ML24144A192

Audit Plan for the Regulatory Audit of the GE Hitachi Topical Report NEDC-33934P/NEDO-33934, Revision 1, BWRX-300 Safety Strategy (Docket No. 99900003)

ML24144A192
Person / Time
Site:	99900003, 99902049
Issue date:	07/16/2024
From:	Carolyn Lauron NRC/NRR/DNRL/NLIB
To:	Hayes M NRC/NRR/DNRL/NLIB
Lauron, C., NRR/DNRL, 301-415-2736
Shared Package
ML24144A188	List: ML24144A192 ML24144A193
References
NEDC-33934P, Rev 1, NEDO-33934, Rev 1
Download: ML24144A192 (13)

Text

Enclosure UNITED STATES NUCLEAR REGULATORY COMMISSION PLAN FOR THE REGULATORY AUDIT OF GE HITACHI TOPICAL REPORT NEDC-33934P/NEDO-33934, REVISION 1, BWRX-300 SAFETY STRATEGY DOCKET NO. 99900003 APPLICANT:

GE-Hitachi Nuclear Energy APPLICANT CONTACT:

Bruce Bennett, Kelli Roberts Banks DATE:

July 22, 2024 - November 30, 2024 (approximate dates)

LOCATION:

U.S. Nuclear Regulatory Commission (USNRC or NRC)

Headquarters (via GEH Electronic Reading Room (eRR))

One White Flint North 11545 Rockville Pike Rockville, MD 20852-2736 AUDIT TEAM:

Steve Jones (NRC)

Santosh Bhatt (NRC)

Ryan Nolan (NRC)

Shanlai Lu (NRC)

Angelo Stubbs (NRC)

Mihaela Biro (NRC)

Stacey Rosenberg (NRC)

Elijah Dickson (NRC)

Yuken Wong (NRC)

Thomas Scarbrough (NRC)

Dinesh Taneja (NRC)

Joseph Ashcraft (NRC)

Calvin Cheung (NRC)

Vijay K. Goel (NRC)

Sheila Ray (NRC)

Sanja Simic (Canadian Nuclear Safety Commission (CNSC))

Douglass Miller (CNSC)

Hazem Mazhar (CNSC)

Suqiang Xu (CNSC)

Michael Xu (CNSC)

Noemie Duvivier (CNSC)

Dan Papaz (CNSC)

Mark Broeders (CNSC)

Yong Chang Liu (CNSC)

Gareth Hopkins (United Kingdom Office of Nuclear Regulation (ONR))

Alex Fife (ONR)

Additional audit team members may be added as needed.

2 PROJECT MANAGERS:

Carolyn Lauron (NRC)

Sean Gallagher (NRC)

BACKGROUND By letter dated March 8, 2024, GE-Hitachi (GEH) submitted to the U.S. Nuclear Regulatory Commission (NRC) a request to review its licensing topical report (LTR)

NEDC-33934P/NEDO-33934, Revision 0, BWRX-300 Safety Strategy.1 By letter dated June 21, 2024, GEH withdrew Revision 0 of the LTR and submitted Revision 1 for NRC staff review.2 GEH requested a review and determination of the acceptability of the LTR for use and the issuance of the NRC advanced safety evaluation with no open items by March 7, 2025.

GEH stated that the LTR is expected to be referenced in future licensing activities either by GEH in support of a Title 10 of the Code of Federal Regulations (10 CFR) Part 52 Design Certification Application or by a license applicant requesting a Construction Permit and Operating License under 10 CFR Part 50 or a Combined Operating License under 10 CFR Part 52 for U.S.

deployment of a BWRX-300. In addition, GEH noted that a similar correspondence has been transmitted to the Canadian Nuclear Safety Commission (CNSC).

Preapplication engagements on the GEH Safety Strategy White Paper and a readiness assessment of the LTR were completed prior to the LTR submission.3, 4 This audit will be conducted in accordance with NRC guidance.5 See Section II for more details on the scope of the audit.

Consistent with the Memorandum of Cooperation among the NRC, the CNSC, and the United Kingdom Office of Nuclear Regulation (ONR), CNSC and ONR staff will observe the conduct of this audit.6 1

Letter from M. Catts to NRC, NEDC-33934P/NEDO-33934, Revision 0, BWRX-300 Safety Strategy Licensing Topical Report, dated March 8, 2024, (Agencywide Document Access and Management System (ADAMS) Accession No. ML24068A207).

2 Letter from M. Catts to NRC, NEDC-33934P/NEDO-33934, Revision 1, BWRX-300 Safety Strategy Licensing Topical Report, dated June 21, 2024, (ML24173A200). Enclosure 1 to letter from M. Catts to NRC, Enclosure 1 - NEDC-33934P, Revision 1, "BWRX-300 Safety Strategy, Proprietary, dated June 30, 2024, (ML24173A201). The public version, NEDO-33934, (ML24173A202). The letter and both versions of the topical report are part of ADAMS package ML24173A199.

3 U.S. NRC, Preapplication Readiness Assessment Report of the GE-Hitachi Nuclear Energy Draft Licensing Topical Report NEDC-33934P, Revision 0, dated June 27, 2024, (ML24165A284).

4 U.S. NRC, February 28, 2024, Meeting Summary Meeting with GEH - BWRX-300 Safety Strategy Draft LTR Readiness Assessment Closeout Meeting, dated June 27, 2024 (ML24165A257).

5 Office of Nuclear Reactor Regulation Office Instruction, LIC-111, Regulatory Audits, dated October 21, 2019, (ML19226A274).

6 Memorandum of Cooperation on Advanced Reactor and Small Modular Reactor Technologies among the United States Nuclear Regulatory Commission, the United Kingdom Office of Nuclear Regulation, dated March 12, 2024, (ML24066A026).

3 I.

REGULTORY AUDIT OBJECTIVE The NRC staff will seek clarification, gain understanding, and verify information related to the subject LTR. The audit will identify whether information is needed to support a regulatory finding and subsequently submitted on the docket. In addition, the review and discussion of the audit material will help focus any requests for additional information.

II. REGULATORY AUDIT SCOPE The audit team will examine supporting documentation provided by GEH via the on-line portal, Certrec, and during virtual meetings. Audit topics include the initial information needs described below and any additional review needs identified by NRC staff in subsequent discussions during the audit.

III. INFORMATION AND OTHER MATERIAL NEEDED FOR THE REGULATORY AUDIT The NRC staff identified the following initial information needed for the regulatory audit by review area. Additional information may be identified in discussions during the audit. Italicized text specifies the NRC staff questions and/or requests.

Nuclear Methods and Systems

1. Provide an example classification list for the systems, structures, and components (SSCs) discussed in the LTR and comparisons of any differences with the currently operating boiling water reactor (BWR) fleet. Provide a justification on how the proposed SSC classification complies with regulations in 10 CFR Part 50. Discuss whether GEH intends for the SSC classification to be within the scope of the LTR.

2. Discuss the intent of the statement, GEH is seeking USNRC agreement... in Section 1.1 of the LTR. Specifically, clarify whether GEH requests an NRC finding.

3. Clarify the statement in Section 1.1 of the LTR which states:

As this LTR is not seeking approval of BWRX-300 design features, GEH is not seeking USNRC approval of compliance with the selected GDC; rather, USNRC approval of the applicability of the GDCs to the identified SSCs is requested.

4. The LTR states that regulatory compliance assessed within the LTR is focused on the Safety Strategy methodology and not the specific design features. Discuss specifically what GEH requests for approval in the LTR given the lack of design details which is typically needed to make such a finding.

5. There are several statements in the LTR that the NRC staff cannot review and/or approve without adequate design details. Examples are found in the proprietary marked text in Section 5.1 of the LTR and will be identified and discussed during the audit.

4

6. The LTR discussion of GDC 4 on page 91 states that:

Safety Class 1 SSCs are protected against dynamic effects including the effects of missiles, pipe whipping, jet impingement, and applicable hydrodynamic loads that may result from equipment failure, except in those cases where it is demonstrated that the probability of fluid system piping rupture is extremely low under conditions consistent with the piping design basis.

10 CFR Part 50 requires loss-of-coolant accident (LOCA) events to be treated as postulated accidents and analyzed regardless of frequency of occurrence and is separate and independent from consideration of dynamic effects in accordance with the requirements of GDC 4. Discuss how the Safety Strategy methodology meets the intent of 10 CFR Part 50 requirements to ensure all losses of coolant are classified/treated as postulated accidents and analyzed regardless of frequency of occurrence.

7. The LTR classifies the initiating events for Anticipated Operational Occurrences (AOOs),

design basis accidents (DBAs) and design extension conditions (DECs), which include additional failures after the postulated initiating event (PIE). The LTR states that GEH is seeking NRC agreement that event categorization is acceptable. Discuss whether the classification of the various initiating events listed in Chapter 15 of the NRC Standard Review Plan (SRP, NUREG-0800) will be part of the LTR or whether an applicant incorporating the LTR will perform this classification in a future licensing action.

Probabilistic Risk Assessment The NRC staff prefers that any examples used by GEH to address these audit questions are non-digital instrumentation and control SSCs.

8. Frequency of Occurrence A. Section 2.2.3, Foundational Defense-in-Depth Analysis Concepts, of the LTR states, Frequencies of occurrence delineate transitions between the event categories, apply to PIEs and to event sequences (which include additional failures after the PIE) and are identified as follows:

AOO - frequency greater than 1E-02 per reactor year DBA - frequency from 1E-02 to 1E-05 per reactor year DEC - frequency less than 1E-05 per reactor year Clarify whether the above frequency cut-off values refer to PIE or event sequence frequencies. Update the LTR, as necessary.

Provide a discussion and demonstration on how the PIE and event sequence frequencies are consistently derived and used to bin into event categories.

B. The LTR defines event sequences as follows: An event sequence consists of a PIE, assumed [defense line] DL mitigation functional failures (making the scenario more severe and less likely than the PIE alone), and the DL function successes that mitigate

5 the PIE. The same PIE typically appears in multiple event sequences based on application of different DL successes and failures in response to the PIE.

The above definition implies certain grouping of event sequences which may produce different event frequencies, depending on how they are parsed/grouped.

Provide a discussion on how event sequences are established and how the methodology avoids parsing frequencies, and consequently, impacting the binning into the event categories.

Provide examples of PIEs and event sequences in each event category.

C. Section 3.2.7.3.1, Event Sequence Frequency Uncertainty, of the LTR states, The BWRX-300 design conservatively accounts for uncertainty by using a 1E-05/year DBA-DEC frequency threshold versus a 1E-04/year threshold that has been endorsed by USNRC in Regulatory Guide (RG) 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors (Reference 7). This RG is not directly applicable to BWRX-300 design; see Section 5.3 for further discussion. This provides a layer of confidence that events with frequencies >1E-04/year are considered as DBAs.

The NRC staff is concerned with the use of the 1E-05/year cut-off value for the DBA-DEC frequency threshold without consideration of uncertainties, key assumptions, and cliff-edge effects. In this section of the LTR, justification of a layer of confidence is given with reference to an RG 1.233 threshold, which is not directly applicable to this LTR.

The NRC staffs concern applies to all regulatory matters where quantitative values generated from PRA models are used as thresholds for decision-making. Consistent with NRC safety goal policy, which is reflected in lower-tier regulatory guidance such as RG 1.200, justify how the use of quantitative values from PRA models are used for decision-making without consideration of uncertainties, key assumptions, and cliff-edge effects.

D. An AOO with a single failure can be considered an event sequence. If its frequency falls in the DEC range, is this considered a DEC or a DBA? If its considered a DEC, can it be mitigated with a DL4 function?

Provide pertinent examples.

9. Reliability Targets A. Section 4.4 describes the reliability targets for DL functions at a high level and provides the following targets by safety category:

Safety Category 1 DL functions: 1E-04 failures per demand.

Safety Category 2 DL functions: 1E-03 failures per demand.

6 Safety Category 3 DL functions: 1E-02 failures per demand.

In contrast, Section 3.1 appears to describe reliability targets at the DL level of of 1E-02 failures/demand for DL2, 1E-04 failures/demand for DL3, and 1E-03 failures/demand for DL4a.

Provide a discussion, with examples, of the reliability targets, their basis, how they are computed, and whether they are specified at the DL level or by safety category, or at the component, system, or function level.

10. DL 4 and Importance in the Overall Safety Strategy A. The Safety strategy report Section 2.2.5.4 describes DL 4. Table 2-1 defines DL4 as follows:

DL4a Detect and mitigate DECs, including event sequences associated with some DBA PIEs and failure of DL3 functions.

DL4b Detect and mitigate DECs to prevent core damage or mitigate the consequences of core damage events (severe accidents).

It is not clear to the staff the distinction between DL4a and DL4b.

Provide a further discussion, with examples, to clarify the distinction between DL4a and DL4b functions.

B. Section 2.2.5.4 states, DL4a functions, and equipment performing those functions, are subject to design requirements derived from the extended deterministic safety analysis (EX-DSA). It further states, DL4b functions, and equipment performing those functions, are subject to design requirements derived both from EX-DSA and severe accident analysis (SAA).

Provide discussion, with examples, how design requirements for DL4a and DL4b functions are derived from the EX-DSA and SAA.

C. Section 2.2.5.4 states, Functions in DL4a are independent and diverse from DL3 functions for any given event sequence in which both are credited.

Explain with examples how this independence and diversity is verified at the event sequence level.

D. Section 4.2.1, Defense Line Functions, discusses the importance of the different defense lines in terms of defense in depth. It states: DL4b functions are not as important as DL3 and DL4a functions, despite the high consequence of failure and DL4a functions are less important than DL3 functions.

These statements, which appears to assign relative importance to different DL functions, appear to be inconsistent with the defense-in-depth attributes in RG 1.174 in Section 2.1.1.3, Evaluation the Impact of the Proposed Licensing Basis Change on Defense-in-Depth, Consideration 1. Preserve a reasonable balance among the layers

7 of defense. The safety report further states make-ready support functions that support DL4a are assigned SC3 safety category, while make-ready support functions that support DL4b are assigned SCN safety category.

Further, Section 3.2.3.2 Complex Sequence Selection states: The scope of the complex sequence selection process corresponds to those Level 1 PSA event sequences that may lead to core damage, and which are more severe or involve more failures than deterministically selected event sequences. The sequences selected in this process are those used to establish performance requirements for DL4b functions which prevent escalation to severe accident conditions.

Explain why these complex sequences, which may have a frequency similar to those mitigated by DL4b functions are considered less important than DL4a functions.

Provide further discussion on whether and how the relative importance of various DLs is considered in SSC classification using the Safety Strategy methodology, especially SC-3 and SCN SSCs.

11. Complex Sequence/Severe Accident Sequence Selection A. Subsection 3.2.7.3.2, Complex Sequence/Severe Accident Sequence Selection Uncertainty, states, in part, While these activities do use PSA results, they employ a risk-informed defense-in-depth approach. Potential core damage sequences are examined and selected by consideration of whether adequate defense-in-depth exists to prevent core damage (complex sequences) or mitigate release (severe accident sequences).

Subsection 3.2.7.3.2 also states, Because there is no numerical metric for determining what sequences get evaluated as complex or severe accident sequences, the list of sequences to consider extends to those with very low frequencies to account for uncertainty, particularly when sequences rely on the high reliability of a single safety feature to keep risk low (e.g. a single passive system), consistent with the consideration of cliff edge effects Complex Sequence/Severe Accident Sequence Selection is important to the process of classification.

Provide a discussion and specific examples of how PRA is included in fault list and complex sequence selection to understand:

a. How these fault lists and complex sequences are selected for evaluation without numerical ranking and how that selection is risk-informed

b. How uncertainty and key assumptions are included in the process consistent with the consideration of cliff edge effects

c. How the process is repeatable and transparent B. The NRC staff will audit information related to the statement in the LTR regarding a numerical metric for determining complex sequence selection.

12. Thermal-hydraulic Uncertainties for Passive Systems A. Section 3.2.7.3, Treatment of Uncertainty in PSA Application to Safety Strategy, summarizes how PSA uncertainties are considered in the overall safety strategy with

8 regards to, among other items, Event Sequence Frequency and Complex Sequence/Severe Accident Sequence Selection. However, it does not mention T-H uncertainties in passive safety systems.

The staff notes that T-H uncertainties in passive designs are an important aspect to consider. Passive safety systems rely on natural driving forces which are small compared to those of pumped systems, and the uncertainty in their values can be of comparable magnitude to the predicted values themselves. Therefore, some event sequences that are not predicted to lead to core damage may lead to core damage when the PRA model considers the T-H uncertainties.

Discuss how thermal hydraulic parameter uncertainties are considered in the safety strategy (specifically, changes to the event sequences, and therefore, SSC classification, based on the uncertainties in T-H parameters that could affect the reliability of a passive system and introduce uncertainty in the determination of success criteria).

13. External Hazard Evaluation A. Based on Section 5.2.2 GDC 2, the staff understands SSCs that are important to safety, as described in GDC 2, include:

Equipment that performs or supports DL3 functions during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of a design basis external hazard event is protected and/or qualified, to remain functional during and after the design basis hazard event Equipment determined to be risk-significant based on external hazard-specific PSA modeling In this context, Section 3.2.2.1 briefly discusses screening of external hazards but does not refer to screening criteria or conservative analysis of external hazards from RG 1.200. RG 1.200, C.1.2.6 Technical Elements for a Screening and Conservative Analysis of Hazards for an At-Power Probabilistic Risk Assessment also states, It is recognized that for those new reactor designs with substantially lower risk profiles (e.g.,

internal events CDF below 1x10-6 per year), the quantitative screening value should be adjusted according to the relative baseline risk value.

Provide details on how the screening and conservative analyses are performed and is consistent with the intent of RG 1.200.

14. Coping Capability Analysis A. Section 3.2.3.3 states that event sequences identified in the complex sequence selection which include successful reactor shutdown and do not involve core damage are in the scope of the coping capability analysis.

In contrast, Section 3.2.3.2 states that the scope of the complex sequence selection process corresponds to those Level 1 PSA event sequences that may lead to core damage, and which are more severe or involve more failures that deterministically selected event sequences.

9 Explain how sequences that do not involve core damage are selected for the coping analysis. Provide example of scenarios analyzed for Coping Capability Analysis described in Section 3.2.6.4.

15. Practical Elimination A. Section 3.2.10 discusses practical elimination at a high level.

The staff notes the International Atomic Energy Agency (IAEA) recently published Specific Safety Guide SSG-88, Design Extension Conditions and the Concept of Practical Elimination in the Design of Nuclear Power Plants [January 2024] As stated in the document, SSG-88 provides IAEAs recommendations for the design of new nuclear power plants on the application of selected requirements in SSR-2/1 related to defense in depth and the practical elimination of plant event sequences that could lead to an early radioactive release or a large radioactive release.

Describe how scenarios are identified as being practically eliminated with examples of such scenarios (e.g., any related to reactor isolation valves).

The staff notes IAEA SSG-88 provides further guidance on practical elimination. Is it GEHs intent to follow this guidance?

16. Establishment of DL1 Requirements Based on the PSA A. Section 3.2.7, Probabilistic safety analysis states that DL1 design requirements are an output of the Level 1 and Level 2 PSA, and are specified based on assumptions made during the Level 1 PSA.

Provide a description with examples on how DL1 design requirements are established using the PSA.

17. Regulatory Treatment of Non-Safety Systems A. Section 1.1, Purpose, states Section 5.5 provides justification that the identification of Regulatory Treatment of Non-Safety Systems (RTNSS) SSCs is not necessary for the BWRX-300 design application, as SSCs are already assigned Safety Class 2 (SC2) or Safety Class 3 (SC3) as part of the BWRX-300 safety classification. GEH is seeking USNRC agreement with this assessment.

Section 5.5 of the report assesses RTNSS. In summary it argues that the RNTSS criteria are covered by the fact that all SSCs identified by the five RTNSS criteria are assigned to SC2 or SC3 safety classes.

As discussed in SECY-94-084, there are deterministic and probabilistic tests for SSCs that should be scoped into the RTNSS program. These tests include:

SSCs needed to meet 10 CFR 50.62 for ATWS Rule (Anticipated Transient without a Reactor Scram) and 10 CFR 50.63 for SBO (Station Blackout).

10 SSCs needed for long term safety (>72 hours) such as additional cooling water and to address seismic events.

SSCs needed for full power and shutdown conditions to meet the Commissions new reactor goals CDF < 1E-4 and LRF < 1E-6 per reactor year AND SSCs needed to maintain initiating event frequencies at the comprehensive baseline PRA levels.

SSCs needed to meet the containment performance goal including containment bypass during severe accidents.

o The conditional containment failure probability should be less than 1 in 10 when weighted over credible core-damage sequences. This goal will maintain a balance between accident prevention and consequence mitigation.

o The containment should maintain its role as a reliable, leak-tight barrier for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage under the more likely severe accident challenges and, following this period, the containment should.

continue to provide a barrier against the uncontrolled release of fission products.

SSC relied upon to prevent significant adverse systems interactions.

Explain how each specific RTNSS criterion is addressed and applied by the Safety Strategy methodology.

Explain how and where augmented design criteria and availability controls will be defined according to the specific functions of each non-safety related SSC that meets the intent of RTNSS via the Safety Strategy methodology consistent with SECY-94-084 and its associated SRM.

18. Technical Specification LCO Criterion 4 A. As stated in Section 1.1, Section 5.4 assesses BWRX-300 Safety Strategy compliance with Technical Specification (TS) Limiting Condition for Operation (LCO) Criteria in accordance with 10 CFR 50.36(c)(2)(ii). GEH is requesting approval of the conclusions made in this assessment.

Section 5.4 states, The BWRX-300 design complies with Criterion 4 by establishing LCOs for SSCs that are determined to be risk-significant using risk importance measures such as Risk Achievement Worth and Fussell-Vesely. Therefore, the BWRX-300 meets the requirements of 10 CFR 50.36(c)(2)(ii) Criterion 4.

The staff is unable to conclude that BWRX-300 meets the requirements of 10 CFR 50.36(c)(2)(ii) Criterion 4 without further information.

Provide further discussion on how SSCs significant to public health and safety will be identified based on the PSA under TS Criterion 4.

The Safety Strategy states, a review of other BWR TS content that are based on operating experience is required to be completed as part of the development of the final

11 TS by license applicants utilizing the BWRX-300 design when requesting an operating license under 10 CFR Part 50.

Because of the passive nature of BWRX-300 SSCs and the limited operating experience with such SSCs, justify why reviewing only BWR operating experience is sufficient to establish SSCs for which operating experience has shown to be significant to public health and safety.

19. Scope of the PSA A. Section 3.2.7.1 states The Level 1 PSA scope includes all plant operational modes (i.e.,

full power, low power, and shutdown) and considers events affecting both the reactor core and the spent fuel pool.

Section 2.2.3 states Design basis hazards, both internal and external, are not categorized according to the frequency values associated with the above event categories. The design basis frequencies are hazard-specific and reflect the regional location of the facility and regulatory expectations for categorization.

Are external and internal hazards intended to be included part of the PSA scope? Clarify the LTR to explain the intent of the cited in the sentence from Section 2.2.3 above.

20. Common Cause Failure A. Figure 3-2, DLs and Event Categories Per Analysis Type and Table 3-1, Application of DLs in Safety Analyses summarized the safety strategy. It shows that CCF on various DLs are being postulated. On page 28 the report states Functional failure analysis is limited to random single failures and to single CCFs. In addition, on page 16, CCF of digital I&C are noted as exception to the frequency-based event categorization.

Describe, with examples, how the CCF is being postulated and used in the safety strategy. Are the CCF postulated at the level of the system, function, or DL? Are CCF postulated for all types of components? Are the CCF probabilities an input into the safety strategy? What is it meant by single CCFs?

21. Single Failures, Independence, and Diversity of DLs A. Section 2.2.4, BWRX-300 Defense-in-Depth Concept states: Among the second, third, and fourth DLs, two independent and diverse lines can mitigate any PIE with a frequency greater than 1E-05 per reactor-year, for PIEs caused by single failures.

Provide examples of how independent and diverse DLs can mitigate these PIEs. What the statement seems to suggest is that a DL3 mitigative feature is backed up by a mitigative feature from a different DL. Is this an accurate statement?

12 Mechanical Engineering

22. The NRC staff will audit supporting information for the BWRX-300 Safety Strategy LTR with respect to the classification and qualification of mechanical equipment in comparison to NRC guidance, such as RG 1.26, Quality Group Classification and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants.

23. Section 4.5, Seismic Categorization states that structures and components that are required to be SC1 are categorized as seismic Category 1A or 1B, except that SC1 components that are in structures that are not seismic Category 1A and whose failure due to failure of the structure results in fail-safe performance of the components safety category function(s) are not required to be seismic Category 1A or 1B. The NRC staff notes that this seismic classification strategy is not consistent with the guidance in RG 1.29, Seismic Design Classification for Nuclear Power Plants. RG 1.29 does not take exception to allow seismic Category I SSCs that fail safe be classified as non-seismic Category I. The NRC staff also notes that this LTR requests approval the seismic classification requirements.

However, this LTR does not seek approval of the BWRX-300 design features. The NRC staff cannot review this exception to RG 1.29 and make a finding without design features of the SC1 components that fail safe. The method in RG 1.29 is one acceptable way for identifying and classifying those features of light-water reactor nuclear power plants that must be designed to withstand the effects of the safe-shutdown earthquake (SSE). The NRC staff would need more details on this different approach to identify and classify Seismic Category I SSCs, including details on the BWRX-300 design features.

Instrumentation & Controls

24. In Table 5-3, clarify what, if any, SC (components) need to comply with IEEE Std. 603-1991 and how/when would this be demonstrated.

25. In Section 4.2.6, Post-Accident Functions, clarify the power supply requirements for PAM B&C variables as stated in Section 6.6 of IEEE-497.

Electrical Engineering

26. Explain how the SSCs designated/identified as SC2, SC3 will be designed, manufactured, constructed, installed, commissioned, operated, tested, inspected, and maintained differently from those designated/identified as SC1 (safety-related).

IV. SPECIAL REQUESTS The NRC staff request that GEH provide subject matter expert(s), if necessary to discuss the details of the audit materials.

V. LOGISTICS AND DELIVERABLES Entrance Meeting:

July 22, 2024 (approximate)

13 Exit Meeting:

Approximately 8 weeks later The audit team will hold audit calls and/or meetings with GEH as necessary to understand the materials. The audit team will inform GEH of any emerging information needs.

The audit report will be issued within 90 days following the exit meeting. The NRC points of contact for this audit are Carolyn Lauron at Carolyn.Lauron@nrc.gov and Sean Gallagher at Sean.Gallgher@nrc.gov.