ML25234A193

Final Safety Evaluation for Electric Power Research Institute Report 3002028939, Risk-informed High-Energy Line Break Evaluation Requirements

ML25234A193
Person / Time
Site:	Electric Power Research Institute
Issue date:	08/29/2025
From:	Licensing Processes Branch
To:
References
EPRI TR 3002028939, EPID L-2024-LRM-0062 pre-app, EPID L-2024-TOP-0003 pre-fee, EPID L-2024-NTR-0006 post-fee, Risk-Informed HELB Methodology 3002028939
Download: ML25234A193 (20)

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Enclosure Safety Evaluation for a Topical Report Summary Information Topical Report No.:

Electric Power Research Institute Technical Report No. 3002028939 Topical Report

Title:

Risk-informed High-Energy Line Break Evaluation Requirements Sponsor:

Electric Power Research Institute (EPRI)

Summary of Request:

EPRI requested the U.S. Nuclear Regulatory Commission to review and approve EPRI Technical Report 3002028939, Risk-Informed High-Energy Line Break Evaluation Requirements, June 2024 Applicability:

All plants licensed under the Atomic Energy Act of 1954, as Amended, implementing Title 10 of the Code of Federal Regulation Part 50, Domestic licensing of production and utilization facilities, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 4, Environmental and dynamic effects design bases.

Submittal:

July 23, 2024, ADAMS Accession No. ML24205A146 (EPRI 2024a)

Supplement(s):

June 12, 2025, ADAMS Accession No. ML25164A225 (EPRI 2025a)

June 26, 2025, ADAMS Accession No. ML25178A798 (EPRI 2025b)

EPID No.:

EPID L-2024-NTR-0006 Principal Contributors to Safety Evaluation:

John Bozga, Mechanical Engineer, Mechanical Engineering & Inservice Testing Branch (EMIB), Division of Engineering and External Hazards (DEX), Office of Nuclear Reactor Regulation (NRR)

Stephen Cumblidge, Mechanical Engineer, Piping and Head Penetration Branch (NPHP), Division of New and Renewed Licenses (DNRL), NRR David Gennardo, Reliability and Risk Analyst, PRA Licensing Branch A (APLA), Division of Risk Assessment (DRA), NRR Kaihwa Hsu, Senior Mechanical Engineer, EMIB, DEX, NRR Ching Ng, Senior Reliability and Risk Analyst, APLA, DRA, NRR Eric Reichelt, Senior Materials Engineer, EMIB, DEX, NRR

ii Table of Contents Summary Information..................................................................................................................... i Table of Contents.......................................................................................................................... ii

1.

Introduction........................................................................................................................ 1 1.1 Description of Request................................................................................................... 1 1.2 Background.................................................................................................................... 1

2.

Regulatory basis................................................................................................................ 1

3.

Summary of Proposed Approach....................................................................................... 2

4.

Technical Evaluation.......................................................................................................... 3 4.1 Risk-Informed HELB Methodology................................................................................. 4 4.1.1.

Consequence of Failure Evaluation...................................................................... 4 4.1.2.

Failure Potential Evaluation................................................................................... 8 4.1.3.

Risk Characterization............................................................................................ 9 4.1.4.

HELB Response Strategies................................................................................... 9 4.1.5.

Risk Impact.......................................................................................................... 10 4.1.6.

Performance Monitoring...................................................................................... 10 4.2 Risk-informed Evaluation.............................................................................................. 11 4.2.1.

Key Principle 1: Licensing Bases Change Meets the Current Regulations......... 11 4.2.2.

Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth (DID) Philosophy.................................................................................................................. 11 4.2.3.

Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins... 12 4.2.4.

Key Principle 4: Change in Risk is Consistent with the Safety Goals.................. 12 4.2.5.

Key Principle 5: Monitor the Impact of the Proposed Change............................. 13 4.3 Environmental Qualification (EQ) of Electrical Equipment Important to Safety............ 13

5.

Limitations and Conditions............................................................................................... 15

6.

Conclusion....................................................................................................................... 15

7.

References....................................................................................................................... 16

8.

Abbreviations................................................................................................................... 18

1.

INTRODUCTION 1.1 Description of Request By letter dated July 23, 2024 (EPRI 2024a), as supplemented on June 12, 2025 (EPRI 2025a) and June 26, 2025 (EPRI 2025b), Electric Power Research Institute (EPRI) submitted EPRI Technical Report (TR) 3002028939, Risk-Informed [RI] High-Energy Line Break [HELB]

Evaluation Requirements (also referred to as TR RI-HELB), to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. The June 26, 2025, submittal provided an amended draft of the topical report. EPRI TR RI-HELB provides an alternative means for assessing and confirming that plant structures, systems, and components (SSCs) that are important to safety are adequately protected against the postulated dynamic and environmental effects of postulated pipe ruptures.

By email dated September 10, 2024 (NRC 2024a), the NRC staff accepted EPRI TR RI-HELB for review.

1.2 Background

The NRC staff issued one round of requests for additional information (RAIs) containing 17 requests for information and 1 request for confirmation:

May 19, 2025, ADAMS Accession No. ML25129A086 (NRC 2025a)

The NRC staff performed an audit to support its review:

NRC audit plan dated October 25, 2024, ADAMS Accession No. ML24298A056 (NRC 2024b)

NRC audit report dated August 22, 2025, ADAMS Accession No. ML25227A108 (NRC 2025b)

2.

REGULATORY BASIS Title 10 of the Code of Federal Regulation (10 CFR) Part 50, Domestic licensing of production and utilization facilities 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants General Design Criterion (GDC) 4, Environmental and dynamic effects design bases -

Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents [(LOCA)]. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

2 While not explicitly part of the review of this TR, the staff also considered the impact of its application to:

10 CFR Section 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, (the EQ rule) requires, in part, licensees to establish a program for qualifying the electric equipment important to safety. The electrical equipment under the scope of this section includes safety-related equipment, non-safety-related electrical equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified by the safety-related equipment, and certain post-accident monitoring equipment. The equipment should remain functional during and following design-basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures.

Each licensee has an approach which has been reviewed by the NRC staff for addressing the requirements of GDC 4 and 10 CFR 50.49. The criteria to meet the relevant requirements are contained in Standard Review Plan (SRP) (NRC 2007a) Section 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment,"

Revision 3 (NRC 2007b), SRP Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Revision 3 (NRC 2016a),

Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment, Revision 3 (NRC 2007a), Branch Technical Position (BTP) 3-3, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, Revision 3 (NRC 2007c), and BTP 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Revision 3 (NRC 2016b), or plant-specific licensing basis.

These NRC guidance documents provide criteria for the protection against high-energy line breaks, including defining high-energy versus moderate-energy systems, defining the applicable high-energy and moderate-energy systems and defining which individual locations within systems where postulated breaks would occur and those locations where postulated breaks would not occur. The criteria describe methods for analyzing pipe whip and jet impingement forces, design of pipe whip restraints and jet impingement shields, methods for evaluating the structural integrity, and methods for evaluating environmental effects.

3.

SUMMARY

OF PROPOSED APPROACH The RI-HELB methodology is based on EPRI TR-112657 (EPRI 2000a), for Risk-Informed Inservice Inspection (RI-ISI), and the extension of RI-ISI to break exclusion region (BER) programs in EPRI TR-1006937 (EPRI 2002a), and is used to determine SSCs impacted by the postulated pipe breaks and the risk significance of these breaks. This proposed RI methodology would provide an alternative to the above deterministic approaches for meeting GDC 4 requirements. Each licensee seeking to implement the alternative would need to follow the appropriate change control process.

This proposed alternative approach is for Class 2, 3, and non-safety-related piping systems with the exception of the portion of piping that is near the containment penetration areas (referred to as the BER). EPRI stated that the scope of this TR is meeting GDC 4, and that initial application is expected to be for currently operating plants that undergo a plant modification. EPRI also

3 indicates that this TR can be used for other applications, and that the RI-HELB evaluation should reflect the intended application.

The RI-HELB process includes the following steps:

(1) Definition of RI-HELB program scope (2) Failure modes and effects analysis (FMEA) of HELB scope

a. Evaluation of pipe failure potential

b. Evaluation of consequences of pipe failures (3) Characterization of risk segments (risk matrix)

(4) HELB response strategies (5) Evaluation of risk impact of changes to the HELB program (6) Incorporation of long-term RI-HELB program (performance monitoring)

EPRI states that the use of this methodology can involve changes to the zones of influence (ZOIs) that are used to evaluate target equipments ability to perform its function following a postulated pipe break. EPRI indicates that such changes should be reflected in the plants EQ program and the plants probabilistic risk assessment (PRA). These changes are outside the scope of the RI-HELB TR.

4.

TECHNICAL EVALUATION The NRC review and approval of EPRI TR RI-HELB, as supplemented, is limited to the RI-HELB methodology contained in EPRI TR RI-HELB, Chapter 2. Any clarifications or limitations based on text in other chapters or RAI responses needed to support a regulatory finding are specifically identified in this safety evaluation (SE). Chapter 1 mostly describes background and history and do not directly relate to the proposed R-IHELB methodology, other than the discussion of applicability in Chapter 1. Chapter 3 is an example application of the RI-HELB methodology described in Chapter 2 and provides additional guidance which may be useful during implementation of EPRI TR RI-HELB. Chapter 4 contains EPRIs assessment of the proposed methodology against the five key principles of risk-informed decisionmaking from Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 (NRC 2018a). Chapter 5 is a summary and Chapter 6 includes the document references. In addition, EPRI provided commentary and background in the supplements and RAI responses.

This evaluation highlights the portions of the TR and RAI responses that underpin the NRC conclusions herein. Other portions of the TR, RAI responses, and supplements which do not affect the text in Chapter 2 should not be considered approved or unapproved based on this review.

4 4.1 Risk-Informed HELB Methodology 4.1.1.

Consequence of Failure Evaluation Fundamental Principles The initial part of the FMEA is focused on pipe segment failures in terms of their impact on core damage and large early release. Pipe segments are defined as lengths of pipe that are exposed to the same degradation mechanism (DM) and whose failure leads to the same consequence.

That is, some lengths of pipe whose failure would lead to the same consequences are split into two or more segments when two or more regions are exposed to different DMs. Similarly, lengths of pipe exposed to the same DM whose failure would lead to different consequences are split into two or more segments.

This alternative methodology uses an approach for piping segments consistent with Sections 3.3 and 3.5 of EPRI TR-112657 (EPRI 2000a).

The consequence evaluation focuses on the direct and indirect impact of a pipe segment failure.

Direct effects include the loss of a train or system and associated possible diversion of flow or an initiating event, such as a LOCA, or both. Indirect effects include high temperatures, dynamic effects arising from pipe whip or jet impingement and other spatial effects, such as from floods and spray, that may affect adjacent SSCs. Spatial consequences of the break are determined based on the location of the analyzed break and the position of SSCs. A walkdown can be performed to validate the location of the SSC with respect to the postulated break.

The consequence evaluation also focuses on the automatic isolation of postulated pipe breaks and operator actions to isolate pipe breaks consistent with Section 3.3.1 of EPRI TR-112657 (EPRI 2000a). The consequence assessment addresses both successful and unsuccessful isolation. Operator recovery actions are consistent with Section 3.3.3.2 of EPRI TR-112657 (EPRI 2000a).

EPRI expanded on the EPRI TR-112657 (EPRI 2000a) automatic isolation and operator actions approach to ensure large and small HELB breaks have been addressed and in its response to RAI 4 EPRI stated:

For large breaks (e.g., double-ended breaks) that are usually limiting with regard to consequence, only automatic isolation should be credited as there is not enough time for operator response (e.g., blowdown has already occurred and causes consequences).

For small breaks that are not large enough to generate existing automatic isolation signals, operator actions can only be credited if the following are met:

There is an alarm and/or clear indication to which the operator will respond The response is directed by procedure The isolation equipment (e.g., valves) is not affected by the break There is enough time to perform isolation and reduce consequences

5 It is possible that smaller piping could become HELB lines where a double-ended break is small with minor consequences and possibly there is no automatic signal. If operator actions are credited, the above requirements must also be applied.

A spectrum of break sizes is considered for a specific run of piping being evaluated to ensure the most limiting consequence is considered. The break size can range from a small leak to a rupture.

Since this guidance is consistent with guidance previously approved in EPRI TR-112657, the NRC staff finds this guidance acceptable.

Consequence Ranking and Categorization Section 2.1.2 of EPRI TR RI-HELB stated that the High consequence category corresponded to a value of conditional core damage probability (CCDP) >10-4 or a value of conditional larger early release probability (CLERP) >10-5. The NRC staff noted that assuming a pipe break frequency of 10-2 per year, the corresponding increase in core damage frequency (CDF) and large early release frequency (LERF) would be >10-6 per year and >10-7 per year, respectively.

The staff noted that in RG 1.174, any application that results in an increase of CDF (or LERF) at this magnitude would at least be considered a small change in risk, as the change would be above the risk acceptance guidelines for a very small change in risk. In this case, the RI-HELB methodology would conservatively require the applicant to follow existing deterministic HELB requirements or would require plant modification to reduce the consequences such that the postulated HELB event would no longer be considered high consequence. The NRC staff notes that the need for and the execution of plant modifications are outside the scope of this review and would be governed by the appropriate change control processes.

For the Low consequence category, CCDP 10-6 and CLERP 10-7 are selected as the threshold. The NRC staff noted that that even with a 10-2 per year pipe break frequency, the resulting increase in risk using these CCDP and CLERP metrics will not be higher than 10-8 per year and 10-9 per year for CDF and LERF, respectively. The NRC staff noted that such changes in risk would be an order of magnitude below the risk acceptance guidelines for a very small change in risk per RG 1.174.

The NRC staff also noted that the consequence measures of CCDP and CLERP are independent of pipe failure likelihood, and they can be used in combination with the qualitative pipe rupture potential in the risk matrix (Figure 2-6 of EPRI TR RI-HELB) to produce meaningful risk measures.

In view of the above, the NRC staff finds that these consequence categorization guidelines are consistent with the intent of RG 1.174 and, when used in the methodology as described in the topical report, provide reasonable assurance that risk increases (if any) resulting from a proposed change would be acceptably small.

6 Consequence Evaluation For consequence evaluation, the consequences of pipe rupture are measured in terms of the CCDP, given a pipe rupture, and the CLERP, given a pipe rupture. These measurements require quantitative risk estimates obtained from the plant-specific probabilistic risk assessment (PRA) models that are available for each plant. As discussed previously, the consequence evaluation focuses on the direct and indirect impact of a pipe segment failure.

In response to RAI 3, Section 2.1.3 of EPRI TR RI-HELB was revised to include the eight consequence evaluation criteria from EPRI TR-1006937 as part of the HELB impact assessment. One of the criteria discusses a typical zone of influence of 10 to 20 pipe diameters based on NUREG/CR-2913, Two Phase-Jet Loads, 1983 (NRC 1983). ZOI is typically defined as the distance of influence for jet spray from breaks in high-energy line piping to the targeted equipment. Each operating plant has established a zone of influence in their HELB licensing basis. The NRC staff notes that licensees have submitted license amendment requests (LARs) to change their zone of influence HELB licensing basis to use NUREG/CR-2913. The staff is not evaluating the need for or approving any changes to the ZOI methodologies as part of this TR.

PRA Acceptability Section 2 of EPRI TR RI-HELB states as with any risk-informed application, the RI-HELB methodology is an integrated decisionmaking process which requires the input and use of multiple disciplines including personnel with expertise in PRA, plant operation, system design, safety/accident analyses, and DM evaluations. EPRI TR RI-HELB also states that the RI-HELB methodology requires the use of a robust plant-specific PRA which includes an assessment of key assumptions and sources of uncertainties. The topical report states that a plant-specific PRA that reflects the as-built/as-operated plant, and that has been peered reviewed and shown to meet Capability Category II of the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard as endorsed in RG 1.200, would meet this requirement. EPRI TR RI-HELB does not specify a revision of RG 1.200, but the NRC staff notes that both RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3 (NRC 2020a) and RG 1.200, Revision 2 (NRC 2009), are considered acceptable means to demonstrate PRA acceptability (also known as PRA technical adequacy) for an internal events and internal flood PRA. The NRC staff also notes that it is acceptable to use gap assessments of earlier peer reviews of the internal events and internal flood PRA against later versions of RG 1.200, in order to demonstrate PRA acceptability.

The NRC staff notes that not all equipment, or associated failure modes due to HELB, may be modeled in the plant-specific PRA. In RAI 9, the NRC staff requested that the applicant clarify how the licensee would evaluate the consequences if the target equipment (i.e., equipment damaged due to the direct or indirect effects of a pipe break) is not modeled in the PRA or if the target equipment is not currently subjected to a HELB but the piping evaluated becomes high energy as a result of the proposed change. In its response to RAI 9, the EPRI explained that piping and equipment that was previously not subject to a HELB but later becomes a target of a HELB must be evaluated as part of the RI-HELB evaluation. This would include updating the HELB scenarios and/or developing new HELB scenarios for piping that becomes high energy.

Furthermore, Section 2, Step 2 of the topical report was revised to state that the RI-HELB evaluation should reflect the intended application and new targets may be identified and adversely impacted. EPRI TR RI-HELB then states, These new targets and the impact of their failure due to the potential new HELB need to be reflected in the RI-HELB evaluation.

7 Furthermore, the NRC staff also notes that in accordance with the ASME/ANS PRA Standard, as endorsed in RG 1.200, a PRA configuration control program should be in place to ensure the PRA is updated and reflects the asbuilt, asoperated plant. Therefore, the direct and indirect effects due to the application of the RI-HELB would be reflected in the PRA consistent with the licensees PRA configuration control program. However, as discussed in the response to RAI 9, PRA update requirements are outside the scope of EPRI TR RI-HELB.

The NRC staff notes that the scope of the PRA model may include both safety-related (SR) and non-safety-related (NSR) equipment. In its response to RAI 8, the applicant stated that RI-HELB evaluation must consider all equipment modeled in the PRA (both SR and NSR). Furthermore, Section 2.1.4 of EPRI TR RI-HELB was updated to state that the impact of a pipe failure and resulting interactions with other components (that is SR, and non-SR) are assessed as part of the consequence evaluation.

Section 2.1.1 of EPRI TR RI HELB discusses the identification of important equipment that could be impacted by the spatial (indirect) effects of a HELB. The NRC staff notes that in human reliability analysis (HRA), available time is an important factor when evaluating the human error probability and available time may be reduced if a HELB were to occur. A HELB can also result in a harsh environment such that local operator actions (e.g., isolation) are no longer feasible.

The NRC staff notes that the topical report does not explicitly address potential impacts to HRA and the associated human error probabilities for operator actions. Therefore, in RAI 12, the NRC staff requested clarification about how adverse impacts to operator actions would be identified and assessed. In its response to RAI 12, EPRI stated that it will update Section 2.1.1 of EPRI TR RI-HELB to highlight the HELB impact on possible local operator actions.

Specifically, local operator actions credited in a specific location should be identified through the analyses and the effect on the actions should be confirmed by a walkdown.

Based on the requirement to use a robust plant-specific PRA, the description of an acceptable PRA, and the clarifications provided with respect to how the PRA would be used to support the RI-HELB evaluations, the NRC finds the approach for PRA Acceptability outlined in the topical report is consistent with the guidance in RG 1.200 and RG 1.174.

Plant Walkdown The approach for plant walkdowns to support the RI-HELB methodology is consistent with Section 5.4 of EPRI TR-112657. The walkdowns are an integral part of the consequence evaluation. The objective of the walkdown is to identify indirect effect impacts which include dynamic effects arising from pipe whip or jet impingement and other spatial effects, such as from floods and spray, that may affect adjacent SSCs. In addition, the walkdown is used to identify any mitigation devices such as pipe whip restraints and jet impingement shields which may provide protection as well as any flood/spray protection features in the plant. Lastly, the walkdown can identify SSCs impacted by environmental effects. Operator actions which are credited should be confirmed by a plant walkdown. The information documented from the walkdown is evaluated in the consequence assessment.

Since this guidance is consistent with guidance previously approved in EPRI TR-112657 by the NRC staff to perform walkdowns as part of a consequence evaluation and assessment, the staff NRC finds this guidance acceptable.

8 4.1.2.

Failure Potential Evaluation Degradation Mechanisms (DMs) Evaluation Section 2.2.1 of EPRI TR RI-HELB describes a method to evaluate the subject piping to determine whether it is susceptible to select DMs that have been observed in high-energy lines.

These DMs include several types of fatigue cracking, several types of corrosion, stress corrosion cracking, and flow-induced corrosion. This section was updated and revised based on RAI 13.

Guidance is given in Section 2.2.1 and in Table 2-4 of EPRI TR RI-HELB as to how to evaluate whether the subject piping is vulnerable to the different DMs. This guidance and criteria for establishing whether a DM is applicable to piping is based on operational experience with the different DMs. Some DMs in Table 2-4 are given proscriptive criteria for determining if the DM applies, and others DM evaluations refer to other EPRI topical reports and owners programs.

The methods and guidance described in Section 2.2.1 and Table 2-4 are derived from EPRI TR-T112657, Revision B-A, and are consistent with previous risk-informed evaluations for RI-ISI, including for the BER. The industry and the NRC have significant experience with the methods described in EPRI TR-112657, Revision B-A, providing reasonable assurance that piping being evaluated using Section 2.2.1 and Table 2-4 would be assigned with the appropriate DMs.

Plant-Specific Service History Review In addition to the general review of DMs, EPRI TR RI-HELB requires a site-specific review of DMs so that the DM categorizations are in line with the individual history of the plant. This step is important as the different types and designs of reactors can have very different experience with DMs, based on the variety of materials and operating conditions.

EPRI expanded upon the EPRI TR-112657 DMs to ensure current operating experience has also been included and in its response to RAI 7a, RAI 7b and RAI 7c, EPRI stated: In addition, dependent upon the scope of the RI-HELB application and the affected piping systems or components, consideration should be given to any additional DMs contained in ASME Section III Nonmandatory Appendix W, as appropriate.

The NRC staff find that the use of the DM evaluation method combined with the plant-specific review provide confidence that the pipe segments would be categorized correctly.

DM Categories EPRI TR RI-HELB categorizes pipes into three Pipe Rupture Potential categories, High, Medium, and Low. The categorization has a preliminary categorization using Table 2-5, which assigns piping subject to flow-accelerated corrosion (FAC) as High, piping with no assigned DM as Low, and all piping with assigned DMs other than FAC as Medium. This is the same method used in EPRI TR-112657, Revision B-A, and is a commonly used method of assigning Pipe Rupture Potentials.

Pipes must be verified to show that they are not subject to water hammer, or they will be moved to a high pipe rupture potential categorization, as discussed in Sections 2.2.2 and 2.2.3 of EPRI

9 TR RI-HELB. Other damage mechanisms may also be reviewed as described in Section 2.2.2, dependent upon the scope of the application and the subject piping systems or components.

The other possible mechanisms described in Appendix W that could result in a high pipe rupture potential categorization include Thermal Aging Embrittlement, Vibration Loads, and unstable Fluid Flow.

The Pipe Rupture Potential categorizations are derived from EPRI TR-112657, Revision B-A, and are consistent with previous risk-informed evaluations for RI-ISI, including for the BER. This method would provide reasonable assurance that Pipe Rupture Potential categorizations made using this method would be appropriate.

4.1.3.

Risk Characterization In EPRI TR RI-HELB, the risk of pipe segment failure is characterized on the basis of the expected likelihood of the event and the expected importance of the consequence. The importance of the consequences is presented by the consequence categories. The likelihood of failure is represented by the DM (or pipe rupture potential) categories. The graphic method is used to illustrate the effects of these two parameters and to serve as a base for the selection of risk important segments. The graphic structure, which is known as the risk matrix, is shown in Figure 2-6 of EPRI TR RI-HELB, and is used to define risk categories (RCs). The NRC staff noted that the risk matrix is the same method used in EPRI TR-112657, Revision B-A, and is a commonly used method to qualitatively evaluate the risk ranking.

4.1.4.

HELB Response Strategies In Section 2.4 of EPRI TR RI-HELB, the HELB response strategies are based on the risk matrix in Figure 2-6 of EPRI TR RI-HELB. The discussion of HELB response strategies assumes that the FAC is the only applicable DM with high pipe rupture potential. (As discussed previously, in order to apply the RI-HELB methodology, the potential for water hammer needs to be verified to be low such that it is not applicable.)

The risk is considered High for RCs 1, 2, and 3, Medium for RCs 4 and 5, and Low for RCs 6 and 7. (The consequences are considered High for RCs 1, 2, and 4.) Specifically, for RCs 1 and 3, the risk is High, in part, due to the high failure potential associated with FAC. The response strategies require that, for the locations identified as susceptible to FAC, the licensees FAC program ensures that the program is addressing the most important locations.

The licensee is then allowed to downgrade the pipe rupture potential such that FAC is no longer considered to be a DM. (This effectively means RC1 and RC3 would no longer be applicable unless there was some other DM identified as a high rupture potential.) For example, if a pipe segment was originally RC1, it would then be characterized as either RC2, if there are other DMs present, or RC4 if FAC was the only applicable DM. All the licensees with piping subject to FAC have a FAC program in place and the FAC program has been shown to be effective to manage FAC. Therefore, the NRC staff finds it acceptable to credit the FAC program for managing FAC such that the DM does not need to be accounted for when defining the RCs.

Similarly, the HELB response strategies note that plant modifications can be performed to reduce the consequence category associated with a HELB event. (As discussed previously, plant modifications are outside the scope of this review.) For example, if a pipe segment was RC2, and the consequences of its failure were reduced from High to Medium, it would become RC5, or if its consequences were reduced to Low, it would become RC6.

10 Pipe segments whose final risk characterization is RC2 or RC4 are required to follow existing deterministic HELB requirements. The NRC staff notes that, based on this logic, if a pipe segment were to remain RC1 it would also follow the existing deterministic HELB requirements.

Pipe segments whose final risk characterization is RC5 are allowed to credit 10 percent inspections based on DM, as an alternative to meeting the deterministic HELB requirements.

The NRC staff notes that a 10 percent inspection would be similar to the requirements in RI-ISI and RI-BER and acceptable because 10 percent sampling will provide reasonable assurance that degradation would be found if a significant number of pipes were degrading. The NRC staff also notes that, based on this logic, if a pipe segment were to remain RC3 it would also be allowed to credit the 10 percent inspections as an alternative. Finally, pipe segments whose final risk categorization is RC6 or RC7 do not need to meet the deterministic HELB requirements and have no alternative requirements applied. Given the Medium risk (and Medium consequence ranking) associated with pipe segments that are RC5 (or RC3) and the Low risk (and Medium or Low consequences) associated with RC6 or RC7 pipe segments, the NRC staff finds this approach for final risk characterization and the associated HELB response strategies to be acceptable.

4.1.5.

Risk Impact In Section 2.5 of EPRI TR RI-HELB, the risk impact of RI-HELB is estimated by taking the CCDP and CLERP values from the analysis and multiplying by the estimated frequency of failure of 10-2 /year. The results are then compared to the guidelines in RG 1.174. The NRC staff evaluation of the risk impact is documented in Section 4.1.1, Subsection entitled Consequence ranking and categorization, of this SE.

4.1.6.

Performance Monitoring RG 1.174 states that the licensee should propose monitoring programs that adequately track the performance of equipment. Section 4 of EPRI TR RI-HELB states, in part, there are no unique aspects of the RI-HELB methodology insofar as monitoring requirements are concerned. In its response to RAI 11, EPRI stated that monitoring programs put in place as a result of the application of the RI-HELB methodology shall be consistent with Section 3 of RG 1.174, Revision 3, to ensure that the RI-HELB evaluation conclusions remain valid.

Section 2.6, Performance Monitoring, was updated accordingly, stating that, when developing the performance monitoring programs, options such as the FAC program and ISI program should be considered. The NRC staff also noted that inspection and inspection population are considered as options for various RCs. The NRC staffs evaluation of the adequacy of the inspection population based on the DM is documented in Section 4.1.4 of this SE.

In its response to RAI 15b, EPRI also updated Section 2.6 of EPRI TR RI-HELB to include: In addition, consistent with RG 1.174, risk-informed decisionmaking process, the licensee is required to review changes to the plant, operational practices, applicable plant and industry operational experience over the entire life of the plant including environmental conditions, and, as appropriate, update the PRA and the RI-HELB evaluations. As discussed in Section 4.1.1 of this SE, updating the PRA is necessary to ensure that it reflects the as-built, as-operated.

However, as discussed in the response to RAI 9, the requirements for how to update the PRA are outside the scope of EPRI TR RI-HELB. Similarly, updating the RI-HELB evaluations, as necessary, ensures that the conclusions of the evaluations remain valid or would allow for new response strategies to be implemented, as needed. As discussed in Section 4 of EPRI TR RI-HELB, the licensee is expected to incorporate the RI-HELB methodology into plant-specific

11 program procedures, including requirements to maintain the associated evaluations and the PRA.

Considering the above, the NRC staff has determined that proposed monitoring programs would satisfy the fifth key principle for risk-informed decisionmaking identified in RG 1.174 and that the RI-HELB method would continue to be acceptable for use with the performance monitoring strategies.

4.2 Risk-informed Evaluation The following sections summarize how the five key principles of risk-informed decisionmaking in RG 1.174, Revision 3, are met, in part, by the implementation of the RI-HELB methodology.

4.2.1.

Key Principle 1: Licensing Bases Change Meets the Current Regulations Key Principle 1 states that the proposed change must meet current regulations unless it is explicitly related to a requested exemption or rule change. The proposed RI-HELB change is an alternative to satisfy the GDC 4 requirement as may be requested under 10 CFR 50.90.

The proposed change is an alternative to the existing guidance on meeting HELB requirements (SRP Sections 3.6.1 and 3.6.2, and BTPs 3-3 and 3-4) based on the risk characterization process and guidance contained in EPRI TR-112657 and EPRI TR-1006937. Each licensee seeking to implement the alternative would need to follow the appropriate change control process.

Considering the above, the NRC staff has determined that the proposed change satisfies the first key principle for risk-informed decisionmaking identified in RG 1.174 and that the licensing basis change will meet current regulations.

4.2.2.

Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth (DID) Philosophy RG 1.174, Revision 3, states that the proposed licensing basis change should be consistent with the DID philosophy. The intent of this key principle of risk-informed decisionmaking is to ensure that any impact of the proposed licensing basis change on DID is fully understood and addressed and that consistency with the DID philosophy is maintained. Section 2.1.1.2 of RG 1.174, Revision 3, identifies seven considerations to evaluate how the proposed licensing basis change impacts DID. The NRC staff noted that Section 4 of EPRI TR RI-HELB provided a discussion of the five key principles that risk-informed licensing basis changes are expected to meet. In its response to RAI 10, the applicant explained that the application of the RI-HELB methodology is consistent with the DID philosophy as outlined in Section 2.1.1.2 of RG 1.174, Revision 3, and revised EPRI TR RI-HELB to include justification for each of the considerations.

Specifically, for preserving reasonable balance among the layers of defense, the applicant stated that the RI-HELB methodology will not increase the likelihood of an HELB or create new significant HELBs. Any HELB events whose final consequence rank is High (i.e., CCDP >10-4 or CLERP >10-5) are required to meet existing deterministic HELB requirements. Additionally, the response explained there will not be a significant impact to the containment function or SSCs supporting that function such as containment fan coolers and sprays. The response also stated the RI-HELB methodology would not reduce the effectiveness of the emergency preparedness

12 program. Therefore, the plant would be expected to maintain a reasonable balance among the layers of defense.

EPRI also stated that the RI-HELB does not substitute programmatic activities for design features to an extent that reduces the reliability and availability of the design features.

Therefore, the approach avoids over-reliance on programmatic activities. The NRC staff notes that any HELB events with a final High consequence rank would continue to meet existing deterministic HELB requirements so that there would not be a significant change in DID-with respect to preserving redundancy, independence, and diversity; adequate defense against potential common cause failure and maintaining multiple fission product barriers.

EPRI explained that the RI-HELB methodology relies on the PRA that reflects the as-built/as-operated plant. Human actions credited in the RI-HELB evaluation, if any, would need to address human reliability considerations consistent with the ASME/ANS PRA Standard (ASME 2009a). Therefore, preserving sufficient defense against human errors. The applicant stated that the RI-HELB methodology is an alternative means for assessing and confirming that plant SSCs that are important to safety are adequate to accommodate the effects of postulated accidents, including appropriate protection against the dynamic and environmental effects of postulated pipe ruptures. Therefore, the plant would continue to meet the intent of the plants design criteria.

Based on the above, the NRC staff concludes that the RI-HELB methodology is consistent with the DID philosophy.

4.2.3.

Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins Key Principle 3 requires that the impact of the proposed licensing basis change is consistent with the principle that sufficient safety margins are maintained. Safety margins used in the piping design calculations are not changed. The High and Medium risk piping is monitored by inservice examination to ensure that DMs do not reduce the pressure boundary integrity to unacceptable levels. No changes to the evaluation of design-basis accidents in the final safety analysis report are being made by the RI-HELB process.

Considering the above, the NRC staff has determined that the proposed change satisfies the third key principle for risk-informed decisionmaking identified in RG 1.174 and that sufficient safety margins will be maintained.

4.2.4.

Key Principle 4: Change in Risk is Consistent with the Safety Goals The proposed RI-HELB methodology uses results and insights from a plant-specific PRA. The PRA must be acceptable and must be subjected to a peer review process. As discussed in Section 4.1.1, specifically Subsection PRA Acceptability, of this SE, using a PRA that meets Capability Category II of the ASME/ANS PRA Standard as endorsed in RG 1.200 would meet this requirement. Key Principle 4 states that when proposed changes result in an increase in CDF or LERF, the increase should be small. The NRC staff findings regarding Key Principle 4 are found in Section 4.1.1, Consequence ranking and categorization, of this SE. The NRC staff found that the consequence evaluation is consistent with guidance previously approved in EPRI TR-112657 and that the consequence categorization guidelines provide reasonable assurance that risk increases would be acceptably small.

13 Considering the above, the NRC staff has determined that the proposed change satisfies the fourth key principle for risk-informed decisionmaking identified in RG 1.174.

4.2.5.

Key Principle 5: Monitor the Impact of the Proposed Change Key Principle 5 requires that the impact of the proposed licensing basis change should be monitored using performance measurement strategies. RG 1.174, Section 3, explains that the primary goal of performance monitoring is to ensure that no unexpected adverse safety degradation occurs because of the change(s) to the licensing basis. Each licensee should consider any new operating experience based on their plant-specific service history review. The NRC staff findings regarding Key Principle 5 are found in Section 4.1.6 of this SE.

Considering the above, the NRC staff has determined that the proposed change satisfies the fifth key principle for risk-informed decisionmaking identified in RG 1.174 and that the methodology would continue to be acceptable for use with the performance monitoring strategies.

4.3 Environmental Qualification (EQ) of Electrical Equipment Important to Safety The NRC staff reviewed the submittal to determine the impact of the EPRI TR on the EQ of electrical equipment. The fundamental basis for 10 CFR 50.49, which governs EQ of electrical equipment important to safety for NPPs, is to ensure that such equipment can reliably perform its safety functions under normal environmental conditions, anticipated operational occurrences, and accident and post-accident environmental conditions. This includes qualification that considers environmental stressors such as temperature, pressure, humidity, chemical exposure, radiation, and aging over the electrical equipment's installed life. The EQ rule requires license holders to establish programs that verify the equipment's capability to function under the most severe design-basis events, thereby maintaining the integrity of safety systems essential for reactor coolant pressure boundary, safe shutdown, and prevention of significant radioactive release. EQ under 10 CFR 50.49 is grounded in protecting against environmentally induced common cause failures and ensuring that electrical equipment can meet its safety requirements throughout the plants operational life in accordance with quality assurance standards specified in 10 CFR Part 50, Appendix B, and GDCs 1, Quality standards and records, 2, Design bases for protection against natural phenomena, 4, Environmental and dynamic effects design bases, and 23, Protection system failure modes, in Appendix A to 10 CFR Part 50.

EPRI TR RI-HELB states, in part, that other related programs (e.g., determining the scope of equipment required to be within an EQ program) are outside the scope of this application.

During the audit, the NRC staff noted that changes to considerations and conditions for pipe breaks (e.g., location, severity, etc.) could inherently affect 10 CFR 50.49 EQ zones, which establish the environmental parameters used to determine which equipment must be qualified and to what threshold. Such changes could potentially reduce or eliminate requirements for equipment qualification. In light of this, the NRC issued an RAI requesting that EPRI provide further clarification as to why 10 CFR 50.49 is outside the scope of EPRI TR RI-HELB, given that NRC staff approval of EPRI TR RI-HELB could have a direct or indirect impact on the equipment currently required to be qualified under 10 CFR 50.49. In its response to RAI 16, EPRI stated:

The statement, other plant designs and related programs (e.g. determining the scope of equipment required to be within an EQ program) are outside the scope of this application

14 was intended to clarify that the adoption of a RI-HELB methodology into a plants design and licensing basis is plant specific. Any change to a plants design and licensing basis including any impact from adopting the RI-HELB methodology on the equipment that is currently required to be qualified per 10 CFR 50.49-would be addressed outside of the scope of EPRI TR 3003028939. The basis for this approach was intended to focus the regulatory review on the risk-informed methodology, which is limited to determining the risk significance of postulated piping failures and appropriate plant response strategies (see Sections 1 and 5 of TR 3003028939). As a result, any changes made specific to the EQ program, or other programs, would be addressed on a plant-by-plant basis as part of the existing licensing process.

The scope of equipment required to be within an EQ program could change as a result of adopting a RI-HELB approach. Plant modifications to lower the risk significance of a postulated HELB has the potential to add equipment to the scope of a § 50.49 EQ program.

The application of a RI-HELB approach could also eliminate the need to environmentally qualify electric equipment only credited to detect and/or mitigate a postulated high-energy pipe rupture, if the postulated high-energy pipe rupture is categorized by the RI-HELB plant-specific evaluations as a low-risk HELB (e.g., RC6 or RC7). For electric equipment credited in multiple HELB scenarios, the most risk-significant scenario should be used in determining HELB response strategies.

During the audit, the NRC staff also noted that EQ requires consideration of design-basis events and accidents, including HELBs. In light of this, the NRC issued an RAI requesting EPRI to further explain why EQ requirements, guidance, and expectations for considering and calculating HELBs (e.g., NUREG-0588, Interim Staff Position on EQ of Safety-Related Electrical Equipment, Revision 1 (NRC 1981)), were not addressed in EPRI TR RI-HELB. In its response to RAI 17, EPRI stated:

The response to RAI 16 clarifies that specific changes made to the EQ program as a result of adopting the methodology in EPRI TR 3003028939 would be addressed on a plant-by-plant basis as part of the License Amendment Request. The adoption of a risk-informed HELB methodology into the current design and licensing basis of a plant would be addressed in the same manner as any other change that affects fundamental inputs to the EQ program.

For currently operating plants, EQ programs would continue to address the postulated HELBs following existing deterministic HELB requirements in accordance with the response strategies delineated in EPRI TR 3003028939 for high consequence (RC1, RC2, RC4) pipe rupture events, unless additional measures are taken to reduce the consequence of failure to medium or low (See Figure 2-8). The determination of the compartmental response (e.g., temperature, pressure, humidity, & flooding) would remain consistent with existing deterministic methods along with accounting for changes in break locations or changes in mass & energy releases. HELBs classified as low risk (RC6 & RC7) do not need to consider the requirements of 10 CFR 50.49-since environmentally induced common cause failures have been addressed as part of the risk classification process.

The additional information provided by EPRI offers assurance that the implementation of the proposed RI-HELB methodology into a plants design and licensing basis will be conducted on a plant-specific basis. Furthermore, any potential impact from adopting the RI-HELB methodology on the qualification of electrical equipment will be managed outside the scope of EPRI

15 TR 3003028939, utilizing existing plant processes. This information also provides assurance that currently operating NPPs will address the effects of postulated HELBs to ensure the continued qualification of electrical equipment in accordance with the requirements of 10 CFR 50.49. Additionally, the NRC staff finds that the applicant's statements regarding HELBs classified as Low risk (RC6 and RC7) not being subject to the requirements of 10 CFR 50.49 and HELBs for High consequence (RC1, RC2, and RC4) pipe rupture events continuing to be addressed in existing EQ programs are beyond the scope of this TR, as such, the NRC staff is not establishing a position or conclusion on these statements in this SE.

Therefore, based on information presented in EPRI TR RI-HELB and EPRIs responses to the RAIs, the NRC staff determined that the proposed methodology should not adversely affect the ability of NPPs to ensure that electrical equipment important to safety remains qualified in accordance with the requirements of 10 CFR 50.49 and GDC 4 of 10 CFR Part 50, Appendix A.

5.

LIMITATIONS AND CONDITIONS The NRC staff did not identify the need to include or issue any limitations or conditions for the use of the approved version of EPRI TR RI-HELB.

6.

CONCLUSION Based on information provided in EPRI TR RI-HELB, as supplemented, the NRC staff finds that the methodology in Chapter 2 of EPRI TR RI-HELB would be consistent with the principles of risk-informed decisionmaking in RG 1.174, Revision 3, and RI HELB acceptable for use as an alternative methodology for compliance with the GDC 4 requirements specified in the TR. The TR is clear that the application of this methodology may involve updates to the EQ program and the plant PRA. Those updates are outside the scope of the TR and the staffs approval.

16

7.

REFERENCES Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities ASME, 2009a, "American Society of Mechanical Engineers / American Nuclear Society,

'Addenda to ASME/ANS RA-S-2008, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009,"

New York, NY, February 2009 ASME, 2023a, American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 1 Appendices, Nonmandatory Appendix W, 2023 edition EPRI, 2002a TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure,"

Revision B-A, February 10, 2000 (Accession No. ML013470102)

EPRI, 2002a, TR-1006937, "Extension of the EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs," April 4, 2002 (Accession No. ML021010461)

EPRI, 2024a, Letter from Michael Ruszkowski, EPRI, to NRC, "Request for NRC Review of Risk-Informed High-Energy Line Break Evaluation Requirements, EPRI Technical Report 3002028939, June 2024," July 23, 2024 (Accession No. ML24205A146)

EPRI, 2025a, Letter from Michael Ruszkowski, EPRI, to Lois James, NRC, "Electric Power Research Institute - Submittal of Responses to the US NRC's Request for Additional Information (May 2025) on EPRI Technical Report 3002028939, Risk-Informed High-Energy Line Break Evaluation Requirements," June 12, 2025 (Accession No. ML25164A225)

EPRI, 2025b, Letter from Michael Ruszkowski, EPRI, to Lois James, NRC, "Electric Power Research Institute - Submittal of Draft Version of Modifications Made to EPRI Technical Report 3002028939, 'Risk-Informed High-Energy Line Break Evaluation Requirements,' in Response to US NRC's Request for Additional Information (May 2025), June 26, 2025 (Accession No. ML25178A798)

NRC, 1981, NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," Revision 1, July 1981 (Accession No. ML031480402)

NRC, 1983, NUREG/CR-2913, Two Phase-Jet Loads, January 1983 (Accession No. ML073510076)

NRC, 1999, Letter from William Bateman, NRC, to Gary Vine, EPRI, "Safety Evaluation Report Related to EPRI Risk-Informed Inservice inspection Evaluation Procedures (EPRI TR-112657, Revision B, July 1999)," October 28, 1999 (Accession No. ML993190477)

NRC, 2002a, Letter from Cornelius Holden, Jr., NRC, to Gary Vine, EPRI, Safety Evaluation of Topical Report TR-1006937, Extension of the EPRI Risk-Informed ISI Methodology to Break Exclusion Region Program' (TAC NO. MB1344)," June 27, 2002 (Accession No. ML021790518)

NRC, 2007a, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (Accession No. ML070660036)

NRC, 2007b, NUREG-0800, Section 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," Revision 3, March 2007 (Accession No. ML070550032)

NRC, 2007c, NUREG-0800, BTP 3.3, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, Revision 3, March 2007 (Accession No. ML070800027)

17 NRC, 2009, NRC RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities," Revision 2, March 1, 2009 (Accession No. ML090410014)

NRC, 2016a, NUREG-0800, Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," Revision 3, December 2016 (Accession No. ML16088A041)

NRC, 2016b, NUREG-0800, BTP 3.4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, Revision 3, December 2016 (Accession No. ML16085A315)

NRC, 2018a, RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, January 2018 (Accession No. ML17317A256)

NRC, 2020a, RG 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020 (Accession No.ML20238B871)

NRC, 2024a, Email from Lois James, NRC to Michael Ruszkowski, EPRI, "NRC Completion Determination Form 898 for EPRI Report 3002028939, Risk-informed High Energy Line Break Evaluation Requirements (EPID L-2024-TOP-0003)," September 10, 2024 (Accession No. ML24214A023)

NRC, 2024b, Email from Lois Jame, NRC to Michael Ruszkowski, EPRI, "Regulatory Audit in NRC Staffs Support of Review of EPRI Report 3002028939, Risk-informed High Energy Line Break Evaluation Requirements (EPID L-2024-NTR-0006)," October 25, 2024 (Accession No. ML24298A056)

NRC, 2025a, Email from Lois James, NRC, to Michael Ruszkowski, EPRI, "Final Request for Additional Information - EPRI Report 3002028939, Risk-informed High Energy Line Break Evaluation Requirements (EPID L-2024-NTR-0006)," May 19, 2025 (Accession No. ML25129A086)

NRC, 2025b, Email from Lois James, NRC, to Michael Ruszkowski, EPRI, "Issuance of Regulatory Audit Report in Support of Review of EPRI Report 3002028939, Risk-informed High Energy Line Break Evaluation Requirements (EPID L-2024-NTR-0006)," August 22, 2025 (Accession No. ML25227A104)

18

8.

ABBREVIATIONS ANS American Nuclear Society ASME American Society of Mechanical Engineers BER break exclusion region CCDP conditional core damage probability CDF core damage frequency CFR Code of Federal Regulation CLERP conditional larger early release probability DEX Division of Engineering and External Hazards DID defense-in-depth DM degradation mechanisms EMIB Mechanical Engineering & Inservice Testing Branch EPRI Electric Power Research Institute EQ environmental qualification FAC flow-accelerated corrosion FMEA failure modes and effects analysis GDC general design criterion HELB high-energy line break HRA human reliability analysis ISI inservice inspection LAR license amendment requests LERF large early release frequency LOCA loss-of-coolant accident LWR light water reactor NPHP Piping and Head Penetration Branch NPP nuclear power plants NRC Nuclear Regulatory Commission NSR non-safety-related PRA probabilistic risk assessment RAI request for additional information RC risk categories RI risk-informed SE safety evaluation SR safety-related SRP standard review plan SSC systems, structures, or components TR technical report