ML24298A068

Plan - Regulatory Audit in NRC Staff’S Support of Review of EPRI Report 3002028939, Risk-informed High Energy Line Break Evaluation Requirements

ML24298A068
Person / Time
Site:	Electric Power Research Institute
Issue date:	10/25/2024
From:	Licensing Processes Branch
To:
References
EPRI TR 3002028939, EPRI TR 3002028939, Rev 0, EPRI TR 3002028939, Revision 0, 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
Download: ML24298A068 (10)

Text

REGULATORY AUDIT PLAN BY THE OFFICE OF NUCLEAR REACTOR REGULATION IN SUPPORT OF THE REVIEW TECHNICAL REPORT 3002028939, RISK-INFORMED HIGH-ENERGY LINE BREAK EVALUATION REQUIREMENTS ELECTRIC POWER RESEARCH INSTITUTE DOCKET NO. 99902021

1.0 BACKGROUND

By letter dated July 23, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24205A146), the Electric Power Research Institute (EPRI) submitted EPRI Technical Report (TR) 3002028939, Risk-Informed High-Energy Line Break Evaluation Requirements [(RI-HELB)], to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. EPRI TR 3002028939 provides an alternative means for assessing and confirming that plant structures, systems, and components (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.

By email dated September 10, 2024 (ADAMS Accession No. ML24214A023), the NRC staff accepted EPRI TR 3002028939 for review.

The NRC staff has determined that a regulatory audit of the EPRI TR 3002028939 should be conducted in accordance with the Office of Nuclear Reactor Regulation Office Instruction (OI) LIC-111, Regulatory Audits, Revision 1, dated October 2019 (ADAMS Accession No. ML19226A274), for the NRC staff to gain a better understanding of EPRIs proposed 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.

A regulatory audit is a planned license or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. The audit is conducted with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. Performing a regulatory audit is expected to assist the NRC staff in efficiently conducting its review and gaining insights to the vendors processes and procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket.

2.0 REGULATORY AUDIT BASES Appendix A, General Design Criteria for Nuclear Power Plants, of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic licensing of production and utilization facilities, provides the principal design criteria that establish the necessary design, fabrication, construction, testing, and performance requirements for SSCs important to safety; that is, SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, states:

SSCs 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. 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.

3.0 REGULATORY AUDIT SCOPE The NRC staff will conduct a virtual audit which will include technical discussion pertaining to EPRI TR 3002028939 so the NRC staff can understand and review the proposed risk-informed alternative to the current HELB evaluation requirements. The outcome of the regulatory audit is for the NRC staff to understand the information needed to develop the safety evaluation.

4.0 INFORMATION NEEDS The areas of focus for the regulatory audit are the information contained in the TR, the audit information needs discussed below, and all associated and relevant supporting documentations (e.g., methodology, process information, calculations, etc.).

4.1 Mechanical Engineering & Inservice Testing 4.1.1 General Questions:

1. Is the intent of the EPRI TR 3002028939 methodology to be used for licensees to address HELB nonconformances and degraded conditions (operability) in addition to changing the HELB design and licensing basis? Does the TR contain any limitations with respect to evaluating nonconforming or degraded conditions?

2. Please explain whether or not this TR can be used to change the licensing basis for the plants that have implemented the EPRI TR1006937, Extension of the EPRI Risk-Informed ISI [Inservice Inspection] Methodology to Break Exclusion Region Programs, and for plants that have not implemented the EPRI TR-1006937 methodology. For example, the requirement for 100 percent volumetric inservice examination of all pipe welds should be conducted during each inspection interval as defined in IWA-2400, American Society of Mechanical Engineers (ASME)

Code,Section XI per MEB [Mechanical Engineering Branch] 3-1 or Branch Technical Position (BTP) 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment (ADAMS Accession No. ML070800008).

3. EPRI TR 3002028939 discusses that this TR will be used for plants that are implementing license renewal. Please provide explanation how this TR will be used for license renewal and how compliance with the license renewal regulatory requirements will be met.

4. EPRI TR 3002028939 does not contain any discussion on an appropriate zone of influence used to determine the SSCs that are potentially subject to a HELB and/or jet impingement load. Please provide detailed explanation on the zone of influence.

Note EPRI TR-1006937 Section 2.3.6 states, in part:

6. Jet Impingement - SRP 3.6.2 [Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping] may be used to evaluate jet impingement targets and potential load impact on structures, systems, and components. In lieu of SRP 3.6.2, plant-specific criteria and analyses may be used, and conservative assumptions and engineering judgments derived from plant design and analysis may be used as follows:

Electrical or active equipment within the zone of influence of the break is assumed to fail (e.g., active valve is assumed to fail in its normal position prior to break) unless otherwise qualified. The typical zone of influence is 10 to 20 pipe diameters (e.g. NUREG/CR 2913, Reference 6).

5. The degradation mechanisms delineated in the EPRI TR 3002028939 were derived from the existing operating reactor fleet. Can the EPRI TR 3002028939 also be used for changes to HELB requirements for new large light water reactors, new small modular reactors, advanced reactors (non-light water reactor (LWR) designs), etc.?

4.1.2 EPRI TR 3002028939, Section 2.1.1, Fundamental Principles:

1. EPRI TR 3002028939, states, in part, As application of the RI-HELB methodology applies to high-energy systems, the likelihood of having significant time available for operator actions may be limited. Typically, only automatic isolation is credited for HELB events if the event does not prevent isolation from functioning. In considering very small breaks that do not generate automatic signals, detection and isolation is considered, but the spatial impacts are much less significant and there has to be time, detection, etc. If isolation is possible, the consequence assessment should be conducted for both cases: successful and unsuccessful isolation. Operator recovery actions are further covered in Section 3.3.3.2 of EPRI TR-112657 For smaller breaks, it is anticipated that operator actions to scram the plant and turbine are expected to occur before an automatic action occurs. There are numerous indications to the operators as follows:

EPRI TR 1006937 states, in part, Physical separation can be credited with regard to the containment structure and isolation. For example, equipment inside containment can be credited with isolating a break outside containment. For high-energy line breaks, only automatic isolation can be credited, and it must be qualified per design basis.

Provide explanation whether or not operating manual recovery actions per Section 3.3.3.2 of EPRI TR-112657 for all high-energy piping are part of the EPRI TR 3002028939 methodology.

4.1.3 EPRI TR 3002028939, Section 2.2.1, DM Evaluation

1. The degradation mechanisms used in EPRI TR 3002028939 are based on EPRI TR-112657 and are amenable to mitigation by inspections.

EPRI TR-112657, states, in part, Now when considering the possible range of impacts that changes in inspection programs could conceivably have on rupture frequencies, the current service experience that is summarized in the preface to our response to requests for additional information (RAI) F-1 (on EPRI RI-ISI Methodology on TR-106706 [18]), this range is in turn limited by the fact that pipe failures can be caused by degradation mechanisms, severe loading conditions, or some combination of these. The vast majority of severe loading condition failures such as vibration fatigue, water hammer, frozen pipes and human error are not amenable to mitigation by inspections that are geared to find damage produced by an active degradation mechanism. (page 6-3)

Section 2.5.2 of EPRI TR-112657 also acknowledges that that vibrational fatigue should be treated outside the RI-ISI program.

a. Provide explanation how the degradation mechanisms which are not amenable to mitigation by inspection as described in EPRI TR-112657 such as but not limited to vibration fatigue, water hammer, flow induced vibration, etc. are addressed in EPRI TR 3002028939.

b. ASME Section III Appendix W contains degradation mechanisms which are not included in EPRI TR 3002028939. Has EPRI performed an analysis/evaluation which concludes that all applicable ASME Section III Appendix W degradation mechanisms have been included?

4.1.4 EPRI TR 3002028939, Section 2.4, HELB Response Strategies

1. RC2: Plant modification to reduce the consequence to Medium (This moves the component to RC5) plus 10 percent inspection based on the degradation mechanism. The staff needs clarification for the 10 percent inspection. What is the frequency/inspection method (UT, RT or VT?) for the 10 percent inspection? What is the impact of the inspection result to the Risk Characterization? Clarify plant modification.

2. RC5 (without flow-accelerated corrosion (FAC)) Plant modification to reduce consequence to Low (RC6) or 10 percent inspection based on degradation mechanism. The staff needs clarification for the 10 percent inspection. What is the frequency/inspection method (UT, RT or VT?) for the 10 percent inspection? What is the impact of the inspection result to the Risk Characterization? Clarify plant modification.

3. RC5 (with FAC): Ensure that FAC program is addressing the most important locations (This move the component to RC6 or RC7 depending on whether there are other degradation mechanisms besides FAC). The staff needs clarification how to address the piping other than the most important locations under the FAC program.

4. What is the basis for the 10 percent inspection? Does the 10 percent inspection result covers 100 percent inspection result?

4.1.5 EPRI TR 3002028939, Section 3.1 3, Example Application of RI-HELB methodology

1. The example provided in EPRI TR 3002028939 is based on a non-safety-related main steam piping example. Currently no restrictions exist for applying EPRI TR 3002028939 methods to safety-related and ASME Class 1, 2, and 3 piping. Will safety-related and ASME Class 1, 2, and 3 piping examples be provided to demonstrate the EPRI TR 3002028939 methodology is adequate to meet GDC-4?

4.2 Probabilistic Risk Assessment (PRA) Licensing 4.2.1 PRA Licensing Document Requests Documentation supporting the example PRA calculations presented EPRI TR 3002028939 Tables 3-1 and 3-6.

PRA notebooks for applicable systems in EPRI TR 3002028939 Section 3.

Plant-specific documentation (e.g., uncertainty notebooks) related to identification of key assumption and sources of uncertainty regarding the application of risk RI-HELB for the systems in EPRI TR 3002028939 Section 3.

4.2.2 PRA Licensing Questions

1. The consequence of pipe rupture is measured in terms of conditional core damage probability and conditional large early release probability. These measurements required quantitative risk estimates obtained from the plant-specific PRA models.

The NRC staff noted that the scope of the PRA model may include both safety-related (SR) and non-safety-related (NSR) equipment. However, in EPRI TR 3002028939 Section 3, SR equipment is specifically emphasized while observations of NSR equipment are minimal. Some examples include:

a. Compartment pressurization pressure calculations for the turbine building produced no areas of concern with respect to SR equipment.

b. The effects of pipe whip on structures and walls and SR components were calculated for postulated main steam and feedwater pipe breaks in the turbine building.

c. It was confirmed that there is no safety-related equipment anchored or in close proximity to the shield wall.

Discuss the apparent differences in addressing safety-related equipment and NSR equipment that could be in the scope of the PRA. Discuss whether additional guidance is needed for in consequence evaluation to properly assesses the impact of failed NSR equipment that could be in the scope of the PRA.

2. Internal Flooding is an integral part of the consequence evaluation. Clarify whether the internal flooding risk must be evaluated quantitatively. Clarify how the licensee would evaluate the consequence if the target equipment (or associated failure mode due to internal flooding) is not modeled in the PRA.

3. It is not clear whether the risk calculation included contribution from fire and external hazards. Clarify the treatments for the impact of fire risk and impact of other external events risks.

4. Risk-informed changes to the licensing basis should not resulted in any substantial change in the effectiveness of the barriers that prevent or mitigate radioactivity release. Release of radioactive materials from the reactor to the environment is prevented by a succession of passive barriers, including the fuel cladding, reactor coolant pressure boundary, and containment structure. RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, Section 2.1.1.2 identifies seven considerations to evaluate how the proposed licensing basis change impacts defense-in-depth. It is not clear in the report how those considerations have been considered such that the licensing basis change is consistent with the defense-in-depth philosophy.

5. The licensee should propose monitoring programs that adequately track the performance of equipment that, when degraded, can affect the conclusions of the licensees engineering evaluation and integrated decisionmaking that support the change to the licensing basis. The staff noted that in EPRI TR 3002028939 Section 4 indicates that there are no unique aspects of the RI-HELB methodology insofar as monitoring requirements are concerned. The staff noted that (1) for RC2, RC4, and RC5 where plant modifications are used to lower the consequences, and (2) RC5 involves 10 percent inspection based on degradation mechanism. It is not clear to staff if current inspection programs are sufficient to monitor the performance of equipment consistent with RG 1.174, Revision 3, Section 3. Confirm that the licensee would include a description of the implementation and monitoring program as described in RG 1.174, Revision 3, Section 3.

4.3 Piping and Head Penetrations Questions 4.3.1 EPRI TR 3002028939, Section 2.2, Failure Potential Evaluation

1. The text in Section 2.2.1 and Table 2-4 generate uncertainty as to how to assess the different degradation mechanisms. Section 2.2 uses a combination of a slightly modified version of Section 3.4.2.3 and Table 3-16 from the 1999 EPRI TR-T112657 Revision B-A, Revised Risk-Informed Inservice Inspection Procedure (ADAMS Accession No. ML013470102). Notably, EPRI TR 3002028939 Table 2-4 contains some changes that reflect the changes in operating experience, EPRI Guidance, and regulations since 1999. The criteria and susceptible regions given in EPRI TR 3002028939 Table 2-4 thus do not match the information for primary water stress corrosion cracking (PWSCC) in the text of EPRI TR 3002028939 Section 2.2.1. There are also differences between the table and the text for other degradation mechanisms.

a. Section 2.2.1 would be significantly clearer and more usable if the text and the table were consistent.

b. What updates to Table 2-4 would be appropriate given the operating experience and changes to regulations since 1999, including updates to documents like MRP-146, Implementation Survey Summary Report (MRP-275), which is now on Revision 2? MRP-146 is mentioned in the Appendix but not in the body of the proposed TR.

c. Page 26 - In the section on PWSCC for PWRs, it states that piping and attachments (i.e., thermowells) are considered susceptible to PWSCC when they are fabricated from mill-annealed Alloy 600 base material and their associated welds Alloy 82/182 that is cold worked or cold worked and welded without subsequent stress relief, are exposed to primary water, and operate at high temperatures. SRP 3.6.3 does not distinguish as to whether only cold worked Alloy 600 and 82/182 welds are susceptible to PWSCC and must be evaluated. It states that if Alloy 600/82/182 material is used, then PWSCC is a concern.

However, if Alloy 690 and 52/152 welds is utilized, PWSCC is not a concern.

Additional information needs to be provided if Alloy 600/82/182 material is used, specifically inspections, cladding/overlays or replacement with Alloy 690/52/152.

There has been much OE since 1999 and this information should be part of the discussion.

d. Page 26 - In the section on Pitting, it states that materials are susceptible to PIT, including austenitic stainless steels, nickel alloys, and carbon and low alloy steels. PIT susceptibility is a strong function of oxygen level and chloride level concentration. Please provide additional information such as EPRI water chemistry procedures are in place.

4.3.2 EPRI TR 3002028939, Section 2.4, HELB Response Strategies

1. Page 36 - What actions are taken if the FAC program is not met? What components are modified EPRI TR 3002028939 Section 2.4 states that for high-risk regions, twenty-five percent (25%) of the inspection population is performed.

Please provide details as to what type of inspections will be performed.

a. Page 36 - What is the definition of Most important places?

b. Page 71 - Under Risk -Informed Decisionmaking Principle 4, what type of examinations will be utilized to demonstrate that risk increases would be small and consistent with the intent of the NRCs policy statement on Risk -Informed Decisionmaking on safety goals for the operations of nuclear power plants.

4.3.3 EPRI TR 3002028939, Section 5, Summary

1. Page 72 - Under Section 5, Summary, please provide additional information as to why other plant designs and related programs (i.e., material modifications) are outside the scope of this application.

2. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 3.6.3, Leak-Before-Break Evaluation Procedures, states that an evaluation over the entire life of the plant for the plant piping system include environmental conditions. Does this apply to EPRI TR 3002028939? Please explain the reasoning.

3. Balance of Plant reviews include capability, reliability and sensitivity of the reactor coolant pressure boundary leakage detection systems inside containment. Please direct the NRC staff to this discussion about leakage detection systems and whether they meet Regulatory Guide (RG) 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, and NUREG-0800, Section 5.2.5, Reactor Coolant Pressure Boundary Leakage Detection.

4.4 Long-Term Operations and Modernization

1. EPRI TR 3002028939 states, in part, that other related programs (e.g.,

determining the scope of equipment required to be within an environmental qualification program) are outside the scope of this application. However, changes to considerations and conditions for pipe breaks (e.g., location, severity, etc.)

inherently could impact 10 CFR 50.49 environmental qualification (EQ) zones (which establish the environmental parameters for determining which equipment needs to be qualified and to what threshold) and reduce or remove requirements for equipment qualification (either 50.49 or GDC 4). Please provide additional explanation as to why 10 CFR 50.49 is outside the scope of this TR since NRC staff approval of EPRI TR 3002028939 could have a direct or indirect impact on the equipment that is currently required to be qualified per 10 CFR 50.49.

2. EQ requires consideration of design basis events and accidents, including HELB.

Please explain why EQ requirements, guidance, and expectations for considering and calculating HELBs (e.g., NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, Revision 1 for comment version) were not addressed in the TR.

5.0 TEAM AND REVIEW ASSIGNMENTS The audit team will consist of the following NRC staff:

NAME ASSIGNMENT DIVISION BRANCH Lois James Senior Project Manager Division of Operating Reactor Licensing (DORL)

Licensing Projects Branch (LPLB)

John Bozga Mechanical Engineer Division of Engineering and External Hazards (DEX)

Mechanical Engineering and Inservice Testing Branch (EMIB)

Kaihwa Hsu Senior Mechanical Engineer DEX EMIB Stephen Cumblidge Materials Engineer Division of New and Renewed Licenses (DNRL)

Piping and Head Penetrations Branch (NPHP)

Eric Reichelt Senior Materials Engineer DNRL NPHP Ching Ng Senior Reliability and Risk Analyst Division of Risk Assessment (DRA)

PRA Licensing Branch A (APLA)

David Gennardo Reliability and Risk Analyst DRA APLA 6.0 LOGISTICS The audit will be conducted from November 8 to December 21, 2024, through an online portal (also known as electronic portal, ePortal, or electronic reading room) established by EPRI.

The audit team will conduct a TEAMS-based entrance meeting with the vendor on November 8, 2024, for the purposes of introducing the team, discussing the scope of the audit, and describing the information to be made available on the portal. Through the audit period between November 8 to December 21, 2024, the NRC staff will hold breakout sessions with representatives of EPRI to answer audit team questions and to have technical discussions. An exit meeting/call will be held at the conclusion of the audit on December 21, 2024.

The NRC staff does not foresee the need for an onsite visit or in-person discussions between the NRC and vendor staff to discuss information to be provided on the portal at this time.

However, if the need for a such a meeting is identified in the future, the audit plan will be revised, and the schedule for the audit will be adjusted accordingly. The NRC project manager (PM) will coordinate any changes to the audit schedule and location with the vendor.

7.0 SPECIAL REQUESTS The audit team would like access to the documents listed in Section 4.0 above through an online portal that allows the audit team to access documents via the internet. The following conditions associated with the online portal must be maintained throughout the duration that the audit team has access to the online portal:

The online portal will be password-protected, and separate passwords will be assigned to the NRC staff who are participating in the audit.

The online portal will be sufficiently secure to prevent the NRC staff from printing, saving, downloading, or collecting any information on the online portal.

Conditions of use of the online portal will be displayed on the login screen and will require acknowledgement by each user.

Username and password information should be provided directly to the NRC staff. The NRC PM will provide to EPRI the names and contact information of the NRC staff who will be participating in the audit. All other communications should be coordinated through the NRC PM.

8.0 DELIVERABLES An audit summary report will be prepared within 90 days of the completion of the audit. If the NRC staff identifies information during the audit that is needed to support its regulatory decision, the NRC staff will issue RAIs to the vendor.