were incorporated into NMP1s internal events or internal flooding PRA models in order to meet the updated requirements of the 2009 PRA standard.
a) Clarify, and describe, if any PRA u pgrades were made to the internal events and internal flooding PRA models since the 2008 peer review.
b) For each identified PRA u pgrade, confirm that each u pgrade has been peer reviewed and the associated findings, if any, have been closed in accordance with NRC approved processes.
-OR-Propose a mechanism to perform a peer review regarding incorporation of any new method incorporated into the NMP1 PRA models. Include in the mechanism to close out all findings that result from the peer review prior to implementing the RMTS program.
Audit Question-08 (APLA - RICT) - Open F&Os Disposition
NEI 06-09-A requires that a licensees PRA be of sufficient quality and level of detail to support the RMTS process, and that the PRA reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.
The disposition to open F&O 1-2 in Enclosure 2 of the LAR states that a human error probability (HEP) ZQDHR_DEPOPERATO represents multiple human failure events (HFEs), which constitutes a joint HEP (JHEP). However, it appears that based on the long timeframe of these three actions, this HEP was excluded from the normal dependency analysis process. The NRC staff notes that the HRA Calculator Dependency Decision Tree can result in low or moderate dependency levels between HFEs with significant times between the actions.
a) Provide justification that the exclusion of the ZQDHR_DEPOPERATO HFE from the dependency analysis and the assumption of the three HFEs (ZQDHRFDEPOPERATO, ZQDHRADEPOPERATO, and ZQDHRIDEPOPERATO) contain no dependency relationship does not significantly impact any RICT calculation.
-OR-
b) Propose a mechanism to include the aforementioned HFEs in the dependency analysis and incorporate into the NMP1 PRA models prior to implementing the RMTS program.
Audit Question-09 (APLA - RICT) - In-Scope LCOs and Corresponding PRA Modeling
The NRCs safety evaluation for NEI 06- 09-A specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis for when there is a difference. Table E1-1 of LAR Enclosure 1 identifies each Limiting Condition for Operation (LCO) in the TSs proposed for inclusion in the RICT program. The table also describes whether the systems and components covered by the LCO are modeled in the PRA and, if so, presents both the design success criteria and PRA success criteria. For certain LCOs, the table explains that the associated structures, systems, and components (SSCs) are not modeled in the PRAs but will be represented using a surrogate event that fails the function performed by the SSC. For some LCOs, the LAR did not provide an adequate description for the NRC staff to conclude that the PRA modeling will be sufficient.
Regarding TS LCO 3.3.4.B, Table E1-1 states that, for any primary containment isolation valves (PCIVs) not modeled, a pre-existing small leak event that is modeled will be used as a surrogate. It is unclear to the NRC staff which pathways will be used for each affected function.
a) Clarify which pre -existing small leak event will be used as a surrogate for each of the system isolation functions affected.
b) Provide justification that the surrogate bounds each of the isolation functions.
Audit Question-10 (APLA - RICT) - Performance Monitoring
The NRC SE for NEI 06- 09-A, states, [t]he impact of the proposed change should be monitored using performance measurement strategies. NEI 06- 09-A considers the use of NUMARC 93- 01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (ML18120A069), as endorsed by RG 1.160, Revision 4 (ML18220B281), for the implementation of the Maintenance Rule. NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.
In addition, the NEI 06- 09-A methodology satisfies the five key safety principles specified in RG 1.177, Revision 2 ( ML20164A034) relative to the risk impact due to the application of a RICT. Moreover, NRC staff position C.3.2 provided in RG 1.177 for meeting the fifth key safety principle acknowledges the use of performance criteria to assess degradation of operational safety over a period. It is unclear how the licensees RICT program captures performance monitoring for the SSCs within-scope of the RMTS program. Therefore:
a) Confirm that the NMP1 Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in NUMARC 93- 01, as endorsed by RG 1.160.
b) Alternatively, describe the approach or method used by NMP1 for SSC performance monitoring, as described in NRC staff position C.3.2 of RG 1.177, Revision 2, for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative or quantitative), along with the appropriate risk metrics, and explain how the approach and criteria demonstrate the intent to monitor the potential degradation of SSCs in accordance with the NRC SE for NEI 06- 09-A.
Audit Question-11 (APLA - RICT) - Consideration of Shared Systems in Calculation of a RICT
RG 1.200, Revision 2, states, [t]he base PRA serves as the foundational representation of the as-built and as-operated plant necessary to support an application.
The LAR does not appear to address the existence of crossties between units. However, the NRC staff has reviewed system documents in the portal that have shared systems. The NRC staff notes that for some of these systems, it appears the sharing of a system is not consistent among units. It appears that some operational aspects, such as alternate alignments, were excluded from the PRA models. For multiunit events (e.g., loss of offsite power and seismic events), credit for a shared system may be limited to one unit.
Clarify what systems are shared, how they are shared, whether they can support the other unit in an accident. Explain how the shared systems are credited for each unit in the PRA models.
This discussion should also address the following:
a) Identify systems that can be cross-tied to another unit. Discuss any differences among the units sharing these systems.
b) Explain how shared systems credited in the real-time risk model that support the RICT calculations are modeled for each unit in a multiunit event. Include in this discussion what aspects of these systems were excluded from the PRA model(s) and why these exclusions do not impact the application.
c) If the impact of events that can create a concurrent demand for a system shared by multiple units and credited in the real-time risk model is not addressed, explain why this modeling exclusion does not have a significant impact on the RICT calculations.
Audit Question-12 (APLA - RICT) - Digital Instrumentation and Control Modeling
Concerning the quality of the PRA model, NEI 06- 09-A states that RG 1.174 and RG 1.200 define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change.
Regarding digital instrumentation and control (I&C), the NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed regulatory applications. In addition, known modeling challenges exist, such as the lack of industry data for digital I&C components, the difference between digital and analog system failure modes, and the complexities associated with modeling software failures including common-cause software failures. Also, though reliability data from vendor tests may be available, this source of data is not a substitute for in-the-field operational data. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT program. Therefore, address the following:
a) Clarify whether digital I&C systems are credited in the PRA models that will be used in the RICT program.
b) If digital I&C systems are credited in the PRA models that will be used in the RICT program, provide justification (e.g., describe and provide the results of a sensitivity study) that demonstrates the modeling uncertainty associated with crediting digital I&C systems has an inconsequential impact on the RICT calculations.
Alternatively, if a justification is not provided, identify which LCOs are determined to be impacted by digital I&C systems modeling for which risk management actions (RMAs) will be applied during a RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation require additional RMAs.
Audit Question-13 (APLA - RICT) - Impact of Seasonal Variations
The Tier 3 assessment in RG 1.177 stipulates that a licensee should develop a program that ensures the risk impact of out of service equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06 A and its associated NRC safety evaluation state that, for the impact of seasonal changes, either conservative assumptions should be made, or the PRA should be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration.
of the LAR states that outside temperatures impact on service water pumps were evaluated and addressed. However, it does not appear to specify the modeling adjustments needed to account for seasonal and time of cycle dependencies and what kind of adjustments will be made. Therefore, address the following to clarify the treatment of seasonal and time of cycle variations:
a) Explain how the RICT calculations address changes in PRA data points, basic events, and SSC operability constraints as a result of extreme weather conditions, seasonal variations, other environmental factors, or time of cycle. Also, explain how these adjustments are made in the configuration risk management program (CRMP) model and how this approach is consistent with the guidance in NEI 06- 09-A and its associated NRC final SE.
b) Describe the criteria used to determine when PRA adjustments due to extreme weather conditions, seasonal variations, other environmental factors, or time of cycle variations need to be made in the CRMP model and what mechanism initiates these changes.
Audit Question-14 (APLA - RICT) - PRA Update Process
Section 2.3.4 of NEI 06-09-A specifies, criteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations.
LAR Enclosure 7 states that if a plant change or a discovered condition is identified and can have significant impact on the RICT calculations, then an unscheduled update of the PRA models will be implemented. More specifically, the LAR states that if the plant changes meet specific criteria defined in the plant PRA and update procedures, including criteria associated with consideration of the cumulative risk impact, then the change will be incorporated into applicable PRA models without waiting for the next periodic PRA update. The LAR does not explain under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedures that will be used to initiate the update.
Considering these observations, describe the conditions under which an unscheduled PRA update (i.e., more than once every two refueling cycles) would be performed and the criteria that would be used to require a PRA update. In the response, define what is meant by significant impact to the RICT Program calculations.
Audit Question-15 (APLA - RICT) - Open Phase Condition
Section C.1.4 of RG 1.200 states the base (e.g., Model of Record) PRA is to represent the as-built, as-operated plant to the extent needed to support the application. The licensee is to have a process that identifies updated plant information that necessitate changes to the base PRA model.
In response to the January 30, 2012, Open Phase Condition (OPC) event at the Byron Generating Station, the NRC issued Bulletin 2012- 01 4. As part of the initial Voluntary Industry Initiative (VII) for mitigation of the potential for the occurrence of an OPC in electrical switchyards 5, licensees have made the addition of an Open Phase Isolation System (OPIS).
As per SRM-SECY-16-0068 6, the NRC staff was directed to ensure that licensees have appropriately implemented OPIS and that licensing bases have been updated accordingly. From the revised voluntary initiative 7 and resulting industry guidance in NEI 19-02 8 on estimating OPC and OPIS risk, it is understood that the risk impact of an OPC can vary widely dependent on electrical switchyard configuration and design. Considering these observations, provide the following information:
4 U.S. NRC Bulletin 2012- 01, Design Vulnerability in Electric Power System (ML12074A115).
5 Anthony R. Pietrangelo to Mark A. Satorius, Ltr re: Industry Initiative on Open Phase Condition - Functioning of Important-to -Safety Structures, Systems and Components (SSCs), dated October 9, 2013 (ML13333A147).
6 U.S. NRC SRM-SECY-16-0068, Interim Enforcement Policy for Open Phase Conditions in Electric Power Systems for Operating Reactors, dated March 9, 2017 (ML17068A297).
7 Doug True to Ho Nieh, Ltr re: Industry Initiative on Open Phase Condition, Revision 3, dated June 6, 2019 (ML19163A176).
8 Nuclear Energy Institute (NEI) 19-02, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights, Revision 0, April 2019 (ML19122A321).
a) For NMP1, discuss the evaluation of the risk impact associated with OPC events including the likelihood of OPC initiating plant trips and the impact of those trips on PRA-modeled SSCs. Also, explain whether an OPIS has been installed and if it has been installed, then discuss its functionality and any operator actions needed to operate the system or needed in response to the system.
b) Clarify whether any installed OPIS equipment and associated operator actions are credited in the PRAs that support this application. If OPIS equipment and associated operator actions are credited, then provide the following information:
ii. Describe the impact that this treatment, if any, has on key assumptions and sources of uncertainty for the RICT program.
iii. Discuss HRA methods and assumptions used for crediting OPIS alarm manual response.
iv. Discuss how OPC related scenarios are modelled for non-internal event scenarios such as internal floods, fire, and seismic.
- v. Regarding inadvertent OPIS actuation:
- Explain whether scenarios regarding inadvertent actuation of the OPIS, if applicable, are included in the PRA models that support the RICT calculations.
- If inadvertent OPIS actuation scenarios are not included in the PRA models, then provide justification that the exclusion of this inadvertent actuation does not impact the RICT calculations.
c) If OPC and OPIS are not included in the application PRA models (whether OPIS equipment is installed or not), then provide justification that the exclusion of this failure mode and mitigating system does not impact the RICT calculations.
d) As an alternative to Part (c), propose a mechanism to ensure that OPC-related scenarios are incorporated into the application PRA models prior to implementing the RICT program.
Audit Question-16 (APLA - RICT) - Exclusion of High Winds Penalty in RICT Estimates
Section 2.3.1, Item 7 of NEI 06- 09-A, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06- 09 states that [w]here PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
Table E1-2 of Enclosure 1 of the LAR provides RICT estimates for TS actions proposed to be in the scope of the RICT program. Note 1 of the table states that the calculations are based on internal events, internal flood, and internal fire PRA model calculations with seismic penalties.
The NRC staff notes that Section 4 of Enclosure 4 identifies the high wind hazard risk is significant to overall plant risk and calculates two different penalties that are listed in Table E5-1.
It is unclear to the NRC staff whether the RICT values of Table E1-2 include the high winds penalties in the calculation.
a) Confirm that the RICT values provided in Table E1-2 of Enclosure 1 of the LAR were calculated considering the high wind penalties.
b) Clarify how the different values of high wind penalties developed in Enclosure 4 to the LAR are applicable to each RICT.
c) If high wind penalties were not provided in Table E1-2, include the high winds penalties, and update the table.
Audit Question-17 (APLC - RICT) - SSC Design Wind Capacity
Section 2.3.1, Item 7 of NEI 06- 09-A, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06 -09 states that [w]here PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
The high winds penalty approach is used to quantify the risk impact and to support the RICT evaluation. Table E4-6 of Enclosure 4 to the LAR provides the design wind speed capacities for several structures, including the emergency diesel generator (EDG) Building and Board Rooms.
The table states a design wind speed capacity of 175 mph for the EDG Building. The NRC staff noted that Table E4-4 states that the DG board room wall capacity could not be confirmed, and the limiting DG building roof capacity value of 175 mph was used in the calculation. It is unclear to the NRC staff the criteria used to determine that the DG board room wall wind capacity is not the limiting part of the EDG Building and Board Rooms and if the decision is conservative for the high winds penalty calculation.
a) Discuss how it was determined that the DG board room wall wind capacity is not the limiting component for the EDG building. Include in this discussion justification that the selection of the DG building roof as the limiting component is a conservative assumption for the high winds penalty calculation.
b) Demonstrate, using an approach such as a sensitivity study, that the use of the EDG building roof capacity as the limiting component does not significantly impact any RICT calculation.
Audit Question-18 (APLC - RICT) - Design Wind Speed Capacity Parameter and Fragility Calculations
Section 2.3.1, Item 7 of NEI 06- 09-A, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06- 09 states that [w]here PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
Table E4-6 of Enclosure 4 to the LAR provides the design wind speed capacities for several key structures. The table includes two columns for the same parameter, Vd. The first column in Table E4-6 defines Vd as the design wind speed capacity and the second column defines it as the 3-second gust wind speed. The licensee used a formula in the EPRI report 3002003107 to calculate the median velocity V m. Section 4.5.1 of EPRI report 3002003107 defines Vd as the design wind speed capacity and it is used as the variable in the calculation of median velocity.
The EPRI report provides no mention of the 3-second gust wind speed. Enclosure 4 of the LAR states The hazard curves from data in NUREG/CR-4461 and the ASCE Hazard Tool are based on the 3-second gust wind speeds. To reconcile the difference in time domains, the fastest mile wind speeds were converted to 3-second gust wind speeds.
a) Justify defining Vd as both the design wind speed capacity and the 3-second gust wind speed in Table E4-6.
b) Explain why the 3-second gust wind speed was chosen as the parameter for the licensees wind capacity and fragility calculations, when the 3-second gust wind speed was not defined in EPRI report 3002003107 for use in the calculation of Vm.
Note (1) of Table E4-6 states, A more detailed fragility analysis was performed to develop the screenhouse parameters, and shows Screenhouse V m = 239 mph with r = 0.08 and u = 0.11.
The NRC staff noted that a licensees calculation in N1-MISC-021 shows missile impact effects on the screenhouse. The median wind capacity V m was estimated as V m = (1.1
- 1.2
- Vd)0.5, which is different from the formula in EPRI report 3002003107 used for other SSCs, but it is listed in Table 4-1 of the EPRI report. However, N1-MISC-021 changed Vd to Vd,hc in the formula, and let V d,hc = 1.780.5 Vd, and used V d = 150 mph, which is not consistent with 144 mph in Table E4-6 for Screenhouse.
c) Explain what is the V d,hc and why it replaces V d in the formula used for calculation of the median velocity Vm. Why Vd = 150 mph is used, instead of 144 mph in Table E4-6 for Screenhouse?
d) Discuss, with justification, the use of a different V m calculation formula for the screenhouse compared to the other SSCs.
Note (2) of Table E4-6 shows a generic c = 0.2 was selected for all SSCs, except for the Screenhouse. Table 4-2 of ERPI report 3003003107 shows ranges of s, d, and p, and by using Eq. 4-10 of the EPRI report, r or u = (s2 + d2 + p2)0.5, can be obtained in a range of 0.2 to 0.5. Then, c should be calculated from c = (r2 + u2)0.5, which should be higher than r or u in a range of 0.2 to 0.5.
e) Discuss how the c value of 0.2 was selected for the NMP1 high winds structure fragility analysis. Include in this discussion justification that the value of 0.2 results in a conservative high winds penalty.
f) Demonstrate, using an approach such as a sensitivity study, that the value of 0.2 for the c parameter does not significantly impact any RICT calculation.
g) If any values were changed due to questions (a) to ( f), provide the updated high wind penalty factors to be used in the RICT program.
Audit Question-19 (APLC - RICT) - Calculation of the Main Stack Contribution to High Wind CDF
Section 2.3.1, Item 7 of NEI 06- 09-A, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06- 09 states that [w]here PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
According to Table E4-7 of Enclosure 4 to the LAR, the Main Stack contributes to 48% of the wind pressure CDF penalty due to failure of key structures at NMP1. The LAR states the stack is the most significant contributor due to its relatively low fragility The LAR also states For the stack, only 31% of the stack failure frequency results in failure of a key structure (EDG building, screenhouse, or reactor building.) The calculation of the 31% stack failure frequency is provided in N1-MISC-021, NMP1 High Winds and Tornado Risk Assessment. The calculation assumes that the stack can fall in any direction with equal probability and only a certain fraction of the stack failure directions will result in the failure of a key structure. From the spatial diagram in N1-MISC-021, it appears that the key structures all lay within approximately a 190- degree arc of the main stack.
Justify that the omission of site wind direction data for determining the stack failure direction probability for this important contributor to high wind CDF penalty is conservative or include wind direction data for the site into the calculations.
Audit Question-20 (APLC - RICT) - NMP1 Tornado Missile Vulnerability Analysis
Section 2.3.1, Item 7 of NEI 06- 09-A, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06- 09 states that [w]here PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
of the LAR provides CDF and large early release frequency (LERF) risk values for NMP1s tornado missile assessment. However, it does not provide details on the methodology used for this analysis. The NRC staff noted that the NMP1 high winds assessment states that the risk is evaluated using a simple tornado missile target model, based on the methods in NEI 17- 02 9. It is unclear to the NRC staff if the NMP1 tornado risk analysis consisted of any deviations from the NEI 17- 02 methodology and data.
a) Describe the simple tornado missile target model, based on the methods in NEI 17-02 used for the licensees tornado missile risk analysis.
b) Clarify if the NMP1 tornado missile risk analysis deviated from the NEI 17- 02 methods and data. Provide details, if any, for each deviation.
c) Provide justification for the use of all deviations, if any, in calculating NMP1s tornado missile risk. Include in this discussion how the deviations do not significantly impact RICT calculations.
9 NEI 17-02, Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document, Revision 1B, dated September 2018 (ML18262A328).
Audit Question-21 (APLC - RICT) - High Wind Risk for Maintenance Configurations
Section 2.3.1, Item 7 of NEI 06-09-A, states that the impact of other external events risk shall be addressed in the RMTS program and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs SE for NEI 06- 09 states that [w]here PRA models are not available, conservative, or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
Under the subsection of High Wind Risk for Maintenance Configurations, the licensee provided one DG or respective DG board unavailable delta CDFTM = 6.0E-6 /yr, and one train of EC or associated equipment unavailable delta CDFTM =1.9E-6 /yr. It is not clear to the NRC staff how these values were calculated.
Explain how these delta CDF values are calculated and the calculation method used.
Audit Question-22 (APLA - 50.69) - Credit for FLEX Equipment and Actions
NRC memorandum dated May 6, 202210 provides the NRCs staff updated assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a probabilistic risk assessment (PRA) model in support of risk -informed decisionmaking in accordance with the guidance of RG 1.20011.
With regards to equipment failure probability, in the memorandum dated May 6, 2022, the NRC staff states in Conclusion 4:
Licensees that choose not to use the generic failure probabilities in PWROG -
18042 to develop plant-specific failure probabilities for portable FLEX equipment modeled in PRAs used for risk-informed applications should submit a justification for the methods and probabilities used to the NRC for review and approval.
It appears that NUREG-6928 fixed equipment failure rates (with a 2x increase) were used as probabilities for FLEX portable equipment. It is unclear to the NRC staff how the NMP1 approach satisfies the concerns of Conclusion 4.
With regards to the uncertainty related to equipment failure probabilities, in the updated NRC memorandum, the NRC staff states in Conclusion 8:
PWROG-18043, Revision 1, notes that there was insufficient data to quantify the failure to load probabilities for portable diesel generators due to lack of detailed data. To account for the uncertainty in the testing activitieslicensees should ensure their preventive maintenance strategies include such testing and that the data reported provides this information. licensees should continue to assess the uncertainty in equipment failure rates and address or disposition it.
The NRC staff notes the results of a FLEX sensitivity study was provided by the licensee; however, it is unclear to the NRC staff if this sensitivity addresses the concerns of Conclusion 8.
For example, the licensees FLEX sensitivity study increased the FLEX maintenance unavailability factors by a factor of 10 (i.e., from 0.01 to 0.1). However, failure probability uncertainties associated with other FLEX failure modes may have a greater impact on risk (e.g.,
FLEX generator fails to run after first hour has a failure probability of 6.16E-2 in the licensees analysis and the uncertainty associated with this failure probability is likely to significantly increase the risk impact). The FLEX failure probabilities assumed in the licensees sensitivity study appears to be noticeably lower than that in PWROG-18042, which was approved by NRC (e.g., fail-to-start probability of the portable diesel-driven pump in PWROG-18042 is a factor of 17 higher than that assumed in the licensees sensitivity study). The results from the FLEX sensitivity study demonstrate a change of 28.8% in CDF value, which seems to indicate that FLEX failure probabilities are a key source of uncertainty.
With regards to human reliability analysis (HRA), in the memorandum dated May 6, 2022, the NRC staff states, in part, in Conclusion 11:
10 U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, dated May 6, 2022 (ML22014A084).
11 U.S. Nuclear Regulatory Commission, Acceptability of Probabilistic Risk Assessment Results for Risk -Informed Activities, RG 1.200, Revision 3, December 2020 (ML20238B871).
EPRI 3002013018 provides updated detailed industry guidance for estimating the human error probabilities (HEPs) of the actions needed to implement mitigating strategies using portable equipment. EPRI 3002013018 provides guidance that is acceptable to the NRC, with the clarifications below...
FLEX actions were developed by the licensee using the methodologies provided in EPRI 3002013018. However, it is unclear to the NRC staff how NMP1 analysis addressed the staff clarifications on the use of the EPRI guidance.
With regards to PRA upgrade, the NRC staff states in the updated NRC memorandum in Conclusion 2:
Therefore, Conclusion 2 remains unchanged [that] for any new risk-informed application that has incorporated mitigating strategiesthe licensee should either perform a focused-scope peer review of the PRA model or demonstrate [that it does not meet the three criteria of an PRA upgrade].
The NRC staff notes that the NMP1 PRA models appear to utilize updated industry guidance, and therefore, it is unclear whether the FLEX analysis is an PRA upgrade for NMP1.
Given these observations, address the following:
a) Describe the methodology used to assess the failure probabilities of any modeled equipment credited in the licensee's mitigating strategies (i.e., FLEX). The discussion should include a justification of the rationale for parameter values, and how the uncertainties associated with the parameter values are considered in 50.69 categorization in accordance with ASME/ANS RA-Sa-2009 12, as endorsed by RG 1.200 (e.g., supporting requirements for HLR-DA-D). This justification should also disposition any significant differences between these FLEX parameter values and those generic failure probabilities in PWROG-18042.
-OR-
Alternatively, propose a mechanism to incorporate into the NMP1 PRA models updated FLEX parameter values prior to implementing 50.69.
b) Provide a discussion detailing the methodology used to assess operator actions related to FLEX equipment and the licensee personnel that perform these actions. The discussion should include:
- i. A summary of how the licensee evaluated the impact of the NRC clarifications in memorandum dated May 6, 2022, with regards to using the EPRI 3002013018 FLEX HRA methodology.
ii. Provide updated FLEX HRA results, if required, to address the NRC clarifications.
iii. Provide justification that the use of the EPRI FLEX HRA methodology does not
12 American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) PRA standard ASME/ANS RA-Sa -2009, "Addenda to ASME/ANS RA -S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications", February 2009, New York, NY (Copyright).
meet the definition of an PRA upgrade as defined by RG 1.200.
Alternatively, if a justification is not provided, propose a mechanism to conduct a focused-scope peer review (FSPR) regarding incorporation of the EPRI FLEX HRA method for the NMP1 PRA models. Include in the mechanism to close out all F&Os that result from the FSPR prior to implementing 50.69.
c) Provide an updated assessment of the impact on risk values and uncertainty analysis provided in the LAR by FLEX equipment credited in NMP1s PRA models. This assessment should include, if required, any modifications to FLEX modeling based on the issues raised in this question. Include in this discussion, the impact of FLEX on the total baseline risk values provided in LAR and whether the uncertainty associated with FLEX modeling is a key source of uncertainty for 50.69. If this uncertainty is "key," then describe and provide a basis for how this uncertainty will be addressed for 50.69 categorization.
Audit Question-23 (APLA - 50.69) - RG 1.200, Revision 1 Gap Assessment
NEI 00-04 requires that a licensees PRA be of sufficient quality and level of detail to support the categorization process, and that the PRA reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.
The LAR states that the internal events and internal flooding peer review was conducted against RG 1.200, Revision 1 that endorsed the 2005 PRA standard that preceded the 2009 PRA standard as endorsed by RG 1.200, Revisions 2 and 3. The NRC staff notes that the licensee provided a recent gap assessment between the two PRA standards. The assessment highlights several new requirements made in Revision 2 of RG 1.200. However, there appears to be no disposition to determine whether these new requirements constituted a PRA upgrade to NMP1s PRA models.
The NRC staff notes that the 2009 ASME/ANS PRA standard defines a PRA upgrade as either:
a new method applied to the NMP1 PRA models, significant changes to accident sequences, or significant changes to accident progression sequences. It is unclear to the NRC staff whether any PRA upgrades subsequent to the 2008 peer review were incorporated into NMP1s internal events or internal flooding PRA models in order to meet the updated requirements of the 2009 PRA standard.
a) Clarify, and describe, if any PRA upgrades were made to the internal events and internal flooding PRA models since the 2008 peer review.
b) For each identified PRA upgrade, confirm that each upgrade has been peer reviewed and the associated findings, if any, have been closed in accordance with NRC approved processes.
-OR-
Propose a mechanism to perform a peer review regarding incorporation of any new method incorporated into the NMP1 PRA models. Include in the mechanism to close out all findings that result from the peer review prior to implementing 50.69.
Audit Question-24 (APLA - 50.69) - Open F&Os Disposition
NEI 00-04 requires that a licensees PRA be of sufficient quality and level of detail to support the RMTS process, and that the PRA reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.
The disposition to open F&O 1-2 in the LAR states that a human error probability (HEP)
ZQDHR_DEPOPERATO represents multiple human failure events (HFEs), which constitutes a joint HEP (JHEP). However, it appears that based on the long timeframe of these three actions, this HEP was excluded from the normal dependency analysis process. The NRC staff notes that the HRA Calculator Dependency Decision Tree can result in low or moderate dependency levels between HFEs with significant times between the actions.
a) Provide justification that the exclusion of the ZQDHR_DEPOPERATO HFE from the dependency analysis and the assumption of the three HFEs (ZQDHRFDEPOPERATO, ZQDHRADEPOPERATO, and ZQDHRIDEPOPERATO) contain no dependency relationship would not significantly impact any SSC categorization.
-OR-
b) Propose a mechanism to include the aforementioned HFEs in the dependency analysis and incorporate into the NMP1 PRA models prior to implementing 50.69.
Audit Question-25 (APLC - 50.69) - Overall Use of NEI 00-04 Figure 5-6
NEI 00-04 13 Figure 5-6 provides guidance to be used to determine SSC safety significance.
The same document states, in part, that if it can be shown that the component either did not participate in any screened scenarios or, even if credit for the component was removed, the screened scenario would not become unscreened, then it is considered a candidate for the LSS category.
In Section 3.2.4 of the LAR the licensee stated that [a]ll other external hazards, except for seismic, were screened for applicability to NMP1 per a plant-specific evaluation... However, the licensee discussed the extreme high winds and tornados, which is not screened, in the same section.
lists all hazards as screened except for internal events, internal flooding, internal fire with PRA models, seismic hazard with an alternate approach, and the extreme winds and tornados with high wind safe shutdown equipment list (HWSSEL). The guidance in NEI 00- 04, Figure 5-6 regarding SSCs that play a role in screening a hazard is not discussed in the of the LAR. Therefore, it appears to the NRC staff based on this description that at the time an SSC is categorized it will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard because that evaluation has already been made. The NRC staff notes that plant changes, plant or industry operational experience, updates to hazard frequency information, and identified errors or limitations in the PRA models could potentially impact the conclusion that an SSC is not needed to screen an external hazard.
a) Clarify whether or not an SSC will be evaluated during categorization of the SSC using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard.
b) If an SSC will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard at the time of categorization because that evaluation has already been made, then explain how plant changes, plant or industry operational experience, updated information in hazard frequencies and identified errors or limitations that could change that decision are addressed.
Audit Question-26 (APLC - 50.69) - Development of the High Winds Safe Shutdown Equipment List (HWSSEL)
Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs.
Section 3.2.4 of the LAR states, the wind pressure / missile hazard safety significance process uses a High Winds Safe Shutdown Equipment List (HWSSEL). It further states that this list will be developed. The LAR continues by stating that the proposed approach will identify SSCs associated with this hazards safe shutdown function and barriers to be high safe significant (HSS).
13 NEI 00-04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline", July 2005 (ML052910035).
It is not clear when the list will be developed, as it was not listed as one of the categorization prerequires in Attachment 1. In addition, the licensee discussed the equipment that fulfills a HWSSEL function, which is not shown in Table 3-1 Categorization Evaluation Summary, that is, the high winds and tornado hazards are not mentioned. If the high winds and tornados are considered as a part of other external hazards, its evaluation level is for component only, not including its function from the HWSSEL, as discussed in Section 3.2.4.
Section 3.3 of NEI 00- 04, Revision 0 provides limited guidance for determining the technical adequacy attributes required for these types of analyses for this specific application. RG 1.201, Revision 1 states that as part of the plant -specific application requesting to implement §50.69, the licensee or applicant will provide the bases supporting the technical adequacy of itsnon-PRA-type analyses for this application.
Address the following regarding the proposed HWSSEL approach:
a) Explain when the HWSSEL will be developed. Justify why it is not included as one of the categorizations prerequires in Attachment 1. If it cannot be justified, provide a mechanism, such as inclusion in Attachment 1, to ensure that the HWSSEL will be developed prior to initiation of the 10 CFR 50.69 program at NMP1.
b) Table 3-1, Categorization Evaluation Summary lists the different methods for categorization and the rules under which the IDP can or cannot change HSS to LSS.
The high winds initiator is not listed separately in this Table. Because high winds are evaluated differently than all other external hazards at NMP1, explain why this hazard is not specifically provided in the table or revise the table to provide high winds separately to show that candidate HWSSEL SSCs cannot be changed by the IDP, and that the evaluation level can be at the function level as well.
c) Provide justification that the HWSSEL method meets the expectations in the Statements of Consideration for 10 CFR 50.69 that non-PRA methods used in the categorization process are conservative. In the justification, identify the industry assessments referenced in the LAR and summarize the industry evaluations and results that support the conclusion that the NMP1 proposed approach to use the HWSSEL is conservative.
d) Provide details on the methodology to be used to develop the HWSSEL. Include in the response how the high wind related SSCs would be processed using this methodology.
e) High wind / tornado / tornado missile protection actions can be credited if they are feasible, but PRA actions generally are not credited unless they are proceduralized and have a failure probability assigned. Some feasible actions have a high failure probability.
Further, certain actions need to be taken outside of Seismic Category-I structures and need expanded time for feasibility due to high wind/tornado/tornado missile conditions.
- i. Justify the consideration of operator actions in the HWSSEL without a detailed evaluation and explain how the feasibility and failure probability of operator actions (which could be high) is incorporated in the analysis for determining SSCs in the HWSSEL?
ii. Explain the mapping that will be performed to assign any operator actions deemed feasible to SSCs for exclusion from the HWSSEL?
f) The terminator oval in Figure 3.2 of the LAR for the No decisions states, [c]andidate Low Safety Significant. Provide a description to clarify what SSCs would be categorized as LSS using the flow chart in Figure 3.2 that are not considered in the high winds /
tornado / tornado missile safe shutdown analysis. In the description, provide examples of SSCs that would be categorized as LSS for this terminator oval.
Audit Question-27 (APLC - 50.69) - External Flooding (Local Intense Precipitation)
Paragraph 50.69(b)(2)(ii) of 10 CFR requires that the quality and level of detail of the systematic processes that evaluate the plant for external events during operation is adequate for the categorization of SSCs.
In Attachment 4 of the LAR External Hazards Screening, the licensee stated that an analysis (S0FLOODF002) was performed to calculate the volume of water that will leak into the buildings during the Local Intense Precipitation ( LIP) without any temporary flood barriers installed. It appears this statement may not be consistent with the assumption 5.1 and 5.4 of the analysis, S0FLOODF002, where a temporary flood barrier either may need to be or must be installed.
The licensee concluded that water intrusion through normally closed exterior doors that would ultimately accumulate on the 250 ft elevation to a maximum depth of 31 inches; as a result, no safety related equipment will be impacted. Therefore, the licensee determined that temporary flood barriers were not required to screen this hazard, only closing of exterior doors were required.
The S0FLOODF002 analysis appears to address the 72-hour Probable Maximum Precipitation (PMP) event. It is unclear to the NRC staff the applicability of the 72-hour PMP analysis for the 24-LIP event given the NMP1 FE analysis. The NRC staff notes that in Section 3.2.1 of its September 20, 2017, flooding focused evaluation14, the following flood protection barriers will be installed for the following areas to address LIP:
- Battery Board Room
- Foam Room
- Auxiliary Control Room
- Reactor Building
It appears that the temporary flood barriers are required to address LIP events, which is not consistent with the S0FLOODF002 analysis. Therefore, address the following:
a) Justify the applicability of the S0FLOODF002 analysis to the LIP event. Include in this discussion any plant or procedure modifications since the NMP1 FE submittal that impacted the use of flood barriers to address the LIP event.
b) Clarify that during a LIP event no flood barriers are necessary to address the areas identified in the NRC staffs 2017 flooding focused evaluation. If barriers are necessary, identify the corresponding areas and explain how the barriers will be treated under the proposed categorization program.
14 Gibson, L.K., U.S. Nuclear Regulatory Commission to Hanson, B.C., Exelon Generation Company, LLC, Nine Mile Point Nuclear Station, Units 1 and 2 - Staff Assessment of Flooding Focused Evaluation (CAC NOS. MG0087 and MG0088), dated September 20, 2017 (ML17251A045).
c) Attachment 4 to the LAR provides the Table of External Hazard Screening. In the External Flooding section of the table, it states the LIP mechanism requires several normally closed exterior doors to slow water intrusion for screening. These doors will be considered high safety significant (HSS) since they are credited for screening the external flood hazard. Provide a list of the specific exterior doors that will be assigned HSS since they are credited for screening the external flood hazard (in accordance with Figure 5-6 in NEI 00-04).
Audit Question-28 (APLC - 50.69) - Screening of Snowfall Risk
Section 2.3.1, Item 7, of NEI 06 -09, Revision 0-A, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by documenting prior to the RMTS program that external events that are not modeled in the PRA are not significant contributors to configuration risk. The SE for NEI 06- 09 states that
[o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk.
LAR Attachment 4, External Hazard Screening, indicates that criterion C5 was used to screen the snow hazard. Criterion C5 states the event develops slowly, allowing adequate time to eliminate or mitigate the threat.
The NMP1 IPEEE states NMP1 drawings indicate that NMP1 has capabilities beyond the 40 PSF design specification. These drawings indicate that the roofs are capable of supporting 100 PSF. Subtracting 25 PSF to account for roof building materials the 75 PSF margin would support approximately 14.4 inches of water The IPEEE goes on to state the 96 PSF value corresponds to approximately 12 feet of fresh snow Given the approximate 3-foot roof wall height and normal winds off L ake Ontario it is judged unlikely that this level of snow could accumulate on roofs.
It is unclear to the staff whether the risk of this hazard is adequately considered for this application since snow loading for the 75 PSF value was not provided, and no discussion of how the threat is monitored or mitigated via internal guidelines or procedures is provided.
a) Given that heavy snowfalls can lead to loss of power events, provide procedures used to monitor and mitigate the winter snow load on critical buildings including the EDG buildings.
b) Justify the screening of risk associated with snowfall from the application by demonstrating that the past major snowfall events result in sufficient margin to withstand such events.
Technical Review Branch/Review Area APLA - Probabilistic Risk Assessment Licensing Branch A APLC - Probabilistic Risk Assessment Licensing Branch C