ML25129A090

Final Request for Additional Information – EPRI Report 3002028939, Risk-informed High Energy Line Break Evaluation Requirements

ML25129A090
Person / Time
Site:
Issue date:	05/19/2025
From:	Lois James Licensing Processes Branch
To:	Ruszkowski M Electric Power Research Institute
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: ML25129A090 (2)

Text

Enclosure ELECTRIC POWER RESEARCH INSTITUTE REPORT 3002028939, RISK-INFORMED HIGH-ENERGY LINE BREAK EVALUATION REQUIREMENTS REQUESTS FOR ADDITIONAL INFORMATION

1. Mechanical Engineering & Inservice Testing Regulatory Basis: 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 systems, structures, and components (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.

RAI 1

Can licensees use Electric Power Research Institute (EPRI) Technical Report (TR) 3002028939, Risk-Informed High-Energy Line Break Evaluation Requirements

[(RI-HELB)], June 2024, to address nonconforming or degraded conditions of high-energy line break (HELB) SSCs with respect to the licensing basis? Does EPRI TR 3002028939 contain any limitations with respect to evaluating nonconforming or degraded conditions of HELB SSCs?

RAI 2

Provide explanation whether or not EPRI TR 3002028939 can be used to change the licensing basis for the plants that have implemented EPRI TR-1006937, "Extension of the EPRI Risk-Informed ISI [inservice inspection] Methodology to Break Exclusion Region Programs," April 4, 2002, as well as plants that have not implemented the EPRI TR-1006937 methodology. Specifically, the requirement for 100 percent volumetric inservice examination of all pipe welds for piping near the containment penetration area 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).

RAI 3

EPRI TR 3002028939 does not contain discussion on an appropriate zone of influence (i.e. distance of jet impingement effects from break to impacted equipment) used to determine the SSCs that are potentially subject to a HELB and/or jet impingement load.

Provide an explanation and description of the zone of influence. Note: EPRI TR-1006937 Section 2.3.6 states, in part: 6. Jet Impingement - SRP [Standard Review Plan] 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, Two-Phase Jet. Loads, January 1983, [ADAMS Accession No. ML073510076], Reference 6).

RAI 4

EPRI TR 3002028939, Section 2.1.1, 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, 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, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Revision B-A, December 1999 (ADAMS Accession No. ML013470102), for all high-energy piping are part of the EPRI TR 3002028939 methodology.

RAI 5

Regarding EPRI Report 3002028939, Section 2.4, RC5 (without flow-accelerated corrosion (FAC)) plant modification to reduce consequence to Low (RC6) or 10 percent

inspection based on degradation mechanism, the NRC staff needs clarification for the 10 percent inspection. What is the frequency/inspection method (ultrasonic testing (UT),

radiograph testing (RT) or visual testing (VT)?) for the 10 percent inspection? What is the impact of the inspection result on the Risk Characterization? Please clarify plant modifications.

RAI 6

EPRI Report 3002028939, Section 2.4, RC5 (with FAC): Ensure that FAC program addresses the most important locations (This moves the component to RC6 or RC7 depending on whether there are other degradation mechanisms besides FAC). The NRC staff need clarification how to address the piping other than the most important locations under the FAC program.

RAI 7

The degradation mechanisms used in EPRI TR 3002028939, Risk-Informed High-Energy Line Break Evaluation Requirements, June 2024, are based on EPRI TR-112657, Revision B-A, 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 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 an explanation of 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 Non-Mandatory Appendix W degradation mechanisms have been included?

c.

Table 2-5 of EPRI TR 3002028939, titled EPRI system for evaluation of pipe rupture potential provides high, medium and low pipe rupture potential based on degradation mechanisms for which a piping segment is susceptible to. What would the pipe rupture potential (high, medium or low) be for a degradation mechanism not ranked in Table 2-5 of EPRI

TR 3002028939 from ASME Section III Non-Mandatory Appendix W such as but not limited to flow inducted vibration or vibrational fatigue?

RCI 1 EPRI TR 3002028939, Section 4, Conformance with Risk-Informed Decision-Making Principles, discusses that the RI-HELB methodology is not applicable to the reactor coolant pressure boundary. Is EPRI TR 3002028939 only applicable to ASME Class 2 and 3 and non-safety-related (NSR) piping or is EPRI TR 3002028939 applicable to ASME Class 1 piping as well? Provide clarification.

2. Probabilistic Risk Assessment (PRA) Licensing Regulatory Basis: 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, states that the engineering analyses (including traditional and probabilistic analyses) conducted to justify a proposed licensing basis change should (1) be appropriate for the nature and scope of the change, (2) be based on the as-built and as-operated and maintained plant, and (3) reflect operating experience at the plant.

RAI 8

As discussed in Sections 2.1.2 and 2.1.3 of EPRI Report 3002028939, the consequence of pipe rupture is categorized in terms of conditional core damage probability (CCDP) and conditional large early release probability (CLERP). These risk metrics will be determined quantitatively using the plant-specific PRA models that reflect the as-built, as-operated plant, as discussed in Section 2 of the report. The staff noted that the scope of the PRA model may include both safety-related (SR) and NSR equipment. However, in the example provided in Section 3 of this TR, 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 SR equipment anchored or in close proximity to the shield wall.

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

RAI 9

As discussed in Sections 2.1.2 and 2.1.3 of EPRI Report 3002028939, the consequence of pipe rupture is categorized in terms of CCDP and CLERP. These risk metrics will be determined quantitatively using the plant-specific PRA models that reflect the as-built,

as-operated plant, as discussed in Section 2 of this TR. Section 2.1.3 of this TR indicated that one of the steps to quantitatively evaluate the consequences is:

Applicable Impacts are set to TRUE using basic events to simulate the impacts. The NRC staff noted that not all equipment, or associated failure modes due to HELB, may be modeled in the plant-specific PRA. 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.

RAI 10

RG 1.174, Revision 3, Section 2.1.1.2 identifies seven considerations to evaluate how the proposed licensing basis change impacts defense-in-depth. The NRC staff noted that Section 4 of the EPRI Report 3002028939 provided a discussion of the five key risk-informed decisionmaking principles that risk-informed licensing basis changes are expected to meet. It is not clear in this TR how those seven considerations have been considered. Discuss how those seven considerations have been addressed when developing the RI-HELB evaluation method. Justify that the licensing basis change using the RI-HELB method is consistent with the defense-in-depth philosophy.

RAI 11

RG 1.174, Section 3, indicates 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. The RG states that 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. Section4 of the EPRI Report 3002028939 states, in part, there are no unique aspects of the RI-HELB methodology insofar as monitoring requirements are concerned. However, in Section 2.4 of this TR, for HELB response strategies, for risk categories (RCs) RC2, RC4, and RC5 plant modifications are recommended to lower the consequences to change the RC. For example, RC1, RC2, and RC3 require a 25 percent inspection population and RC4 and RC5 require a 10 percent inspection population based on the degradation mechanism. It is not clear to staff if current inspection programs are sufficient to monitor the performance consistent with RG 1.174, Revision 3, Section 3.

Confirm that the licensee will include a description of the monitoring programs and their implementation ensuring that no unexpected adverse safety degradation occurs because of the change(s) to the licensing basis such that the RI-HELB evaluation conclusions would remain valid.

RAI 12

As discussed in Sections 2.1.2 and 2.1.3 of EPRI Report 3002028939, the consequence of pipe rupture is categorized in terms of CCDP and CLERP. Section 2.1.1 of this TR discusses the identification of important equipment that could be impacted by the spatial (indirect) effects of a HELB. Section 2.1.3 of this TR describes how impacts to equipment are assessed using the PRA when quantitatively evaluating the consequences. The NRC staff note that in human reliability analysis (HRA), available time is an important factor when evaluating the human error probability (HEP) and

available time may be reduced if a high-energy line break were to occur. A HELB can also result in a harsh environment such that local operator actions (e.g., isolation) are no longer feasible. Although this TR does discuss operator actions, it does not explicitly address potential impacts to HRA and the associated human error probabilities for operator actions. Clarify how adverse impacts to operator actions would be identified and assessed, including discussion of if HEPs would be adjusted, as part of the RI-HELB method.

3. Piping and Head Penetrations Questions Regulatory Basis: Title 10 of the CFR Part 50, Appendix A, General Design Criterion (GDC) 4 allows the use of analyses reviewed and approved by the Commission to eliminate from the design basis the dynamic effects of the pipe ruptures postulated in SRP Section 3.6.2. The staff reviews and approves the plant-specific piping system submitted from licensees and applicants to eliminate these dynamic effects. A staff approved leak-before-break (LBB) analysis permits licensees to remove protective hardware such as pipe whip restraints and jet impingement barriers, redesign pipe connected components, their supports and their internals, and other related changes in operating plants.

RAI 13

The text in EPRI TR 3002028939, Section 2.2.1 and Table 2-4 generate uncertainty as to how to assess the different degradation mechanisms. Section 2.2 of this TR 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. 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.

d.

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

e.

What updates to Table 2-4 of this TR 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.

f.

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. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, SRP 3.6.3, Leak-Before-Break Evaluation Procedures Revision 1, March 2007 (ADAMS Accession No. ML063600396), 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 are 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 operating experience (OE) since 1999, and this information should be part of the discussion.

g.

Page 26 - In the section on Pitting (PIT), 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.

RAI 14

a.

EPRI TR 3002028939, Section 2.4, HELB Response Strategies, 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.

b.

EPRI TR 3002028939, Section 2.4, HELB Response Strategies, Page 36

- What is the definition of Most important places?

c.

EPRI TR 3002028939, Section 2.4, HELB Response Strategiesm, 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.

RAI 15

a.

EPRI TR 3002028939, Section 5, Summary. 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.

b.

NUREG-0800, Section 3.6.3, Revision 1, March 2007, 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.

c.

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 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, Revision 1, March 2007 (ADAMS Accession No. ML070610277).

4. Long-Term Operations and Modernization Regulatory Basis: Title 10 of the CFR Section 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, describes a required program for qualifying the electric equipment that must include and be based on the following:

(1) Temperature and pressure. The time-dependent temperature and pressure at the location of the electric equipment important to safety must be established for the most severe design-based accident during or following which this equipment is required to remain functional.

(2) Humidity. Humidity during design-based accidents must be considered.

(3) Chemical effects. The composition of chemicals used must be at least as severe as that resulting from the most limiting mode of plant operation (e.g., containment spray, emergency core cooling, or recirculation from containment sump). If the composition of the chemical spray can be affected by equipment malfunctions, the most severe chemical spray environment that results from a single failure in the spray system must be assumed.

(4) Radiation. The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.

(5) Aging. Equipment qualified by test must be preconditioned by natural or artificial (accelerated) aging to its end-of-installed life condition. Consideration must be given to all significant types of degradation which can have an effect on the functional capability of the equipment. If preconditioning to an end-of-installed life condition is not practicable, the equipment may be preconditioned to a shorter designated life.

The equipment must be replaced or refurbished at the end of this designated life unless ongoing qualification demonstrates that the item has additional life.

(6) Submergence (if subject to being submerged).

(7) Synergistic effects. Synergistic effects must be considered when these effects are believed to have a significant effect on equipment performance.

(8) Margins. Margins must be applied to account for unquantified uncertainty, such as the effects of production variations and inaccuracies in test instruments. These margins are in addition to any conservatisms applied during the derivation of local environmental conditions of the equipment unless these conservatisms can be quantified and shown to contain appropriate margins.

RAI 16

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 10 CFR 50.49 or 10 CFR 50, Appendix A, 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.

RAI 17

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 (ADAMS Accession No. ML031480402)) were not addressed in the TR.