ML25021A140

Relief Request-71: Resubmittal of Relief Request-30

ML25021A140
Person / Time
Site:	Palo Verde
Issue date:	01/28/2025
From:	Tony Nakanishi Plant Licensing Branch IV
To:	Heflin A Arizona Public Service Co
Orders, William
References
EPID L-2024-LLR-0050
Download: ML25021A140 (1)

Text

January 28, 2025 Adam Heflin Executive Vice President/

Chief Nuclear Officer Mail Station 7605 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-20341

SUBJECT:

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 -

RELIEF REQUEST-71: RESUBMITTAL OF RELIEF REQUEST-30 (EPID L-2024-LLR-0050)

Dear Adam Heflin:

By letter dated July 31, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24213A323), Arizona Public Service Company (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for approval of a proposed alternative to a Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(c),

Reactor coolant pressure boundary, American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV) Code,Section III, Code Case parameter for Palo Verde Nuclear Generating Station (Palo Verde), Units 1, 2 and 3. Specifically, the licensee requests approval of an alternative to ASME Section III, sub-section NB-3356, Code Case 1361-2, Socket Welds,Section III, to allow a diametral clearance (cMAX) of 0.062 inch between the replacement pressurizer heater sleeves and the heater sheaths, instead of 0.045 inch as specified in the Code Case. This reactor coolant pressure boundary relief request resubmittal is being tracked by the licensee as Inservice Inspection Program Relief Request (RR)-71 and is to renew existing RR-30.

The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the stresses and the fatigue usage factors of the Palo Verde, Units 1, 2, and 3 are within ASME Code requirements. The licensees current fatigue program addressed the RR-71 environmentally assisted fatigue issue. Therefore, the staff concludes that RR-71 may be authorized remain applicable.

If you have any questions, please contact the Palo Verde Project Manager, William Orders at 301-415-3329 or by email at William.Orders@nrc.gov.

Sincerely, Tony T. Nakanishi, Chief Plant Licensing Branch LPL4 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-528, STN 50-529, and STN 50-530

Enclosure:

Safety Evaluation cc: Listserv Tony T.

Nakanishi Digitally signed by Tony T. Nakanishi Date: 2025.01.28 15:42:33 -05'00'

Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 71, REQUEST FOR ALTERNATIVE TO 10 CFR 50.55a(c) REACTOR COOLANT PRESSURE BOUNDARY, AMERICAN SOCIETY OF MECHNAICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION III, CODE CASE 1361-2 PARAMETER ARIZONA PUBLIC SERVICE COMPANY PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-528, 50-529 AND 50-530

1.1 BACKGROUND

By application dated July 31, 2024 (Reference 1), Arizona Public Service Company (APS, the licensee) submitted an alternative request to the U.S. Nuclear Regulatory Commission (NRC) to renew NRC-authorized Relief Request (RR)-30 and to address environmentally assisted fatigue (EAF) that applied to wetted reactor coolant pressure boundary (RCPB) components that are monitored for fatigue as part of license extension obligations for Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Palo Verde or PVNGS). NRC-authorized RR-30 for Palo Verde is an alternative to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Code Case N-1361-2, using a maximum 0.062 inch (0.055 inch nominal) diameter clearance between the pressurizer heater and heater sleeve in lieu of the Code Case requirement of 0.045 inch. The NRC authorized RR-30 on November 19, 2004 (Reference 2).

The licensee requested that the NRC authorize RR-71 prior to entering a period of extended operation.

1.2 DESCRIPTION

OF ALTERNATIVE ASME BPV Code Case 1361-2 lists several requirements associated with this type of joint design for a heater sleeve to heater sheath fillet weld, and the design parameter relevant to this alternative request is the diametral clearance between connecting parts (cMAX) noted on figure 1 of the Code Case. Figure 1 of Code Case 1361-2 specifies cMAX to be 0.045 inch.

The licensee is requesting to use a maximum 0.062 inch (0.055 inch nominal) diametral clearance (cMAX) between the pressurizer heater sleeve and pressurizer heater sheath as an alternative to the Code Case 1361-2 requirement of 0.045 inch. The proposed alternative request applies to all 36 heater sleeves in each of the Palo Verde units pressurizer and is requested for the remainder of plant life at the three units.

The difference between these requests is the renewal requirement of meeting the environmental assisted fatigue rules that apply to the RCPB components monitored for fatigue as part of license extension commitments.

2.0 REQULATORY REQUIREMENTS Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), Acceptable level of quality and safety, a licensee may request NRC staff authorization of a proposed alternative to a 10 CFR 50.55a(c), Reactor coolant pressure boundary, ASME BPV Code,Section III, Code Case parameter.

The regulation in 10 CFR 54.21(c)(1)(iii) states, The effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

PVNGS Updated Final Safety Analysis Report (UFSAR) (Reference 3), section 19.2.1, states, in part that, The Metal Fatigue of Reactor Coolant Pressure Boundary program uses cycle counting and usage factor tracking to ensure that actual plant experience remains bounded by design assumptions and calculations reflected in the PVNGS UFSAR.

Section 50.55a, Codes and standards, of 10 CFR specifies standards approved for incorporation by reference in the NRC regulations, including ASME BPV Code,Section III.

ASME BPV Code Case 1361-2 states in part:

Appurtenances with outside diameter equal to that of 2-inch standard pipe size and less may be constructed using weld joints in accordance with Fig 1, provided the following requirements are met:

1. The design of the joint shall be such that stresses will not exceed the limits described in NB-3220 and tabulated in tables I-1.1 and I-1.2.

2. A fatigue strength reduction factor of not less than 4 shall be used in the fatigue analyses of the joints.

3. The finished welds shall be examined by a magnetic particle or by a liquid Penetrant method in accordance with Section V and the Acceptance Standards of NB-5000.

4. End closure connections may be made with fillet welds or partial penetration welds provided the conditions stated above are met.

Figure 1 note cMAX = diametral clearance between connecting parts cMAX =

0.045 in.

ASME BPV Code,Section III, subsection NB, paragraph NB-3121, states that material subject to environmental effects shall have provision made for these effects during the design or specified life of the component.

3.0 TECHNICAL EVALUATION

The proposed alternative for design impacts the ASME BPV Code stress qualification and the fatigue life. The stress qualification for both requests is the same. The fatigue life for the RR-71 request must address the EAF effect. RR-30 did not address the EAF as a license extension commitment. The evaluation for Code stress qualification and EAF license extension commitment is evaluated in following subsections.

3.1 ASME BPV Code Stress Qualification The construction Code for Palo Verde units is ASME BPV Code,Section III, 1971 Edition through Winter 1973 Addenda. The licensee performed the stress calculation using a 1.30 inches heater sleeve inner diameter with the tolerance on the heater sheath outer diameter of 1.245 inches (the maximum diametrical clearance between the components of 0.062 inch (0.055 inch nominal)). The stress calculation utilized the results from the original stress report to determine acceptability of the replacement heater sleeve/sheath weld configuration. In the stress analysis, the loads and weld cross sectional properties are modified by a ratio factor that addressed the change in diametral clearance between the parts. The NRC staff reviewed the licensees stress calculation results, which demonstrated compliance with the ASME BPV Code,Section III, paragraph NB-3220, allowable for primary membrane, primary membrane plus bending, primary plus secondary, and fatigue stresses.

The NRC staff previously authorized RR-30 with the proposed alternative at Palo Verde, for the remainder of the plant life as documented in the safety evaluation report dated November 19, 2004 (Reference 2). RR-71 does not change the Code stress qualification design. Therefore, the Code stress qualification basis for RR-30 remains acceptable for RR-71.

3.2 Renewal Evaluation The NRC staff reviewed the licensees RR-71 and performed an audit (Reference 4) to ensure that RR-71 has addressed the renewal requirement of meeting the EAF rules that apply to the RCPB components monitored for fatigue as part of license extension commitments.

As stated in the Palo Verde UFSAR section 19.2, [t]he Metal Fatigue of Reactor Coolant Pressure Boundary program uses cycle counting and usage factor tracking to ensure that actual plant experience remains bounded by design assumptions and calculations reflected in the PVNGS UFSAR. The licensees license renewal application (LRA) (Reference 5), section 4, Time Limited Aging Analyses, invokes regulatory guidance provided by NUREG-1801 (Reference 6) for aging effects. From NUREG-1801, Chapter X, Time-Limited Aging Analyses Evaluation of Aging Management Programs Under 10 CFR 54.21(C)(1)(iii), Program X.M1 Fatigue Monitoring, for purposes of monitoring and tracking, applicants should include for a set of sample reactor coolant system components, fatigue usage calculations that consider the effects of the reactor water environment. For fatigue monitoring, for purposes of monitoring and tracking, the licensee includes, for a set of sample reactor coolant system components, fatigue usage calculations that consider the effects of the reactor water environment. The set of reactor coolant system components indicate those reactor coolant components subject to the same transients. This sample set should include the locations identified in NUREG/CR-6260 (Reference 7) and additional plant-specific component locations in the RCPB if they may be more limiting than those considered in NUREG/CR-6260. An EAF screening was performed by Structural Integrity Associates (SIA) on behalf of the licensee, which identified the bounding components that must be managed for EAF during the period of extended operation.

Palo Verde Technical Specification 5.5.5 requires the establishment of a Component Cyclic or Transient Limit program to track the cyclic and transient occurrences to ensure that components are maintained within the design limits. The licensee used FatiguePro computer Code to perform cycle counting, and cycle-based fatigue cumulative usage factor calculations.

The NRC staff reviewed the licensees methodology for the Fen Calculation. The Palo Verde Fen methodology from NUREG/CR-6583 (Reference 8) for carbon/low alloy steels, NUREG/CR-5704 (Reference 9) for stainless steels, and NUREG/CR-6909 (Reference 10) for Ni-Cr-Fe (nickel-chromium-iron) is used for the material for each location. The Fen methodology is recommended by the NRC. The staff finds the Fen methodology to be acceptable.

The NRC staff reviewed the screening methodology including thermal zones grouping and the conservative Fen estimation to determine the sentinel locations chosen to have bounding estimated environmental usage factors (Uen). The bounding Fen conservatively applying the lowest strain rate and choosing a maximum temperature to determine estimated Fen. The licensee used a bounding Fen to determine the bounding Uen to screen out the locations. The licensee selected a pressurizer heater sleeve with a 1/2 nozzle repair location per screen methodology. The staff finds this acceptable.

The NRC staff also reviewed the design transients, manually counted transients, and the 60-year projection during the regulatory audit. The 60-year cycle projection was performed using the linear projection method. The staff finds this 60-year transient cycle to be appropriate. The licensee performed FatiguePro EAF analysis using the 60-year projected transient cycles.

The FatiguePro result for each Palo Verde unit shows that the EAF for the pressurizer heater sleeve is currently below the limit of 1.0 and is projected to remain 1.0 through 60 years of operation.

On the basis that the appropriated Design Transients, Transients Cycle Numbers input, NRC endorsed Fen Methodology, and the output of the EAF usage factor is less than 1.0 in accordance with the ASME BPV Code requirement, the NRC staff concludes this alternative to be acceptable.

4.0 CONCLUSION

Based on the above, the NRC staff determined that the stresses and the fatigue usage factors of the Palo Verde, Units 1, 2, and 3 are within the ASME BPV Code requirements. The licensees current fatigue program addressed the RR-71 EAF issue. Therefore, the staff concludes that RR-71 may be authorized.

5.0 REFERENCES

1.

Horton, T., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Docket Nos.

STN 50-528, 50-529, and 50-530, Renewed Operating License Nos. NPF-41, NPF-51, NPF-74 Transmittal of Relief Request (RR) No. 71: Re-Submittal of RR-30, dated July 31, 2024 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML24213A323).

2.

Gramm, R. A., U.S. Nuclear Regulatory Commission, letter to G. R. Overbeck, Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 -

Relief Request No. 30 Re: American Society of Mechanical EngineersSection III Boiler

& Pressure Vessel Code Case 1361-2 Parameter (TAC NOS. MC5080, MC5081, AND MC5082), dated November 19, 2004 (ML043240213).

3.

Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Revision 22 to Updated Final Safety Analysis Report, dated June 2023 (ML23181A166).

4.

Orders, W., U.S. Nuclear Regulatory Commission, letter to A. Heflin, Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 -

Regulatory Audit Summary Concerning Review of Relief Request No. 71: Resubmittal Of Relief Request-30 (EPID L-2024-LLR-0050), dated December 30, 2024 (ML24351A203).

5.

Mims, D. C., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, Docket Nos. STN 50-528, 50-529, and 50-530, License Renewal Application, Chapter 4, Time-Limited Aging Analyses, dated December 11, 2008 (ML083510615).

6.

U.S. Nuclear Regulatory Commission, Generic Aging Lessons Learned (GALL) Report, NUREG-1801, Volume 1, Summary, and Volume 2, Tabulation of Results, Revision 1, dated September 2005 (ML052110005 and ML052110006, respectively) and Revision 2, December 2010 (ML103490041).

7.

Idaho National Engineering Laboratory, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, NUREG/CR-6260 (INEL-95/0045), February 1995 (ML031480219).

8.

Argonne National Laboratory, Effects of LWR [Light-Water Reactor] Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, NUREG/CR-6583 (ANL-97/18), March 1998 (ML031480391).

9.

Argonne National Laboratory, Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels, NUREG/CR-5704 (ANL-98/31), April 1999 (ML031480394).

10.

Argonne National Laboratory and Electric Power Research Institute, Effect of LWR Water Environments on the Fatigue Life of Reactor Materials, Final Report, NUREG/CR-6909, Revision 1, May 2018 (ML16319A004).

Principal Contributor: K. Hsu, NRR Date: January 28, 2025

ML25021A140

• concurrence via email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*

NRR/DEX/EMIB/BC*

NAME WOrders PBlechman SBailey(TScarbrough for)

DATE 1/17/2025 1/27/2025 1/27/2025 OFFICE NRR/DORL/LPL4/BC*

NRR/DORL/LPL4/PM*

NAME TNakanishi WOrders DATE 1/28/2025 1/28/2025