ML25147A123

Regulatory Audit Report in Support of Review of Electric Power Research Institute Technical Report 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, Revision 0 (EPID L-2023-TOP-0045)

ML25147A123
Person / Time
Site:	Electric Power Research Institute
Issue date:	06/12/2025
From:	Lois James Licensing Processes Branch
To:
References
EPID L-2023-TOP-0031 pre-app, EPID L-2023-TOP-0045 pre-fee, EPID L-2023-NTR-0008 post-fee, EPRI TR 3002025288, Rev 0, 3002025288, EPID L-2023-NTR-0008
Download: ML25147A123 (11)

Text

Enclosure REGULATORY AUDIT REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION IN SUPPORT OF THE REVIEW TECHNICAL REPORT 3002025288, REVISION 0 ENHANCED RISK-INFORMED CATEGORIZATION METHODOLOGY FOR PRESSURE BOUNDARY COMPONENTS ELECTRIC POWER RESEARCH INSTITUTE DOCKET NO. 99902021

1.0 BACKGROUND

By letter dated August 17, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23234A266), as supplemented on November 30, 2023 (ADAMS Accession No. ML23334A210), and June 14, 2024 (ADAMS Accession No. ML24180A016), the Electric Power Research Institute (EPRI) submitted EPRI Technical Report (TR) 3002025288, Enhanced Risk-Informed Categorization Methodology for Pressure Boundary Components, dated June 2023, to the U.S. Nuclear Regulatory Commission (NRC) for review and approval.

The EPRI TR 3002025288 presents an enhanced methodology for categorizing pressure boundary components in support of Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, applications.

By letter dated July 11, 2024 (ADAMS Accession No. ML23352A054), the NRC staff accepted EPRI TR 3002025288 for review.

By email dated August 30, 2024 (ADAMS Accession No. ML24241A160), the NRC staff forwarded an audit plan to Fernando Ferrante, Program Manager of Risk & Safety, EPRI, to initiate the audit of the TR.

The audit began with an entrance meeting between the NRC staff and EPRI on September 13, 2024, and ended with an exit meeting on December 18, 2024.

2.0 REGULATORY AUDIT BASES The NRC promulgated regulations to permit power reactor licensees and license applicants to implement an alternative regulatory framework with respect to special treatment, where special treatment refers to those requirements that provide increased assurance beyond normal industrial practices that structures, systems, and components (SSCs) perform their design-basis functions. Under this framework (10 CFR 50.69), licensees using a risk-informed process for categorizing SSCs according to their safety significance can remove SSCs of low safety significance from the scope of certain identified special treatment requirements.

The 10 CFR 50.69 process allows a licensee to categorize the safety significance of its SSCs using a robust categorization process defined in Nuclear Energy Institutes guidance in Nuclear Energy Institute (NEI) 00-04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline" (ADAMS Accession No. ML052900163), as endorsed by the NRC in Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (ADAMS Accession No. ML061090627).

3.0 REGULATORY AUDIT SCOPE The NRC staff conducted a virtual audit which included technical discussion pertaining to potential requests for additional information (RAIs) as well as comprehension questions on the TR itself.

4.0 AUDIT

SUMMARY

4.1 Audit Discussion Audit questions and discussions focused primarily on the following topic areas: probabilistic risk assessment (PRA) adequacy; plant-specific 10 CFR 50.69 examples; methodology limitations and prerequisites; and the handling of uncertainty.

4.1.1 Probabilistic Risk Assessment (PRA) Adequacy In Section 4.1 of the EPRI TR, it is stated that licensees need to ensure that prerequisites of PRA technical adequacy be maintained (Prerequisite 1). In Attachment A of the supplement dated November 30, 2023 (ADAMS Accession No. ML23334A210), the applicant also added Prerequisite 4, to ensure the PRA will reflect the as-built, as-operated plant.

4.1.2 Plant-Specific 10 CFR 50.69 Examples The proposed criteria for categorizing SSCs were compared to the risk analysis for a variety of boiling water and pressurized water designs. This analysis was done to determine if the proposed criteria would adequately capture the high-risk SSCs. Additionally, the SSCs determined to be high safety significant (HSS) were compared to the SSCs determined to be HSS by the Arkansas Nuclear One (ANO) precedent to determine how well the proposed criteria align with accepted methods for categorizing SSCs.

4.1.3 Methodology Limitations and Prerequisites The staff discussed the applicability of the methodology to designs not verified in the generation of the methodology. The vendor responded that the methodology was not intended to be applied to designs newer than the operating fleet for which the methodology was verified.

The staff engaged with vendor on the intent in identifying programs concerning degradation as necessary prerequisites to using the methodology. The vendor responded by providing examples of programs that would suffice.

4.1.4 Handling of Uncertainty The staff discussed the utility of using a factor of 3 to assess sensitivity of the methodology to uncertainty. Vendor explained that this sensitivity was drawn in analogy from NEI 00-04 recommendations concerning active systems.

4.2 Audit Documents DOCUMENT TITLE REVISION /

DATE 90-E-0053-09 Consequence Evaluation of ANO-2 Service Water System 3/26/1998 90-E-0053-12 Consequence Evaluation of ANO-2 E.F.W., Containment Spray, and Main Steam and Feedwater System Piping 8/14/1997 98-E-0058-03, R0 Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the Chemical Volume Control System at ANO-2 8/8/1997 98-E-0058-05, R0 Implementation of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure for the Low Pressure Safety Injection and Shutdown Cooling Systems at ANO-2 8/8/1997 ANO-1_N560 ANO-1 N-560 Consequence Evaluation May 1998 BWR [boiling water reactor] 4 Core Spray Passive Categorization 10CFR50.69 Passive Categorization BWR 4 Core Spray System (CS)

Not Given BWR 4 HPCI Passive Categorization 10CFR50.69 Passive Categorization BWR 4 High Pressure Coolant Injection System Not Given BWR 4 RCIC Passive Categorization 10CFR50.69 Passive Categorization BWR 4 Reactor Core Isolation System Not Given BWR 4 RHR [residual heat removal] Passive Categorization 10CFR50.69 Passive Categorization BWR 4 Residual Heat Removal System (RHR)

Not Given BWR 4 Service Water Passive Categorization 10CFR50.69 Passive Categorization BWR 4 Service Water Not Given Draft_Table 2_RHR Segment Degradation Mechanism Draft_Table 2_RHR Segment Degradation Mechanism Not Given Duke Energy RA 0282 Letter from Kevin Ellis, Duke Energy, to NRC, "Second Response to Request for Additional Information (RAI)

Regarding Proposed Alternative to Use American Society of Mechanical Engineers Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/Replacement 10/20/2023 DOCUMENT TITLE REVISION /

DATE Activities in Class 2 and 3 SystemsSection XI, Division 1,"

dated October 30, 2023 EN-DC-147 Risk-informed Repair Replacement Activities (RI-RRA) ANO Unit 1 Service Water Revision 8, 6/16/2022 EN-DC-147 R010 GGNS 10 CFR 50.69 Categorization for the SSW System 10/23/2024 EPRI Passive PB TR Audit Questions EPRI Passive PB TR Audit Questions 9/26/2024 EPRI Passive PB TR Audit Questions EPRI Passive PB TR Audit Questions 9/30/2024 EPRI Report 3002003029 Nondestructive Evaluation N-716 Revision 1 Pilot Study Results and Lessons Learned Dec 2014 EPRI Report 3002012984 10 CFR 50 69 Categorization Guidance Document Jun 2018 EPRI Report 3002024904 Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments Aug 2023 EPRI TR-107530 V2 Application of the EPRI Risk-Informed Inservice Inspection Evaluation Procedure Volume 2: BWR Pilot Study Data Dec 1997 ER-CL-330-1004 Clinton Power Station Unit 1 Risk-Informed Inservice Inspection Evaluation Fourth Ten Year Inspection Interval Revision 0 4/24/1987 Example Importance File Results for IE and IF Example Importance File Results for initiating events (IE) and internal flooding (IF)

Not Given Example Importance File Results for IE and IF Rev 1 Example Importance File Results for IE and IF Rev 1 Not Given GGNS-SE-24-00001 Categorization for the SSW System Tab J Defense-in-Depth Assessment Not Given PSA-ANO1-06-5069-05P, R0 ANO-1 50.69 Categorization: Reactor Building Spray (BS)

Passive Consequence Development in Support of 50.69 Passive Characterization 6/26/2024 PSA-ANO1-06-5069-10P, R0 ANO-1 50.69 Categorization: Core Flood (CF) Passive Consequence Development in Support of 50.69 Passive Characterization 6/26/2024 PSA-ANO1-06-5069-15P, R0 ANO-1 50.69 Categorization: Main Feedwater (MFW) Passive Consequence Development in Support of 50.69 Passive Characterization Oct 2024 Vogtle RAI Response 4 dated May 2 2014 Vogtle Electric Generating Plant Unit 1 and Unit 2 Pilot 10 CFR 50.69 License Amendment Request Response to Request for Additional Information 5/2/2014 West 3-Loop AFW

[auxiliary feedwater]

Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 3 Loop Auxiliary Feedwater (AFW)

Not Given West 3-Loop CVCS

[chemical and volume control system] & SI

[safety injection]

Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 3 Loop CVCS/SI Not Given DOCUMENT TITLE REVISION /

DATE West 3-Loop Feedwater Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse3 Loop Condensate and Feedwater System (FW)

Not Given West 3-Loop RHR Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 3 Loop Residual Heat Removal Not Given West 3-Loop SW

[service water]Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 3 Loop Service Water (SW)

Not Given West 4-Loop AFW Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 4 Loop Auxiliary Feedwater Not Given West 4-Loop CVCS Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 4 Loop Chemical Volume and Control (CVCS)

Not Given West 4-Loop Feedwater Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 4 Loop Condensate and Feedwater Not Given West 4-Loop RHR Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 4 Loop Residual Heat Removal System Not Given West 4-Loop Service Water Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 4 Loop Nuclear Cooling Water System Not Given West 4-Loop SI Passive Categorization 10 CFR 50.69 Passive Categorization Westinghouse 4 Loop Safety Injection Not Given 5.0 AUDIT PARTICIPANTS The NRC audit participants included the following NRC staff:

U.S. NUCLEAR REGULATORY COMMISSION NAME DIVISION Mihaela Biro Division of Risk Assessment (DRA)

Jeff Circle DRA Stephen Cumblidge Division of New and Renewed Licenses (DNRL)

David Gennardo DRA Gerond George Division of Operating Reactor Licenses (DORL)`

Lois James DORL Meena Khanna DRA Yiu Law Division of Reactor Oversight Samuel Lee DNRL Matthew Mitchell DNRL Demetrius Murray DORL Eric Palmer DNRL Bob Pascarelli DRA Bo Pham DORL Aida Rivera-Varona DORL Michele Sampson DORL Dan Widrevitz DNRL The EPRI audit participants team included the following staff:

ELECTRIC POWER RESEARCH INSTITUTE (EPRI)

NAME ORGANIZATION Fernando Ferrante EPRI Richard Fougerousse EPRI Doug Kull EPRI Ashley Lindeman EPRI James Moody EPRI Pat O'Regan EPRI Michael Ruszkowski EPRI Jennifer Maye Varnedoe Duke Energy

6.0 CONCLUSION

The audit accomplished the objectives and goals listed in Section 3.0 of the audit plan by allowing direct interaction with EPRI and support technical experts. The NRC staff were able to obtain clarification audit questions, to examine technical notes supporting responses to the audit questions, and to discuss differences in technical opinion. The clarifications and examination allowed the NRC staff to assess the need for RAIs and develop the RAIs more efficiently.

APPENDIX A - AUDIT QUESTIONS

1. Paragraph 50.69(e)(3) of Title 10 of the Code of Federal Regulations (10 CFR) states, the licensee shall consider data collected in Section 50.69(d)(2)(i) for RISC-3 SSCs

[Risk-Informed Safety Class (RISC)-3 structures, systems, and components (SSCs)] to determine if there are any adverse changes in performance such that the SSC unreliability values approach or exceed the values used in the evaluations conducted to satisfy Section 50.69(c)(1)(iv). The licensee shall make adjustments as necessary to the categorization or treatment processes so that the categorization process and results are maintained valid.

a. Provide industry operating experience data that was collected on failure rate changes for passive pressure-retaining SSCs that have been categorized as RISC-3 in an approved 10 CFR 50.69 program.

2. EPRI TR 3002025288 Section 4.4 describes how the methodology can be applied to a pressure boundary function of each individual plant unit as a system for 10 CFR 50.69 categorization and alternative treatment purposes. Consistent with 10 CFR 50.69 rule language and several citations in the final rules Statements of Considerations, the system boundaries for the pressure boundary function are limited to pressure retention.

a. Describe how the approach is taken in EPRI TR 3002025288 for passive pressure boundary SSC categorization as belonging to a single plant unit system comply with 10 CFR 50.69(c)(1)(v).

3. Paragraph 50.69(c)(1)(i) states that the SSC categorization process must consider results and insights from the plant-specific PRA [probabilistic risk assessment]. This PRA must at a minimum model severe accident scenarios resulting from internal initiating events occurring at full power operation. The PRA must be of sufficient quality and level of detail to support the categorization process and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC. In response, Table 7 of TR 3002025288 further states, As stated previously, the plant needs to have a robust internal events PRA, including IF

[internal flooding], that addresses failure of all pressure boundary components (main steam line breaks, main feedwater line breaks, internal flooding events, interfacing system loss-of-coolant-accident (LOCA), and so on). Because this methodology is being used in support of 10 CFR 50.69 applications, the plant-specific PRA needs to be sufficient to support the license amendment request (LAR) approval process, including consideration of PRA assumptions and sources of uncertainty.

a. Criteria 11-13 are the only criteria in the methodology that involves a direct use of the licensees PRA model-of-record. It appears that each of the three criteria have hard risk thresholds. Explain how uncertainty is taken into account within the use of these thresholds to categorize a passive pressure-retaining component. Also explain how the potential cumulative impact of changes is addressed.

b. Discuss and justify how current risk thresholds for Criteria 11-13 take into account cases of lower initiating event frequencies coupled with higher failure consequences.

Discuss how these higher failure consequences are considered.

c. As a risk-informed process, discuss how the preservation of defense-in-depth and maintenance of safety margins accounted for in using Criteria 11-13.

d. In computing the CDF/LERF and CCDP/Conditional Large Early Release Probability (CLERP) for Criteria 11-13, how are various embedded events such as recovery actions (i.e., FLEX) and human reliability analyses (HRA) taken into account?

e. For plants which have a high seismic contribution to pipe rupture, how are the various analyses (e.g., seismic PRA, seismic margins analysis (SMA)) taken into account for Criteria 11-13?

4. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV) Case N-660 and its NRC-approved replacement, ANO2-R&R-004, Revision 1, have been developed for Classes 2 and 3 components only. In the SE (ML090930246) for the ANO2 R&R-004, Revision 1, methodology, NRC staff referenced 50.55a(c)(2)(i) regarding Class 1 components. Criterion 1 in Section 4.2 of EPRI TR 3002025288 which allows for categorization of Class 1 piping as low safety significant (LSS) provided it be re-classified as non-Class 1 and following two conditions.

a. Discuss in detail how using the methodology and some plant-specific examples for cases where originally classified Class 1 SSCs could be categorized as LSS.

5. The detailed risk background information for the SSCs covered by the ten predetermined criteria were not provided in the topical report or supplemental information. Essential to the review of the TR, the NRC staff needs to evaluate the basis of the risk significance of the SSCs covered by the ten predetermined criteria for a variety of representative power plants to determine if the criteria would appropriately categorize the SSCs as HSS or LSS.

a. The NRC staff would like to review CDF, CCDP, and failure mode effects analysis (FMEA) for SSCs in the following systems:

Feedwater Piping, Heat Exchangers, Vessels, and Passive Components of Active Systems Auxiliary/Emergency Feedwater Piping, Heat Exchangers, Vessels, and Passive Components of Active Systems Emergency Core Cooling System Piping and Passive Components of Active Systems Service Water Piping Standby Liquid Control System Piping Residual Heat Removal/Decay Removal/ Shutdown Cooling System Piping, Heat Exchangers, Vessels, and Passive Components of Active Systems Pressurizer and Pressurizer relief piping and Passive Components of Active Systems Chemical Volume and Control System For:

One or more representative Westinghouse four-loop nuclear power plants (NPPs)

One or more Westinghouse three-loop NPPs A Babcock & Wilcox (B&W) Lowered Loop NPP A Combustion Engineering (CE) two-loop NPP One or more General Electric (GE) Type 4 (BWR/4) NPPs A GE Type 6 (BWR/6) NPP

6. Section 2 of EPRI TR 3002025288 describes how the 50.69 categorization process is performed in accordance with NEI guidance document NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, as endorsed in Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, Revision 1. Figure 1 of EPRI TR 3002025288 shows passive categorization as a separate path for preliminary categorization of pressure-retaining components, prior to integrated decision-making panel (IDP) review and final categorization. The NRC notes that the guidance in NEI 00-04 includes considerations for fire, seismic, and other external hazards, which may be assessed qualitatively, as well as additional qualitative criteria and requirements for assessment of defense-in-depth (DID). (These are also shown in Figure 1.) The NRC notes that the NRC-approved methodology for passive categorization in ANO2-R&R-004, Revision 1, also includes considerations for assessing shutdown operations, external events, and DID.

It is unclear how the methodology in EPRI TR 3002025288, whether independently or in conjunction with the guidance in NEI 00-04, requires the assessment of shutdown operations and external events for potential impact on the categorization of passive, pressure-retaining components.

a. Please explain how these considerations are addressed. If these considerations are addressed solely by the IDP, explain how this is communicated to the IDP.

b. Also elaborate on the assessment of qualitative criteria and DID for passive categorization, and if any additional guidance is required for the IDP when applying the methodology in EPRI TR 3002025288.

7. The document titled Clinton Power Station Unit 1 Risk-Informed Inservice Inspection Evaluation Fourth Ten Year Inspection Interval Revision 0 (Clinton N-578 RI-ISI Evaluation CoF [consequence of failure] Results.pdf) has a table on page 113 of 233 (69th page in attachment) listing the final categorizations. It lists 12 SSCs as HSS in the standby liquid control system. Which of the SSCs would be designated as HSS using the proposed categorization criteria 1-10?

8. The document titled 10CFR50.69 Passive Categorization Westinghouse 4 Loop Safety Injection (West 4-Loop SI Passive Categorization.docx) describes 32 non-class 1 SSCs in the refueling water storage tank (RWST) system as HSS on pages 28 and 29. Which of these SSCs would be designated as HSS using the proposed categorization Criteria 1-10?

9. The document titled 10CFR50.69 Passive Categorization Westinghouse 4 Loop Residual heat Removal (West 4-Loop RHR Passive Categorization.docx) on pages 16 and 17 lists 27 SSCs in the RHR-DISCH as HSS. Which of these SSCs would be designated as HSS using the proposed categorization criteria 1-10?

10. The document titled 10CFR50.69 Passive Categorization BWR 4 Residual Heat Removal System (BWR 4 RHR Passive Categorization.docx) Table 2 RHR-TORUS Components on page 18 lists 9 SSCs as HSS. Which of these SSCs would be designated as HSS using the proposed categorization criteria 1-10?

11. The document titled 10CFR50.69 Passive Categorization BWR 4 High Pressure Coolant Injection System (BWR 4 HPCI Passive Categorization.docx) Table 2 HSS Segment HPCI-STM lists 6 SSCs as HSS. Which of the SSCs would be designated as HSS using the proposed categorization criteria 1-10?

12. EPRI TR 3002025288 states that plant-specific 10 CFR 50.69 system categorization was verified for robustness through evaluation of several BWR and pressurized water reactor (PWR) plants. This group of plants is described as consisting of multiple designs and included earlier-vintage and later-vintage designs. It is unclear if this effort verified the applicability of this approach generically to new designs such as those incorporated by reference in the Appendices to 10 CFR Part 52 (e.g., AP1000, NuScale).

If the methodology was not meant to include such designs, clarify how this is controlled in the methodology.

13. EPRI TR 3002025288 includes, in Section 4.1, that robust program[s] for localized corrosion, flow accelerated corrosion (FAC), and erosion must be ensured before implementing the categorization in Section 4.2 of the methodology. The necessary quality and effectiveness of such programs is verified through, self-assessment, benchmarking, or peer review for localized corrosion; and reference to EPRI reports for FAC and erosion. The descriptions include optional language such as should.

a. It is unclear whether an applicant must meet the descriptions of the three programs or what alternatives would be acceptably similar. Clarify how this should be determined and whether optional elements of the descriptions (those including language like should) are genuinely optional.

b. On what basis are these programs determined to be sufficiently robust, and what would constitute an indication that these programs were insufficiently robust in implementation or due to future alterations?

c. The methodology does not explicitly require that these programs continue after implementation of the methodology. Would the methodology continue to be appropriate if these programs were discontinued or modified?

d. The methodology references specific revisions of EPRI reports as necessary robust programs. What would an applicant using the methodology do if/when those references were updated? Would the 50.69 categorization need to be revisited?

14. EPRI TR 3002025288 includes, in Section 4.2 under criteria 13, that users should rely on industry guidance for a number of risk impacts. It is unclear if NRC review and approval is being sought to generically accept use of unspecified industry guidance (examples are given but are not required) as being sufficient for regulatory review of performance of Criteria 13. Please confirm or clarify if this was the intent.

15. EPRI TR 3002025288 criteria 1 differentiates components based on whether the components can be isolated from the reactor coolant system by two valves in series.

Table 3 amends this to note that the piping between these two valves may be medium/low consequence. It is unclear how a valve whose function is dependent on a lower classification can retain a higher classification function as a matter of categorization. Clarify why a valve connected by, for example, RISC-3 piping to another valve should be treated as sufficiently reliable to constitute a two valves in series function as described in Criteria 1.

16. [This question expands and clarifies an earlier audit question on the same topic]

EPRI TR 3002025288, Section 4.3, states that analysis using a factor of 3 reduction in reliability for systems categorized as RISC-3 is conservative and appropriate, citing NEI 00-04. It is unclear why this factor is conservative and appropriate in the reversed context of this methodology, where components are presumed LSS by default, in contrast to the traditional 50.69 methodology which presumes components are HSS by default. Notably, the proposed methodology is relatively simplified compared to the traditional use of NEI 00-04 for supporting 10 CFR 50.69 applications which includes a relatively fine-grained assessment of subject systems.

NEI 00-04 does not state that a factor of 3 is appropriate, rather it provides a range of values useful in conducting sensitivity studies of an analysis. No basis is given for this range in NEI 00-04 beyond that it would provide trend insights for the consequences of reductions in reliability due to reduced treatments.

Clarify on what basis a factor of 3 is determined to be conservative. In particular, provide any operating experience meta-analysis and/or data distributions supporting that a factor of 3 is conservative, or realistic for passive systems.

Addressing this uncertainty is particularly important in the context of other relaxations in treatment that may occur due to changes in ASME Code requirements, for example, that may be implemented separately and concurrently with this methodology.