ML25262A158
| ML25262A158 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/19/2025 |
| From: | Spina J Arizona Public Service Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 102-09003-JLS/MDD | |
| Download: ML25262A158 (1) | |
Text
10 CFR 50.55a A member of the STARS Alliance, LLC Callaway
- Diablo Canyon
- Palo Verde
- Wolf Creek Jennifer Spina Vice President Nuclear Regulatory & Oversight Palo Verde Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072 Mail Station 7602 Tel: 623.393.4621 102-09003-JLS/MDD September 19, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Palo Verde Nuclear Generating Station (PVNGS) Unit 1 Docket No. STN 50-528 Renewed Operating License No. NPF-41 Relief Request 75 - Proposed Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval for Palo Verde Unit 1 Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, paragraph (z)(1), Arizona Public Service Company (APS) requests Nuclear Regulatory Commission (NRC) staff authorization of Relief Request 75, on the basis that the proposed alternative provides an acceptable level of quality and safety.
The Pressurized Water Reactor Owners Group (PWROG) submitted Topical Report (TR)
WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, for the NRC to review in 2006. Subsequent submittals have taken place and Revision 3 of the Topical Report [Agencywide Documents Access and Management System (ADAMS) Accession number ML11306A084] received NRC staff acceptance as documented by Safety Evaluation (SE) report dated, July 26, 2011.
WCAP-16168-NP-A, Revision 3 was developed around risk informed methodologies in Regulatory Guide 1.174. The current inspection interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174.
The proposed alternative is requested from the requirement of IWB-2411 Inspection Program, that volumetric examination of essentially 100% every 10-year interval be extended up to 20 years which will result in a reduction in man-rem exposure and examination costs. The Enclosure contains the required information needed to support a relief request for extending the Unit 1 B-A and B-D weld exams from the 1R26 outage to 1R32 plus or minus one refueling outage A pre-submittal meeting was held between APS and the NRC staff on August 7, 2025.
APS requests authorization of this proposed alternative by August 2026.
102-09003-JLS/MDD ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Relief Request 75 - Proposed Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval for Palo Verde Unit 1 Page 2 APS makes no new commitment to the NRC by this letter. Should you need further information regarding this letter, please contact Michael. D. DiLorenzo, Licensing Department Leader, at (623) 393-3495.
Sincerely, JS/MDD/cr
Enclosure:
Arizona Public Service Company, Palo Verde Nuclear Generating Station, Unit 1 Extended Reactor Vessel Inservice Interval - Relief Request Number 75 cc:
J. D. Monninger NRC Region IV Regional Administrator W. T. Orders NRC NRR Project Manager for PVNGS E. R. Lantz Senior Resident Inspector for PVNGS Spina, Jennifer (Z08962)
Digitally signed by Spina, Jennifer (Z08962)
Date: 2025.09.19 11:15:25 -07'00'
Enclosure ARIZONA PUBLIC SERVICE COMPANY PALO VERDE NUCLEAR GENERATING STATION, UNIT 1 EXTENDED REACTOR VESSEL INSERVICE INSPECTION INTERVAL Relief Request Number 75
Enclosure Relief Request 75 i
Contents Page 1.0 ASME CODE COMPONENT AFFECTED................................................................. 1 2.0 APPLICABLE CODE EDITION AND ADDENDA....................................................... 1 3.0 APPLICABLE CODE REQUIREMENTS.................................................................. 1 4.0 REASON FOR REQUEST................................................................................... 1 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE.................................................. 2 6.0 DURATION OF PROPOSED ALTERNATIVE........................................................... 7 7.0 PRECEDENTS................................................................................................. 7
8.0 REFERENCES
................................................................................................. 8 List of Tables Table 1 Critical Parameters for the Application of Bounding Analysis for PVNGS Unit 1................................................................................................... 3 Table 2 Additional Information Pertaining to Reactor Vessel Inspection for PVNGS Unit 1.....................................................................................................4 Table 3 Details of TWCF Calculation for PVNGS Unit 1 at 54 Effective Full Power Years (EFPY)............................................................................................6
Enclosure Relief Request 75 ii Nomenclature Acronym Definition ALJGW As-Left J-Groove Weld APS Arizona Public Service Company ASME American Society of Mechanical Engineers ATTB Ambient Temperature Temper Bead CEA Control Element Assembly EPFM Elastic Plastic Fracture Mechanics GTAW Gas Tungsten Arc Welding HAZ Heat Affected Zone HIC Hydrogen Induced Cracking ISI Inservice Inspection JGW J-Groove Weld LAS Low Alloy Steel LEFM Linear Elastic Fracture Mechanics NDE Nondestructive Examination NRC Nuclear Regulatory Commission OCJ One-Cycle Justification OD Outside Diameter PDI Performance Demonstration Initiative PT (Liquid) Penetrant Testing PVNGS Palo Verde Nuclear Generating Station PWHT Post Weld Heat Treatment PWSCC Primary Water Stress Corrosion Cracking PZR Pressurizer RCS Reactor Coolant System RFO Refueling Outage RG Regulatory Guide RTNDT Reference Temperature Nil Ductility Transition UT Ultrasonic Testing
Enclosure Relief Request 75 1
Proposed Alternative for Palo Verde Unit 1 In Accordance with 10 CFR 50.55a(z)(1)
-Alternative Provides Acceptable Level of Quality and Safety-
- 1. ASME Code Component(s) Affected The affected component is the Palo Verde Nuclear Generating Station (PVNGS) Unit 1 reactor vessel (RV), specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.
Category B-A welds are defined as Pressure Retaining Welds in Reactor Vessel.
Category B-D welds are defined as Full Penetration Welded Nozzles in Vessels.
Examination Category Item No.
Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section (Throughout this request, the above examination categories are referred to as the subject examinations and the ASME BPV Code,Section XI, is referred to as the Code.)
- 2. Applicable Code Edition and Addenda ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2013 Edition (Reference 1).
- 3. Applicable Code Requirement IWB-2411, Inspection Program, requires volumetric examination of essentially 100% of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. The fourth 10-year inservice inspection (ISI) interval for PVNGS Unit 1 is scheduled to end on July 17, 2028. The applicable Code for the fifth 10-year ISI will be selected in accordance with the requirements of 10 CFR 50.55a.
- 4. Reason for Request An alternative is requested from the requirement of the IWB-2411 Inspection Program, that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each 10-year interval.
Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.
Enclosure Relief Request 75 2
- 5. Proposed Alternative and Basis for Use APS proposes not to perform the ASME Code required volumetric examination of the PVNGS Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fourth inservice inspection interval, currently scheduled for Fall 2026. APS will perform the next ASME Code required volumetric examination of the PVNGS Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in the fifth inservice inspection interval in Fall 2035 plus or minus one refueling outage. The proposed inspection date is within plus or minus one refueling outage of the latest revised implementation plan, OG-10-238 (Reference 2), which reflects the next inspection being performed in 2036.
In accordance with 10 CFR 50.55a(z)(1), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).
The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval (Reference 4). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for PVNGS Unit 1 were compared to those obtained from the Combustion Engineering (CE) pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for PVNGS Unit 1 are bounded by the results of the CE pilot plant qualifies PVNGS Unit 1 for an ISI interval extension.
Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the CE pilot plant to those of PVNGS Unit 1. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.
Enclosure Relief Request 75 3
Table 1: Critical Parameters for the Application of Bounding Analysis for PVNGS Unit 1 Parameter Pilot Plant Basis Plant-Specific Basis Additional Evaluation Required?
Dominant Pressurized Thermal Shock (PTS) Transients in the NRC PTS Risk Study are Applicable NRC PTS Risk Study (Reference 5)
PTS Generalization Study (Reference
- 6)
No Through-Wall Cracking Frequency (TWCF) 3.16E-07 Events per year (Reference 4) 6.75E-13 Events per year (Calculated per Reference 4)
No Frequency and Severity of Design Basis Transients 13 heatup/cooldown cycles per year (Reference 4)
Bounded by 13 heatup/cooldown cycles per year No Cladding Layers (Single/Multiple)
Single Layer (Reference 4)
Single Layer No Table 2 below provides a summary of the latest reactor vessel inspection for PVNGS Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the PVNGS Unit 1 reactor vessel.
Enclosure Relief Request 75 4
Table 2: Additional Information Pertaining to Reactor Vessel Inspection for PVNGS Unit 1 Inspection methodology:
The latest RV ISI for PVNGS Unit 1 was conducted in accordance with the requirements of Appendix VIII of the ASME Code,Section XI, 1995 Edition with 2000 Addenda. Evaluation of recordable indications was performed to the acceptance standards of Section XI, 2001 Edition with 2003 Addenda. Future inservice inspections will be performed to ASME Section XI, Appendix VIII requirements.
Number of past inspections:
Two 10-Year inservice inspections have been performed.
Number of indications found:
There were 50 total indications identified in the beltline region during the most recently completed inservice inspection. These subsurface indications are located in the intermediate and lower shell longitudinal welds (Items 8 through 13 in in Table 3). All 50 indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Of the 50 indications, there are 10 indications within the inner 1/10th or inner 1 of the reactor vessel wall thickness. The 10 indications are acceptable per the requirements in the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).
Of the 50 indications described above, the second 10-year inspection was the first ISI examination that detected the indications except for one indication that was recorded during the previous examination. There is no site-specific flaw growth data since these indications are considered to be fabrication indications. These indications were evaluated as acceptable per ASME Section XI Table IWB-3510-1.
A disposition of the 10 flaws against the limits of the Alternate PTS Rule is shown in the table below. Nine of the flaws were located in the weld materials and one flaw was located in the plate material of the reactor vessel.
Through-Wall Extent, TWE (in.)
Scaled maximum number of flaws per 1,293 inches of weld length in the inspection volume that are greater than or equal to TWEMIN and less than TWEMAX.
Number of PVNGS Unit 1 Flaws Evaluated (Axial/Circ.)
TWEMIN TWEMIN TWEMAX 0
0.075 No Limit 0 (0/0) 0.075 0.475 215 9 (7/2) 0.125 0.475 117 2 (2/0) 0.175 0.475 29 2 (2/0) 0.225 0.475 11 1 (1/0) 0.275 0.475 5
1 (1/0) 0.325 0.475 3
1 (1/0) 0.375 0.475 1
1 (1/0) 0.425 0.475 1
0 (0/0) 0.475 Infinite 0
0 (0/0)
Enclosure Relief Request 75 5
Table 2: Additional Information Pertaining to Reactor Vessel Inspection for PVNGS Unit 1 (Continued)
Through-Wall Extent, TWE (in.)
Scaled maximum number of flaws per 11,762 square-inches of inside surface area in the inspection volume that are greater than or equal to TWEMIN and less than TWEMAX.
Number of PVNGS Unit 1 Flaws Evaluated (Axial/Circ.)
TWEMIN TWEMAX 0
0.075 No Limit 0 (0/0) 0.075 0.375 94 1 (1/0) 0.125 0.375 37 0 (0/0) 0.175 0.375 9
0 (0/0) 0.225 0.375 3
0 (0/0) 0.275 0.375 0
0 (0/0) 0.325 0.375 0
0 (0/0) 0.375 Infinite 0
0 (0/0)
The plant-specific total length (1,293 inches) of reactor vessel beltline welds that were volumetrically inspected and the plant-specific total surface area (11,762 square-inches) of reactor vessel beltline plates that were volumetrically inspected are both comprised of three intermediate shell longitudinal welds, the intermediate shell to lower shell girth weld, and three lower shell longitudinal welds. Weld length and area were also adjusted based on examination coverage of each weld as this is considered a conservative approach.
Proposed inspection schedule for balance of plant life:
The third inservice inspection is scheduled for Fall 2026. This inspection will instead be performed in Fall 2035 plus or minus one refueling outage. The proposed inspection date is within plus or minus one refueling outage of the latest revised implementation plan, OG-10-238 (Reference 2), which reflects the next inspection being performed in 2036.
Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).
Enclosure Relief Request 75 6
Table 3: Details of TWCF Calculation for PVNGS Unit 1 at 54 Effective Full Power Years (EFPY)
Inputs (1)
Inter. & Lower Shell Twall [inches]:
11.09 Upper Shell Twall [inches]:
9.06 No.
Region and Component Description Material Heat No.
Cu
[wt%]
Ni
[wt%]
R.G.
1.99 Pos.
[ºF]
RTNDT (u)
[ºF]
Fluence
[Neutron/cm2, E > 1.0 MeV]
1 Inter. Shell Plate 124-102 A
C4188-2 0.06 0.61 1.1 37 40 2.56E+19 2
Inter. Shell Plate 124-102 B
C4142-1 0.07 0.66 1.1 44 30 2.56E+19 3
Inter. Shell Plate 124-102 C
C4188-1 0.06 0.61 1.1 37 40 2.56E+19 4
Lower Shell Plate 142-102 A
62722-1 0.03 0.64 1.1 20
-20 2.56E+19 5
Lower Shell Plate 142-102 B
62467-1 0.04 0.65 1.1 26
-10 2.56E+19 6
Lower Shell Plate 142-102 C
62817-1 0.03 0.62 1.1 20
-40 2.56E+19 7
Inter. Shell to Lower Shell Girth Weld 101-171 4P7869 0.031 0.096 1.1 28.7
-70 2.56E+19 8
Inter. Shell Axial Weld 101-124 A 4P6052 0.047 0.049 1.1 30.7
-50 1.50E+19 9
Inter. Shell Axial Weld 101-124 B 4P6052 0.047 0.049 1.1 30.7
-50 1.99E+19 10 Inter. Shell Axial Weld 101-124 C 4P6052 0.047 0.049 1.1 30.7
-50 1.99E+19 11 Lower Shell Axial Weld 101-142 A 90071 0.035 0.079 1.1 29.3
-80 1.50E+19 12 Lower Shell Axial Weld 101-142 B 90071 0.035 0.079 1.1 29.3
-80 1.99E+19 13 Lower Shell Axial Weld 101-142 C 90071 0.035 0.079 1.1 29.3
-80 1.99E+19 Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)
Controlling Material Region No.
XX RTMAX-XX
[°R]
Fluence
[Neutron/cm2, E >1.0 MeV]
FF (Fluence Factor)
T30 [ºF]
TWCF95-XX Limiting Axial Weld - AW 1 and 3 2.5 543.62 1.99E+19 1.1878 43.95 0.00E+00 Limiting Plate - PL 1 and 3 2.5 546.00 2.56E+19 1.2521 46.33 2.701E-13 Limiting Circumferential Weld - CW 1 and 3 2.5 546.00 2.56E+19 1.2521 46.33 0.00E+00 TWCF95-TOTAL (AWTWCF95-AW + PLTWCF95-PL + CWTWCF95-CW):
6.75E-13 Note:
(1) Material properties are based on PVNGS Updated Final Safety Analysis Report (UFSAR) and fluence inputs are based on WCAP-16835-NP (Reference 9).
Enclosure Relief Request 75 7
- 6. Duration of Proposed Alternative This request is applicable to the PVNGS Unit 1 inservice inspection program for the fourth and fifth 10-year inspection intervals.
- 7. Precedents Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Relief Request No. RR-40, Reactor Vessel Weld Examination Interval Extension (TAC Nos. ME1634, ME1635, and ME1636), dated February 22, 2010, Agencywide Document Access and Management System (ADAMS) Accession Number ML100290415.
Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574), dated April 30, 2013, ADAMS Accession Number ML13106A140.
Vogtle Electric Generating Plant, Units 1 and 2 - Request for Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT-06 (TAC Nos. MF2596 and MF2597), dated March 20, 2014, ADAMS Accession Number ML14030A570.
Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAC Nos. MF1922 and MF1923), dated March 26, 2014, ADAMS Accession Number ML14079A546.
Sequoyah Nuclear Plant, Units 1 and 2 - Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval (TAC Nos. MF2900 and MF2901), dated August 1, 2014, ADAMS Accession Number ML14188B920.
Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596), dated December 10, 2014, ADAMS Accession Number ML14303A506.
Wolf Creek Generating Station - Request for Relief Nos. I3R-08 and I3R-09 for the Third 10-Year Inservice Inspection Program Interval (TAC Nos. MF3321 and MF3322), dated December 10, 2014, ADAMS Accession Number ML14321A864.
Callaway Plant, Unit 1 - Request for Relief I3R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAC No.
MF3876), dated February 10, 2015, ADAMS Accession Number ML15035A148.
Braidwood Station, Units 1 and 2 - Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)
(CAC Nos. MF8191 and MF8192), dated March 15, 2017, ADAMS Accession Number ML17054C255.
South Texas Project, Units 1 and 2 - Relief from the Requirements of the ASME Code Regarding the Third 10-Year Inservice Inspection Program Interval (EPID L-2018-LLR-0010), dated July 24, 2018, ADAMS Accession Number ML18177A425.
Donald C. Cook Nuclear Plant, Unit No. 1 - Approval of Alternative to the ASME Code Regarding Reactor Vessel Weld Examination - Relief Request ISIR-4-08 (EPID: L-2018-LLR-0106), dated October 26, 2018, ADAMS Accession Number ML18284A310.
R. E. Ginna Nuclear Power Plant - Issuance of Relief Request ISI-18 Regarding Fifth 10-year Inservice Inspection Program Interval (EPID L-2018-LLR-0104), dated April 22, 2019, ADAMS Accession Number ML19100A004.
Enclosure Relief Request 75 8
Point Beach Nuclear Plant, Units 1 and 2 - Approval of Relief Requests 1-RR-13 and 2-RR-13 Regarding Extension of Inspection Interval for Reactor Pressure Welds from 10 to 20 years (EPID L-2019-LLR-0060), dated March 4, 2020, ADAMS Accession Number ML20036F261.
St. Lucie Plant, Unit 2 - Authorization of RR#15 Regarding Extension of ASME Requirements Related to Reactor Pressure Vessel Weld Examinations from 10 to 20 Years (EPID L-2020-LLR-0283), dated September 30, 2021, ADAMS Accession Number ML21236A131.
Oconee Nuclear Station, Units 1, 2, and 3 - Authorization and Safety Evaluation for Alternative Reactor Pressure Vessel Inservice Inspection Intervals (EPID L-2021-LLR-0004),
dated November 19, 2021, ADAMS Accession Number ML21281A141.
Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Authorization of Relief Request Nos. 8 and 9 Regarding Extension of Inspection Interval for Reactor Pressure Vessel Welds (EPID L-2021-LLR-0038), dated May 10, 2022, ADAMS Accession Number ML22123A192.
Arkansas Nuclear One, Unit 1 - Authorization of Request for Alternative ANO1-ISI-037 Regarding Extension of Reactor Vessel Inservice Inspection Interval (EPID L-2023-LLR-0028), dated April 10, 2024, ADAMS Accession Number ML24086A541.
- 8. References
- 1. ASME Boiler and Pressure Vessel Code,Section XI, 2013 Edition, American Society of Mechanical Engineers, New York.
- 2. PWROG Letter OG-10-238, Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval. PA-MSC-0120, July 12, 2010 (ADAMS Accession Number ML11153A033).
- 3. NRC Regulatory Guide 1.174, Revision 1, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, U.S. Nuclear Regulatory Commission, November 2002 (ADAMS Accession Number ML023240437).
- 4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, October 2011 (ADAMS Accession Number ML11306A084).
- 5. NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), U.S.
Nuclear Regulatory Commission, March 2010 (ADAMS Accession Number ML15222A848).
- 6. NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants, U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482).
- 7. Code of Federal Regulations, 10 CFR Part 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Washington D.C., Federal Register, Volume 75, No. 1, dated January 4, 2010, and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010.
Enclosure Relief Request 75 9
- 8. NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988 (ADAMS Accession Number ML003740284).
- 9. Westinghouse Report, WCAP-16835-NP, Revision 2, Palo Verde Nuclear Generating Station Units 1, 2, and 3 Basis for RCS Pressure and Temperature Limits Report, September 2015.