L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
| ML23087A250 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/28/2023 |
| From: | Domingos C Northern States Power Company, Minnesota, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-PI-23-007 WCAP-18746-NP, Rev. 2 | |
| Download: ML23087A250 (1) | |
Text
1717 Wakonade Drive Welch, MN 55089 March 28, 2023 L-PI-23-007 10 CFR Part 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Unit 1 and Unit 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
References:
- 1) NSPM Letter L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), dated December 2, 2022. (NRC ADAMS Accession Numbers ML22343A257 and ML22343A258)
- 2) NSPM Letter L-PI-23-002, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 Reactor Vessel Material Surveillance Program Report, dated March 16, 2023. (NRC ADAMS Accession Numbers ML23075A345 thru ML23075A352)
Reference 1 included report WCAP-18746-NP, Revision 1, Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, which was prepared using the proposed methods. WCAP-18746-NP has been revised to Revision 2 as a result of the analysis of the Unit 2 Reactor Vessel Surveillance Capsule N, which was submitted to the NRC with Reference 2. WCAP-18746-NP, Revision 2, is enclosed and wholly replaces WCAP-18746-NP, Revision 1.
The information included with this letter does not alter the evaluations performed in accordance with 10 CFR 50.92 in Reference 1. In accordance with 10 CFR 50.91, a copy of this supplement is being provided to the designated Minnesota official.
Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
fl Xcel Energy
Document Control Desk L-Pl-23-007 Page 2 Please contact Mr. Jeff Kivi at (612) 330-5788 or Jeffrey.L.Kivi@xcelenergy.com if there are any questions or if additional information is needed.
I declare under penalty of perjury, that the foregoing is true and correct.
Executed on =::.:.~~""'-----"'=~~2::::::::::::-
Christopher.
Site Vice President, Montie d Prairie Island Nuclear Generating Plants Northern States Power Company - Minnesota Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota
ENCLOSURE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AND UNIT 2 WCAP-18746-NP Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation Revision 2 (104 Pages Follow)
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 WCAP-18746-NP January 2023 Revision 2 Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation
@Westinghouse
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- Electronically approved records are authenticated in the electronic document management system.
Westinghouse Electric Company LLC 1000 Westinghouse Dr.
Cranberry Township, PA 16066
© 2023 Westinghouse Electric Company LLC All Rights Reserved WCAP-18746-NP Revision 2 Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation Margaret L. Long*
Reactor Vessel/Containment Vessel Design and Analysis Riley I. Benson*
Nuclear Operations January 2023 Reviewers:
Donald M. McNutt III*
Approved:
Lynn A. Patterson*, Manager RV/CV Design and Analysis RV/CV Design and Analysis Arturo Miralles Ferrete*
Jesse J Klingensmith*, Manager Radiation Engineering and Analysis Radiation Engineering and Analysis
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 ii WCAP-18746-NP January 2023 Revision 2 RECORD OF REVISION Revision Description Completed 0
Original Issue July 2022 1
NSPM-RV000-CN-ME-000001 was updated to correct words Axial and Tangential which were swapped in Tables 4-2, 5-2, F-1, and F-2 as documented in CAP IR-2022-9008.
October 2022 2
Updated to incorporate Unit 2 Capsule N data per WCAP-18795-NP. Added wording for Unit 2 Capsule N and WCAP-18795-NP throughout document and reference sections where applicable. Revised four capsules to list five in Sections 2.1, E.1, and E.2. Deleted section that was previously Appendix E; Appendix F is now Appendix E.
Updated Tables 4-2, 5-2, 5-4, 7-5, 7-6, C-2, D-2, E-1, E-2, and E-3; and Figure C-2.
January 2023
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 iii WCAP-18746-NP January 2023 Revision 2 TABLE OF CONTENTS LIST OF TABLES....................................................................................................................................... iv LIST OF FIGURES.................................................................................................................................... vii EXECUTIVE
SUMMARY
........................................................................................................................ viii 1
INTRODUCTION........................................................................................................... 1-1 2
RADIATION ANALYSIS AND NEUTRON DOSIMETRY.......................................... 2-1
2.1 INTRODUCTION
........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS........................................................................... 2-2 2.3 NEUTRON DOSIMETRY.............................................................................................. 2-4 2.4 CALCULATIONAL UNCERTAINTIES........................................................................ 2-5 3
FRACTURE TOUGHNESS PROPERTIES.................................................................... 3-1 4
SURVEILLANCE DATA................................................................................................ 4-1 5
CHEMISTRY FACTORS................................................................................................ 5-1 6
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS... 6-1 6.1 OVERALL APPROACH................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS........................................... 6-5 6.4 BOLTUP TEMPERATURE REQUIREMENTS............................................................. 6-5 7
CALCULATION OF ADJUSTED REFERENCE TEMPERATURE............................. 7-1 8
HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.......... 8-1 9
HEATUP AMD COOLDOWN LIMITS APPLICABILITY AND MARGIN ASSESSMENT................................................................................................................ 9-1 10 REFERENCES.............................................................................................................. 10-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)...................................................... A-1 APPENDIX B OTHER RCPB FERRITIC COMPONENTS................................................................. B-1 APPENDIX C UPPER-SHELF ENERGY EVALUATION................................................................... C-1 APPENDIX D PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION................................................................................................. D-1 APPENDIX E CREDIBILITY EVALUATION OF THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE PROGRAM..................................................................................................................... E-1
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 iv WCAP-18746-NP January 2023 Revision 2 LIST OF TABLES Table 2-1 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 1............................................................................................. 2-8 Table 2-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 1......................................................................................... 2-10 Table 2-3 Calculated Surveillance Capsule Lead Factors for Unit 1............................................. 2-12 Table 2-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1............................................................................. 2-13 Table 2-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 1............................................................................. 2-14 Table 2-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1.............................................................................................. 2-15 Table 2-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 1........................................................................................................ 2-16 Table 2-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 1.................................................. 2-17 Table 2-9 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Material for Unit 1........................................................................... 2-18 Table 2-10 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 2........................................................................................... 2-19 Table 2-11 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 2......................................................................................... 2-21 Table 2-12 Calculated Surveillance Capsule Lead Factors for Unit 2............................................. 2-23 Table 2-13 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2............................................................................. 2-24 Table 2-14 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 2............................................................................. 2-25 Table 2-15 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2.............................................................................................. 2-26 Table 2-16 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 2........................................................................................................ 2-27 Table 2-17 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 2.................................................. 2-28 Table 2-18 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Materials for Unit 2.......................................................................... 2-29
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 v
WCAP-18746-NP January 2023 Revision 2 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 1 Reactor Vessel Materials................................................................ 3-2 Table 3-2 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 2 Reactor Vessel Materials................................................................ 3-3 Table 3-3 Summary of Prairie Island Units 1 and 2 Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values......................................................................................................... 3-4 Table 4-1 Prairie Island Unit 1 Surveillance Capsule Data.............................................................. 4-2 Table 4-2 Prairie Island Unit 2 Surveillance Capsule Data.............................................................. 4-3 Table 5-1 Calculation of Prairie Island Unit 1 Chemistry Factors Using Surveillance Capsule Data
......................................................................................................................................... 5-2 Table 5-2 Calculation of Prairie Island Unit 2 Chemistry Factors Using Surveillance Capsule Data
......................................................................................................................................... 5-3 Table 5-3 Summary of Prairie Island Unit 1 Positions 1.1 and 2.1 Chemistry Factors................... 5-4 Table 5-4 Summary of Prairie Island Unit 2 Positions 1.1 and 2.1 Chemistry Factors................... 5-5 Table 7-1 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 1 Reactor Vessel at 54 EFPY................. 7-3 Table 7-2 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 2 Reactor Vessel at 54 EFPY................. 7-3 Table 7-3 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location.............................................. 7-4 Table 7-4 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location.............................................. 7-5 Table 7-5 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location.............................................. 7-6 Table 7-6 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location.............................................. 7-7 Table 8-1 Summary of the ART Values Used in the Generation of the Prairie Island Units 1 and 2 Heatup and Cooldown Curves at 54 EFPY...................................................................... 8-2 Table 8-2 Prairie Island Units 1 and 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)................................................................................ 8-6 Table 8-3 Prairie Island Units 1 and 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)......................................................................... 8-8 Table A-1 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Heatup Curves........... A-2 Table A-2 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Cooldown Curves..... A-3
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 vi WCAP-18746-NP January 2023 Revision 2 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 1 Beltline Materials................................................................................................. C-3 Table C-2 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 2 Beltline Materials................................................................................................. C-4 Table D-1 RTPTS Calculations for the Prairie Island Unit 1 Reactor Vessel Materials at 54 EFPY. D-2 Table D-2 RTPTS Calculations for the Prairie Island Unit 2 Reactor Vessel Materials at 54 EFPY. D-3 Table D-3 Evaluation of Prairie Island Unit 1 ERG Limit Category............................................... D-4 Table D-4 Evaluation of Prairie Island Unit 2 ERG Limit Category............................................... D-5 Table E-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Prairie Island Unit 2 Surveillance Data...................................................................................... E-3 Table E-2 Best-Fit Evaluation for Prairie Island Unit 2 Surveillance Materials............................. E-4 Table E-3 Calculation of Residual vs. Fast Fluence for Prairie Island Unit 2................................. E-6
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 vii WCAP-18746-NP January 2023 Revision 2 LIST OF FIGURES Figure 2-1 Arrangement of Surveillance Capsules in the Prairie Island Units 1 and 2 Reactor Vessels
....................................................................................................................................... 2-30 Figure 2-2 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Core Midplane... 2-31 Figure 2-3 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Nozzle Centerline
....................................................................................................................................... 2-32 Figure 2-4 Prairie Island Units 1 and 2 Section View of the Reactor Geometry at =33°............. 2-33 Figure 8-1 Prairie Island Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100oF/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/KIc)....................................................................................................... 8-4 Figure 8-2 Prairie Island Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc).......................................................................... 8-5 Figure 9-1 Heatup Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws........................................... 9-2 Figure 9-2 Cooldown Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws........................................... 9-3 Figure C-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 1.............................................. C-5 Figure C-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 2.............................................. C-6
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 viii WCAP-18746-NP January 2023 Revision 2 EXECUTIVE
SUMMARY
This report presents the evaluation of the Prairie Island Units 1 and 2 reactor pressure vessels (RPV) with respect to reactor vessel integrity, particularly with consideration of the Prairie Island Units 1 and 2 Capsule N testing results and RAPTOR-M3G fluence analysis as documented in Section 2. Prairie Island Units 1 and 2 have been approved for license extension for a total of 60 years of operation; thus, the evaluations in this report are projected through 54 effective full-power years (EFPY), which is deemed end-of-license extension (EOLE). Note that Capsule N from Prairie Island Unit 2 has not been analyzed at the time of issuance of this report, but a revision is planned to incorporate those results.
A summary of results for the Prairie Island Units 1 and 2 reactor vessel integrity evaluations are provided below. Based on the results presented herein, it is concluded that the Prairie Island Units 1 and 2 RPV will continue to meet RPV integrity regulatory requirements through the extended period of operation.
Although reactor vessel integrity requirements continue to be met, the incorporation of the Prairie Island Units 1 and 2 Capsule N testing results and RAPTOR-M3G fluence analysis resulted in an increased limiting Adjusted Reference Temperature (ART) values for Prairie Island Units 1 and 2 through 54 EFPY.
In order to address this, heatup and cooldown Pressure-Temperature (P-T) limit curves were generated based on the revised ART values with additional margin and using the more current methodologies in WCAP-14040-A, Revision 4 and WCAP-18124-NP-A.
The P-T limit curves were generated for 54 EFPY using the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G, which is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4. Heatup rates of 60 and 100F/hr, and cooldown rates of 0 (steady-state), 20, 40, 60, and 100F/hr were used to generate the P-T limit curves, with the flange requirements and without margins for instrumentation errors. The Prairie Island Units 1 and 2 End of License Extension (EOLE) corresponding to 60 years of operation is 54 EFPY. The EOLE P-T limit curves without margins for instrumentation errors can be found in Figure 8-1 and Figure 8-2. The P-T limit curves currently contained in the Pressure-Temperature Limit Report (PTLR) were generated with the more restrictive methodology contained in Revision 2 of WCAP-14040-A. Therefore, the curves generated in this report in Figure 8-1 and Figure 8-2 are compared to the PTLR P-T limits curves to determine if they remain applicable. The comparisons of the curves are shown in Figure 9-1 and Figure 9-2 which determined the current PTLR curves are bounding through 54 EFPY.
Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 54 EFPY.
Appendix B contains discussion of the other ferritic Reactor Coolant Pressure Boundary (RCPB) components relative to P-T limits. As discussed in Appendix B, all of the other ferritic RCPB components meet the applicable requirements of Section III of the ASME Code.
Appendix C contains an upper-shelf energy (USE) evaluation for all Prairie Island Units 1 and 2 reactor vessel beltline and extended beltline materials. Per Appendix C, all beltline and extended beltline materials are projected to maintain USE values above the 50 ft-lb screening criterion per 10 CFR 50 Appendix G at 54 EFPY.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 ix WCAP-18746-NP January 2023 Revision 2 Appendix D contains a pressurized thermal shock (PTS) evaluation for all Prairie Island Units 1 and 2 reactor vessel beltline and extended beltline materials. Per Appendix D, all beltline and extended beltline materials have projected RTPTS values below the screening criteria set forth in 10 CFR 50.61. Additionally, Prairie Island Units 1 and 2 will remain in Category I of the Emergency Response Guidelines through 54 EFPY.
Appendix E previously contained the validation of the radiation transport models based on neutron dosimetry measurements for Unit 2. This section has been deleted as it was provided as a placeholder until WCAP-18795-NP for Unit 2 Surveillance N was completed. The validation of the radiation transport models based on neutron dosimetry measurements for Unit 1 is contained in WCAP-18660-NP [17] for Unit 1 and in WCAP-18795-NP [20] for Unit 2.
Appendix E contains the credibility evaluation of the Prairie Island Unit 2 surveillance program which had previously been provided in Appendix F.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 1-1 WCAP-18746-NP January 2023 Revision 2 1
INTRODUCTION The purpose of this report is to evaluate the Prairie Island Units 1 and 2 beltline materials with consideration of the testing results of the reactor vessel surveillance program Capsule N from both Units 1 and 2 to determine the impact on the Pressure-Temperature Limit Report (PTLR) P-T limit curves. The Units 1 and 2 beltline materials are also evaluated to determine their reference temperature for pressurized thermal shock (RTPTS) and upper-shelf energy (USE) values at end of license extension (EOLE), which corresponds to 54 effective full-power years (EFPY).
Reference nil-ductility transition temperature (RTNDT) increases, and the USE decreases as the material is exposed to fast-neutron irradiation. To find the most limiting RTNDT and USE at any time period in the reactor's life, Regulatory Guide 1.99 (RG 1.99), Revision 2 [1] is used to calculate the RTNDT and USE percent decease due to the associated radiation exposure. The resulting RTNDT values are used to adjust the unirradiated RTNDT (RTNDT(U)) in order to satisfy the requirements of 10 CFR Part 50.61 [10], the PTS Rule, and to verify/calculate P-T limit curves in accordance with the requirements of 10 CFR Part 50, Appendix G [4]. (Note, the methodology to calculate RTPTS is stipulated in 10 CFR Part 50.61; however, it is identical to RG 1.99.) The resulting limiting USE values are used to satisfy the requirements of 10 CFR 50, Appendix G.
The incorporation of the Unit 1 Capsule N surveillance data resulted in an increased limiting Adjusted Reference Temperature (ART). Therefore, new Prairie Island Units 1 and 2 heatup and cooldown P-T limit curves for 54 EFPY were developed. The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (plus additional uncertainties to account for future perturbations and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [2].
Specifically, the KIc stress intensity factors and the less restrictive Circ-Flaw methodology of the 1998 through the 2000 Addenda Edition of ASME Code,Section XI, Appendix G [3] were used, when applicable.
The KIc curve is a lower bound static fracture toughness curve based on crack initiation toughness. The limiting material is indexed to the KIc curve so that allowable stress intensity factors can be obtained for the material as a function of temperature. Allowable operating limits are then determined using the allowable stress intensity factors. The P-T limit curves were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [4] have been incorporated in the P-T limit curves.
The Prairie Island Units 1 and 2 PTLR currently implements P-T limit curves developed in WCAP-14780
[9]. (Note, the curves in WCAP-14780 were originally developed for Unit 1 through 35 EFPY but have since been shown to bound Unit 2 as well and have been extended to 54 EFPY.) WCAP-14780, consistent with Revision 2 of WCAP-14040-NP-A used the more restrictive KIa stress intensity factors and Axial-Flaw methodology. KIa is a toughness based on the lower bound of crack arrest toughness, and the Axial-Flaw methodology was used even though the limiting ART values are located in a circumferential weld.
Because of the different methodologies, the new curves developed herein are compared to those currently in the PTLR to determine which are bounding.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-1 WCAP-18746-NP January 2023 Revision 2 2
RADIATION ANALYSIS AND NEUTRON DOSIMETRY
2.1 INTRODUCTION
Two discrete ordinates (Sn) transport analyses were performed for the Prairie Island Units 1 and 2 reactors to determine the neutron radiation environment within the reactor pressure vessels and surveillance capsules. In these analyses, neutron exposure parameters in terms of fast neutron (E > 1.0 MeV) fluence and iron atom displacements (dpa) were established on a plant-and fuel-cycle-specific basis. The dosimetry analyses documented in WCAP-18660-NP [17] for Unit 1 and in WCAP-18795-NP [20] for Unit 2, show that the +/-20% (1) acceptance criteria specified in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [14], is met, based on comparison with the five in-vessel surveillance capsules tested to-date for Prairie Island Unit 1 and the five in-vessel surveillance capsules tested to-date for Prairie Island Unit 2, respectively. Comparisons of the results from the dosimetry evaluations with the analytical predictions served to validate the plant-specific neutron transport calculations. These validated calculations subsequently form the basis for projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY).
The use of fast neutron (E > 1.0 MeV) fluence to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. However, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.
Because of this potential shift away from a threshold fluence toward an energy-dependent damage function for data correlation, ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, [12] recommends reporting displacements per iron atom along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy-dependent dpa function to be used for this evaluation is specified in ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom [13]. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has been promulgated in Revision 2 to Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials [1].
All of the calculations and dosimetry evaluations described in this section were based on nuclear cross-section data derived from ENDF/B-VI. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [14]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET [15].
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-2 WCAP-18746-NP January 2023 Revision 2 2.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Prairie Island Units 1 and 2 reactor vessels is shown in Figure 2-1. Six irradiation capsules located between the thermal shield and vessel wall are included in each reactor design that constitutes the reactor vessel surveillance program. Capsules S, T, V, N, P, and R are located at azimuthal angles of 57°, 67°, 77°, 237°, 247°, and 257°, respectively. These full-core positions correspond to the following quadrant symmetric locations represented in Figure 2-2: 13° from the core cardinal axes (for the 77° and 257°), 23° from the core cardinal axes (for the 67° and 247°), and 33° from the core cardinal axes (for the 57° and 237°). The stainless-steel specimen containers are approximately 1-inch square in cross-section and are approximately 63 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.
From a neutronic standpoint, the surveillance capsules and associated support structures are significant.
The presence of these materials has a significant effect on both the spatial distribution of neutron exposure rate and the neutron spectrum in the vicinity of the capsules. However, the capsules are far enough apart that they do not interfere with one another. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the Prairie Island Units 1 and 2 reactor vessels and surveillance capsules, plant-specific 3D forward transport calculations were carried out to directly solve for the space-and energy-dependent neutron exposure rate, (r,,z,E).
For the Prairie Island Units 1 and 2 transport calculations, the model depicted in Figure 2-2 through Figure 2-4 was utilized. The model contained a representation of the reactor core, the reactor internals, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing these analytical models, nominal design dimensions were generally employed for the various structural components. In addition, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a plant-and fuel-cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, etc.
A section view of the model is shown in Figure 2-4. Figure 2-4 shows the RAPTOR-M3G quadrant model at the 33° azimuth displaying the 33° surveillance capsule. Both models extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation more than fourteen (14) feet below the active fuel to about fourteen (14) feet above the active fuel.
The model consists of 186 radial mesh, 200 azimuthal mesh, and 435 axial mesh. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the RAPTOR-M3G calculations was set at a value of 0.001.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-3 WCAP-18746-NP January 2023 Revision 2 The core power distributions used in the plant-specific transport analyses for the first 32 fuel cycles at each of Prairie Island Units 1 and 2 included plant-and cycle-dependent fuel assembly initial enrichments, burnups, radial and axial power distributions. Actual operating characteristics through Cycle 32 have been evaluated for each unit; projections beyond Cycle 32 for each unit were based on the plant-specific Cycle 32 spatial power distributions with a 10% bias on the peripheral and re-entrant corners, water temperatures, and reactor power level as directed by Xcel Energy. The plant-and cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions were used to develop spatial-and energy-dependent core source distributions averaged over each individual fuel cycle for each unit. Therefore, the results from the neutron transport calculations provided data in terms of unit-fuel-cycle-averaged neutron exposure rate, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.
All of the transport calculations supporting these analyses were carried out using the RAPTOR-M3G discrete ordinates code and the BUGLE-96 cross-section library [5]. The BUGLE-96 library provides a coupled 47-neutron, 20-gamma-group cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion, and angular discretization was modeled with an S16 order of angular quadrature.
Results of the discrete ordinates transport analyses pertinent to the surveillance capsule evaluations are provided in Table 2-1 through Table 2-3 and Table 2-10 through Table 2-12 for Units 1 and 2, respectively.
In Table 2-1 and Table 2-10, the calculated fast neutron fluence rate and fluence (E > 1.0 MeV) are provided at the geometric center of the capsules, as a function of irradiation time for the Prairie Island Unit 1 and Unit 2 reactors, respectively. Similar data presented in terms of iron atom displacement rate and integrated iron atom displacements are given in Table 2-2 and Table 2-11 for Units 1 and 2, respectively.
In Table 2-3 and Table 2-12, lead factors associated with surveillance capsules are provided as a function of operating time for the Prairie Island Units 1 and 2 reactors, respectively. The lead factor is defined as the ratio of the neutron fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the maximum neutron fluence (E > 1.0 MeV) at the pressure vessel clad/base metal interface.
Neutron exposure data pertinent to the Unit 1 pressure vessel clad/base metal interface are given in Table 2-4 and Table 2-5 for neutron fluence (E > 1.0 MeV) rate and fluence (E > 1.0 MeV), respectively, and in Table 2-6 and Table 2-7 for dpa/s and dpa, respectively. Neutron exposure data pertinent to the Unit 2 pressure vessel clad/base metal interface are given in Table 2-13 and Table 2-14 for neutron fluence (E > 1.0 MeV) rate and fluence (E > 1.0 MeV), respectively, and in Table 2-15 and Table 2-16 for dpa/s and dpa, respectively. In each case, the data are provided for each operating cycle of the Prairie Island Units 1 and 2 reactors, respectively. Neutron fluence (E > 1.0 MeV) and dpa are also projected to future operating times extending to 60 EFPY for each unit. The vessel exposure data are presented in terms of the maximum exposure experienced by the pressure vessel at azimuthal angles of 0°, 15°, 30°, and 45°, and at the azimuthal location providing the maximum exposure relative to the core cardinal axes.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-4 WCAP-18746-NP January 2023 Revision 2 In Table 2-8 and Table 2-9, maximum projected fluences and dpa, respectively, of the various pressure vessel materials of Unit 1 are given. In Table 2-17 and Table 2-18, maximum projected fluences and dpa, respectively, of the various pressure vessel materials of Unit 2 are given.
These data tabulations include both plant-and fuel-cycle-specific calculated neutron exposures at the end of Cycle 32 and projections to 60 EFPY. The projections beyond Cycle 32 for both units were based on the plant-specific Cycle 32 spatial power distributions with a 10% bias on the peripheral and re-entrant corners, water temperatures, and reactor power level.
2.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures reported in Section 2.2 is demonstrated by a direct comparison against the measured sensor reaction rates and a least-squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serve to validate the calculated results, only the average of the direct comparison of measured-to-calculated results for the five previously analyzed surveillance capsules removed from each of Prairie Island Unit 1 (Capsules V, P, R, S, and N) and Prairie Island Unit 2 (Capsules V, T, R, P, and N) is provided in this section of the report.
The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsules V, P, R, S, and N [17] that were withdrawn from the Unit 1 reactor, is summarized below.
Reaction M/C Reaction Rate Average
% Std. Dev.
63Cu (n,) 60Co 0.95 7.7 54Fe (n,p) 54Mn 0.86 6.7 58Ni (n,p) 58Co 0.91 5.6 238U(Cd) (n,f) 137Cs 0.97 7.6 237Np(Cd) (n,f) 137Cs 1.06 9.8 Average of M/C Results 0.94 9.8 The average measured-to-calculated (M/C) reaction rate ratios for the threshold reactions of the five previously withdrawn surveillance capsules for Prairie Island Unit 1 range from 0.86 to 1.06, and the average M/C ratio is 0.94 9.8% (1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 2.2; therefore, the calculations for those cycles are deemed applicable for Prairie Island Unit 1.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-5 WCAP-18746-NP January 2023 Revision 2 The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsules V, T, R, P, and N [20] that were withdrawn from the Unit 2 reactor, is summarized below.
Reaction M/C Reaction Rate Average
% Std. Dev.
63Cu (n,) 60Co 0.94 7.3 54Fe (n,p) 54Mn 0.90 7.6 58Ni (n,p) 58Co 0.95 4.6 238U(Cd) (n,f) 137Cs 1.01 8.7 237Np(Cd) (n,f) 137Cs 1.09 7.5 Average of M/C Results 0.98 9.8 The average measured-to-calculated (M/C) reaction rate ratios for the threshold reactions of the five previously withdrawn surveillance capsules for Prairie Island Unit 2 range from 0.90 to 1.09, and the average M/C ratio is 0.98 9.8% (1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 2.2; therefore, the calculations for those cycles are deemed applicable for Prairie Island Unit 2.
2.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Prairie Island Units 1 and 2 surveillance capsules and reactor pressure vessels is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:
- 1. Simulator Benchmark Comparisons: Comparisons of calculations with measurements from simulator benchmarks, including the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 Experiment.
- 2. Operating Reactor and Calculational Benchmarks: Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment. Also considered are comparisons of calculations to results published in the NRC fluence calculation benchmark.
- 3. Analytic Uncertainty Analysis: An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments.
- 4. Plant-Specific Benchmarking: Comparisons of the plant-specific calculations with all available dosimetry results from the Prairie Island Units 1 and 2 surveillance programs.
The first phase of the methods qualification (simulator benchmark comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase,
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-6 WCAP-18746-NP January 2023 Revision 2 however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (operating reactor and calculational benchmark comparisons) addressed uncertainties in these additional areas that are primarily methods-related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations, as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Prairie Island Units 1 and 2 analyses was established from results of these three phases of the methods qualification.
The fourth phase of the uncertainty assessment (comparisons with Prairie Island Units 1 and 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. This comparison is used only as a check and is not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures described in Section 2.2. This comparison will be completed once again for Unit 2 when the evaluation of Unit 2 Capsule N is completed. This report will be updated accordingly when that is completed.
The following table summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Westinghouse Report WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET [15].
Description Capsule and Vessel IR Simulator Benchmark Comparisons 3%
Operating Reactor and Calculational Benchmarks 5%
Analytic Uncertainty Analysis 11%
Additional Uncertainty for Factors not Explicitly Evaluated 5%
Net Calculational Uncertainty 13%
The net calculational uncertainty was determined by combining the individual components in quadrature.
Therefore, the resultant uncertainty was treated as random, and no systematic bias was applied to the analytical results.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-7 WCAP-18746-NP January 2023 Revision 2 The NRC-issued Safety Evaluation for WCAP-18124-NP-A appears in Section A of [15]. The NRC identified two Limitations and Conditions associated with the application of RAPTOR-M3G and FERRET, which are reproduced here for convenience:
- 1. Applicability of WCAP-18124-NP, Revision 0 is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided.
Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to the response parameters of interest (e.g. pressure-temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and the reactor coolant system inlet and outlet nozzles and reactor vessel internal components.
- 2. Least squares adjustment is acceptable if the adjustments to the M/C ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the discrepancy should be disqualified.
The neutron exposure values applicable to the surveillance capsules and the maximum reactor pressure vessel neutron exposure values used to derive the surveillance capsule lead factors are completely covered by the benchmarking and uncertainty analyses in WCAP-18124-NP-A. Note, however, that this report does contain neutron exposure values for materials that are outside the qualification basis of WCAP-18124-NP-A (i.e. extended beltline materials). For the materials considered to be located in the extended beltline region, a comprehensive analytical uncertainty analysis applicable to the Prairie Island Units 1 and 2 RPV extended beltline region is summarized in WCAP18124-NP-A, Revision 0, Supplement 1-P/NP [19]. Note that the NRC has issued the Final Safety Evaluation Report for WCAP18124-NP-A, Revision 0, Supplement 1-P/NP. All RPV extended beltline calculations for Prairie Island Units 1 and 2 were performed using the WCAP-18124-NP-A Revision 0 Supplement 1-P/NP methodology.
Limitation #2 applies in situations where the least-squares analysis is used to adjust the calculated values of neutron exposure. In this report, the least-squares analysis is provided only as a supplemental check on the results of the dosimetry evaluation. The least-squares analysis was not used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure. Therefore, Limitation #2 does not apply.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-8 WCAP-18746-NP January 2023 Revision 2 Table 2-1 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence Rate (n/cm2-s) 13° 23° 33° 1
1.40 1.40 1.38E+11 7.87E+10 7.59E+10 2
0.78 2.18 1.43E+11 8.53E+10 8.25E+10 3
0.86 3.03 1.53E+11 8.47E+10 7.90E+10 4
0.89 3.92 1.50E+11 8.91E+10 8.60E+10 5
0.99 4.91 1.55E+11 8.72E+10 8.29E+10 6
0.87 5.79 1.57E+11 8.96E+10 8.57E+10 7
1.01 6.80 1.38E+11 8.47E+10 9.03E+10 8
0.89 7.69 1.70E+11 9.61E+10 9.30E+10 9
0.94 8.63 1.34E+11 8.91E+10 8.85E+10 10 0.92 9.55 1.78E+11 9.07E+10 8.15E+10 11 0.93 10.48 1.83E+11 1.06E+11 9.80E+10 12 1.18 11.65 1.32E+11 9.06E+10 8.77E+10 13 1.24 12.89 9.65E+10 7.01E+10 6.70E+10 14 1.21 14.11 7.82E+10 5.70E+10 5.66E+10 15 1.25 15.36 7.91E+10 5.59E+10 5.54E+10 16 1.29 16.65 9.15E+10 6.86E+10 6.29E+10 17 1.47 18.12 9.34E+10 6.88E+10 6.13E+10 18 1.55 19.68 8.21E+10 5.66E+10 5.35E+10 19 1.21 20.89 8.25E+10 6.71E+10 6.41E+10 20 1.61 22.50 8.97E+10 6.58E+10 5.97E+10 21 1.60 24.09 8.88E+10 5.82E+10 5.52E+10 22 1.72 25.81 9.07E+10 6.26E+10 6.19E+10 23 1.36 27.17 9.97E+10 6.72E+10 6.37E+10 24 1.61 28.78 8.79E+10 5.96E+10 5.60E+10 25 1.43 30.21 8.92E+10 6.34E+10 6.25E+10 26 1.42 31.63 9.10E+10 6.08E+10 5.84E+10 27 1.33 32.96 8.95E+10 5.59E+10 5.36E+10 28 1.74 34.71 8.79E+10 5.60E+10 5.37E+10 29 1.69 36.40 9.11E+10 6.01E+10 5.87E+10 30 1.79 38.19 8.87E+10 6.13E+10 6.30E+10 31 1.87 40.06 9.16E+10 5.93E+10 5.81E+10 32 2.03 42.08 8.77E+10 5.88E+10 5.96E+10
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-9 WCAP-18746-NP January 2023 Revision 2 Table 2-1 (continued) Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence (n/cm2)
V (13°)
P (23°)
R (13°)
S (33°)
N (33°)
T (23°)
1 1.40 1.40 6.09E+18 3.47E+18 6.09E+18 3.35E+18 3.35E+18 3.47E+18 2
0.78 2.18 5.57E+18 9.60E+18 5.38E+18 5.38E+18 5.57E+18 3
0.86 3.03 7.86E+18 1.37E+19 7.51E+18 7.51E+18 7.86E+18 4
0.89 3.92 1.04E+19 1.79E+19 9.93E+18 9.93E+18 1.04E+19 5
0.99 4.91 1.31E+19 2.28E+19 1.25E+19 1.25E+19 1.31E+19 6
0.87 5.79 2.71E+19 1.49E+19 1.49E+19 1.55E+19 7
1.01 6.80 3.15E+19 1.78E+19 1.78E+19 1.83E+19 8
0.89 7.69 3.63E+19 2.04E+19 2.04E+19 2.09E+19 9
0.94 8.63 4.02E+19 2.30E+19 2.30E+19 2.36E+19 10 0.92 9.55 2.54E+19 2.54E+19 2.62E+19 11 0.93 10.48 2.82E+19 2.82E+19 2.93E+19 12 1.18 11.65 3.15E+19 3.15E+19 3.27E+19 13 1.24 12.89 3.41E+19 3.41E+19 3.54E+19 14 1.21 14.11 3.63E+19 3.63E+19 3.76E+19 15 1.25 15.36 3.85E+19 3.85E+19 3.98E+19 16 1.29 16.65 4.10E+19 4.10E+19 4.26E+19 17 1.47 18.12 4.39E+19 4.39E+19 4.58E+19 18 1.55 19.68 4.65E+19 4.86E+19 19 1.21 20.89 4.90E+19 5.12E+19 20 1.61 22.50 5.20E+19 5.45E+19 21 1.60 24.09 5.48E+19 5.74E+19 22 1.72 25.81 5.81E+19 6.08E+19 23 1.36 27.17 6.08E+19 6.37E+19 24 1.61 28.78 6.37E+19 6.67E+19 25 1.43 30.21 6.65E+19 6.96E+19 26 1.42 31.63 6.91E+19 7.23E+19 27 1.33 32.96 7.14E+19 7.47E+19 28 1.74 34.71 7.44E+19 7.78E+19 29 1.69 36.40 7.75E+19 8.10E+19 30 1.79 38.19 8.10E+19 8.44E+19 31 1.87 40.06 8.45E+19 8.79E+19 32 2.03 42.08 9.17E+19 48 1.04E+20 51 1.10E+20 54 1.16E+20 60 1.28E+20
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-10 WCAP-18746-NP January 2023 Revision 2 Table 2-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Iron Atom Displacement Rate (dpa/s) 13° 23° 33° 1
1.40 1.40 2.53E-10 1.38E-10 1.34E-10 2
0.78 2.18 2.61E-10 1.50E-10 1.46E-10 3
0.86 3.03 2.80E-10 1.49E-10 1.39E-10 4
0.89 3.92 2.75E-10 1.56E-10 1.52E-10 5
0.99 4.91 2.84E-10 1.53E-10 1.46E-10 6
0.87 5.79 2.87E-10 1.57E-10 1.51E-10 7
1.01 6.80 2.53E-10 1.49E-10 1.59E-10 8
0.89 7.69 3.12E-10 1.69E-10 1.64E-10 9
0.94 8.63 2.44E-10 1.56E-10 1.56E-10 10 0.92 9.55 3.27E-10 1.59E-10 1.43E-10 11 0.93 10.48 3.37E-10 1.86E-10 1.73E-10 12 1.18 11.65 2.40E-10 1.58E-10 1.55E-10 13 1.24 12.89 1.75E-10 1.22E-10 1.18E-10 14 1.21 14.11 1.42E-10 9.90E-11 9.93E-11 15 1.25 15.36 1.43E-10 9.72E-11 9.71E-11 16 1.29 16.65 1.66E-10 1.19E-10 1.11E-10 17 1.47 18.12 1.69E-10 1.20E-10 1.08E-10 18 1.55 19.68 1.49E-10 9.84E-11 9.37E-11 19 1.21 20.89 1.49E-10 1.17E-10 1.13E-10 20 1.61 22.50 1.63E-10 1.14E-10 1.05E-10 21 1.60 24.09 1.61E-10 1.01E-10 9.68E-11 22 1.72 25.81 1.65E-10 1.09E-10 1.09E-10 23 1.36 27.17 1.81E-10 1.17E-10 1.12E-10 24 1.61 28.78 1.59E-10 1.04E-10 9.82E-11 25 1.43 30.21 1.62E-10 1.10E-10 1.10E-10 26 1.42 31.63 1.65E-10 1.06E-10 1.03E-10 27 1.33 32.96 1.63E-10 9.74E-11 9.40E-11 28 1.74 34.71 1.60E-10 9.76E-11 9.42E-11 29 1.69 36.40 1.66E-10 1.05E-10 1.03E-10 30 1.79 38.19 1.61E-10 1.07E-10 1.11E-10 31 1.87 40.06 1.66E-10 1.03E-10 1.02E-10 32 2.03 42.08 1.59E-10 1.02E-10 1.05E-10
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-11 WCAP-18746-NP January 2023 Revision 2 Table 2-2 (continued) Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Displacements (dpa)
V (13°)
P (23°)
R (13°)
S (33°)
N (33°)
T (23°)
1 1.40 1.40 1.12E-02 6.09E-03 1.12E-02 5.91E-03 5.91E-03 6.09E-03 2
0.78 2.18 9.76E-03 1.76E-02 9.49E-03 9.49E-03 9.76E-03 3
0.86 3.03 1.38E-02 2.52E-02 1.32E-02 1.32E-02 1.38E-02 4
0.89 3.92 1.82E-02 3.29E-02 1.75E-02 1.75E-02 1.82E-02 5
0.99 4.91 2.29E-02 4.18E-02 2.21E-02 2.21E-02 2.29E-02 6
0.87 5.79 4.97E-02 2.62E-02 2.62E-02 2.73E-02 7
1.01 6.80 5.78E-02 3.13E-02 3.13E-02 3.20E-02 8
0.89 7.69 6.65E-02 3.59E-02 3.59E-02 3.67E-02 9
0.94 8.63 7.38E-02 4.06E-02 4.06E-02 4.14E-02 10 0.92 9.55 4.47E-02 4.47E-02 4.60E-02 11 0.93 10.48 4.98E-02 4.98E-02 5.14E-02 12 1.18 11.65 5.55E-02 5.55E-02 5.73E-02 13 1.24 12.89 6.02E-02 6.02E-02 6.21E-02 14 1.21 14.11 6.40E-02 6.40E-02 6.59E-02 15 1.25 15.36 6.78E-02 6.78E-02 6.97E-02 16 1.29 16.65 7.23E-02 7.23E-02 7.46E-02 17 1.47 18.12 7.73E-02 7.73E-02 8.01E-02 18 1.55 19.68 8.19E-02 8.49E-02 19 1.21 20.89 8.62E-02 8.94E-02 20 1.61 22.50 9.15E-02 9.52E-02 21 1.60 24.09 9.64E-02 1.00E-01 22 1.72 25.81 1.02E-01 1.06E-01 23 1.36 27.17 1.07E-01 1.11E-01 24 1.61 28.78 1.12E-01 1.17E-01 25 1.43 30.21 1.17E-01 1.21E-01 26 1.42 31.63 1.22E-01 1.26E-01 27 1.33 32.96 1.26E-01 1.30E-01 28 1.74 34.71 1.31E-01 1.36E-01 29 1.69 36.40 1.36E-01 1.41E-01 30 1.79 38.19 1.43E-01 1.47E-01 31 1.87 40.06 1.49E-01 1.53E-01 32 2.03 42.08 1.60E-01 48 1.81E-01 51 1.91E-01 54 2.02E-01 60 2.23E-01
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-12 WCAP-18746-NP January 2023 Revision 2 Table 2-3 Calculated Surveillance Capsule Lead Factors for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Lead Factor 13° 23° 33° 1
1.40 1.40 2.98(a) 1.70 1.64 2
0.78 2.18 3.04 1.76 1.70 3
0.86 3.03 3.06 1.75 1.67 4
0.89 3.92 3.07 1.77 1.70 5
0.99 4.91 3.07 1.76(b) 1.69 6
0.87 5.79 3.07 1.76 1.69 7
1.01 6.80 3.11 1.80 1.75 8
0.89 7.69 3.10 1.79 1.74 9
0.94 8.63 3.08(c) 1.80 1.76 10 0.92 9.55 3.07 1.77 1.72 11 0.93 10.48 3.07 1.77 1.71 12 1.18 11.65 3.06 1.80 1.73 13 1.24 12.89 3.07 1.83 1.76 14 1.21 14.11 3.08 1.86 1.79 15 1.25 15.36 3.09 1.88 1.81 16 1.29 16.65 3.10 1.91 1.84 17 1.47 18.12 3.12 1.94 1.86(d) 18 1.55 19.68 3.13 1.96 1.87 19 1.21 20.89 3.14 1.99 1.90 20 1.61 22.50 3.15 2.01 1.92 21 1.60 24.09 3.14 2.01 1.92 22 1.72 25.81 3.15 2.02 1.93 23 1.36 27.17 3.15 2.02 1.93 24 1.61 28.78 3.15 2.03 1.94 25 1.43 30.21 3.15 2.04 1.95 26 1.42 31.63 3.16 2.05 1.95 27 1.33 32.96 3.16 2.04 1.95 28 1.74 34.71 3.16 2.04 1.95 29 1.69 36.40 3.15 2.04 1.95 30 1.79 38.19 3.16 2.05 1.97 31 1.87 40.06 3.15 2.05 1.97(e) 32 2.03 42.08 3.15 2.05 1.97 48(f) 3.15 2.05 1.99 51 3.15 2.06 2.00 54 3.15 2.06 2.00 60 3.15 2.06 2.01 Notes:
(a) Capsule V was removed after Cycle 1.
(b) Capsule P was removed after Cycle 5.
(c) Capsule R was removed after Cycle 9.
(d) Capsule S was removed after Cycle 17.
(e) Capsule N was removed after Cycle 31.
(f) The projections beyond Cycle 32 are based on Cycle 32 with a 10%
bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-13 WCAP-18746-NP January 2023 Revision 2 Table 2-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence Rate (n/cm2-s) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.40 1.40 4.61E+10 2.74E+10 1.82E+10 1.60E+10 4.64E+10
-73 2
0.78 2.18 4.77E+10 2.95E+10 2.03E+10 1.73E+10 4.79E+10 121 3
0.86 3.03 4.96E+10 2.93E+10 1.88E+10 1.70E+10 4.98E+10
-1 4
0.89 3.92 4.81E+10 2.90E+10 2.03E+10 1.75E+10 4.83E+10 3
5 0.99 4.91 5.05E+10 3.01E+10 1.97E+10 1.78E+10 5.08E+10 67 6
0.87 5.79 5.06E+10 3.06E+10 2.03E+10 1.75E+10 5.09E+10
-73 7
1.01 6.80 4.13E+10 2.68E+10 2.07E+10 1.92E+10 4.16E+10 5
8 0.89 7.69 5.57E+10 3.24E+10 2.20E+10 1.81E+10 5.60E+10 5
9 0.94 8.63 4.60E+10 2.66E+10 2.09E+10 1.73E+10 4.62E+10 3
10 0.92 9.55 5.84E+10 3.36E+10 1.95E+10 1.86E+10 5.87E+10 5
11 0.93 10.48 5.98E+10 3.52E+10 2.36E+10 1.73E+10 6.02E+10 5
12 1.18 11.65 4.43E+10 2.65E+10 2.10E+10 1.63E+10 4.45E+10 5
13 1.24 12.89 2.94E+10 2.00E+10 1.61E+10 1.44E+10 2.96E+10 5
14 1.21 14.11 2.39E+10 1.64E+10 1.35E+10 1.27E+10 2.40E+10 73 15 1.25 15.36 2.36E+10 1.63E+10 1.31E+10 1.26E+10 2.37E+10 5
16 1.29 16.65 2.76E+10 1.91E+10 1.54E+10 1.29E+10 2.77E+10 3
17 1.47 18.12 2.75E+10 1.95E+10 1.51E+10 1.23E+10 2.76E+10 5
18 1.55 19.68 2.51E+10 1.69E+10 1.29E+10 1.17E+10 2.52E+10 5
19 1.21 20.89 2.31E+10 1.77E+10 1.55E+10 1.36E+10 2.32E+10 67 20 1.61 22.50 2.75E+10 1.92E+10 1.50E+10 1.25E+10 2.76E+10 71 21 1.60 24.09 2.94E+10 1.80E+10 1.33E+10 1.19E+10 2.96E+10 5
22 1.72 25.81 2.87E+10 1.88E+10 1.49E+10 1.34E+10 2.88E+10 67 23 1.36 27.17 3.17E+10 2.04E+10 1.54E+10 1.39E+10 3.18E+10 67 24 1.61 28.78 2.76E+10 1.81E+10 1.36E+10 1.21E+10 2.78E+10 67 25 1.43 30.21 2.73E+10 1.82E+10 1.48E+10 1.30E+10 2.74E+10 5
26 1.42 31.63 2.83E+10 1.87E+10 1.40E+10 1.24E+10 2.85E+10
-73 27 1.33 32.96 2.86E+10 1.82E+10 1.29E+10 1.17E+10 2.88E+10
-73 28 1.74 34.71 2.79E+10 1.79E+10 1.29E+10 1.18E+10 2.80E+10
-73 29 1.69 36.40 2.93E+10 1.86E+10 1.40E+10 1.22E+10 2.95E+10
-73 30 1.79 38.19 2.73E+10 1.82E+10 1.47E+10 1.37E+10 2.75E+10
-73 31 1.87 40.06 2.94E+10 1.85E+10 1.38E+10 1.22E+10 2.96E+10
-73 32 2.03 42.08 2.80E+10 1.79E+10 1.40E+10 1.30E+10 2.81E+10
-73
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-14 WCAP-18746-NP January 2023 Revision 2 Table 2-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence (n/cm2) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.4 1.4 2.04E+18 1.21E+18 8.04E+17 7.08E+17 2.05E+18
-73 2
0.78 2.18 3.14E+18 1.89E+18 1.28E+18 1.11E+18 3.16E+18
-73 3
0.86 3.03 4.47E+18 2.66E+18 1.78E+18 1.57E+18 4.49E+18
-3 4
0.89 3.92 5.81E+18 3.48E+18 2.35E+18 2.06E+18 5.84E+18
-3 5
0.99 4.91 7.38E+18 4.38E+18 2.96E+18 2.61E+18 7.42E+18
-1 6
0.87 5.79 8.77E+18 5.21E+18 3.52E+18 3.09E+18 8.81E+18
-1 7
1.01 6.8 1.01E+19 6.07E+18 4.18E+18 3.71E+18 1.01E+19
-1 8
0.89 7.69 1.16E+19 6.97E+18 4.80E+18 4.22E+18 1.17E+19 1
9 0.94 8.63 1.30E+19 7.76E+18 5.42E+18 4.73E+18 1.31E+19 1
10 0.92 9.55 1.47E+19 8.74E+18 5.98E+18 5.27E+18 1.48E+19 3
11 0.93 10.48 1.65E+19 9.77E+18 6.68E+18 5.78E+18 1.65E+19 3
12 1.18 11.65 1.81E+19 1.08E+19 7.45E+18 6.38E+18 1.82E+19 3
13 1.24 12.89 1.93E+19 1.15E+19 8.09E+18 6.95E+18 1.94E+19 3
14 1.21 14.11 2.02E+19 1.22E+19 8.60E+18 7.43E+18 2.03E+19 3
15 1.25 15.36 2.11E+19 1.28E+19 9.12E+18 7.93E+18 2.12E+19 3
16 1.29 16.65 2.22E+19 1.36E+19 9.75E+18 8.45E+18 2.23E+19 3
17 1.47 18.12 2.35E+19 1.45E+19 1.04E+19 9.02E+18 2.36E+19 3
18 1.55 19.68 2.47E+19 1.53E+19 1.11E+19 9.59E+18 2.48E+19 3
19 1.21 20.89 2.56E+19 1.60E+19 1.17E+19 1.01E+19 2.57E+19 3
20 1.61 22.5 2.70E+19 1.69E+19 1.24E+19 1.07E+19 2.71E+19 5
21 1.6 24.09 2.84E+19 1.78E+19 1.31E+19 1.13E+19 2.86E+19 5
22 1.72 25.81 3.00E+19 1.88E+19 1.39E+19 1.21E+19 3.01E+19 5
23 1.36 27.17 3.13E+19 1.97E+19 1.45E+19 1.26E+19 3.15E+19 5
24 1.61 28.78 3.27E+19 2.06E+19 1.52E+19 1.33E+19 3.29E+19 5
25 1.43 30.21 3.39E+19 2.14E+19 1.59E+19 1.38E+19 3.41E+19 5
26 1.42 31.63 3.52E+19 2.23E+19 1.65E+19 1.44E+19 3.54E+19 5
27 1.33 32.96 3.64E+19 2.30E+19 1.70E+19 1.49E+19 3.66E+19 5
28 1.74 34.71 3.79E+19 2.40E+19 1.78E+19 1.55E+19 3.81E+19 5
29 1.69 36.4 3.95E+19 2.49E+19 1.85E+19 1.62E+19 3.97E+19 5
30 1.79 38.19 4.10E+19 2.59E+19 1.93E+19 1.69E+19 4.12E+19 5
31 1.87 40.06 4.27E+19 2.70E+19 2.01E+19 1.77E+19 4.29E+19 3
32 2.03 42.08 4.45E+19 2.81E+19 2.10E+19 1.85E+19 4.47E+19 3
48(a) 5.02E+19 3.17E+19 2.39E+19 2.11E+19 5.05E+19 3
51 5.32E+19 3.36E+19 2.53E+19 2.24E+19 5.34E+19 3
54 5.61E+19 3.54E+19 2.67E+19 2.38E+19 5.63E+19 3
60 6.19E+19 3.90E+19 2.96E+19 2.65E+19 6.22E+19 3
Notes:
(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-15 WCAP-18746-NP January 2023 Revision 2 Table 2-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Displacement Rate (dpa/s) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.40 1.40 7.56E-11 4.56E-11 3.00E-11 2.61E-11 7.59E-11
-73 2
0.78 2.18 7.83E-11 4.88E-11 3.30E-11 2.81E-11 7.87E-11 119 3
0.86 3.03 8.14E-11 4.97E-11 3.11E-11 2.77E-11 8.17E-11
-1 4
0.89 3.92 7.89E-11 4.93E-11 3.36E-11 2.85E-11 7.92E-11 3
5 0.99 4.91 8.30E-11 5.07E-11 3.25E-11 2.89E-11 8.33E-11 67 6
0.87 5.79 8.30E-11 5.10E-11 3.35E-11 2.85E-11 8.34E-11
-73 7
1.01 6.80 6.79E-11 4.55E-11 3.42E-11 3.13E-11 6.82E-11 5
8 0.89 7.69 9.14E-11 5.51E-11 3.63E-11 2.95E-11 9.18E-11 5
9 0.94 8.63 7.54E-11 4.51E-11 3.44E-11 2.82E-11 7.57E-11 3
10 0.92 9.55 9.59E-11 5.71E-11 3.23E-11 3.02E-11 9.63E-11 5
11 0.93 10.48 9.83E-11 5.98E-11 3.90E-11 2.83E-11 9.87E-11 5
12 1.18 11.65 7.26E-11 4.48E-11 3.46E-11 2.67E-11 7.29E-11 5
13 1.24 12.89 4.82E-11 3.37E-11 2.65E-11 2.35E-11 4.84E-11 5
14 1.21 14.11 3.92E-11 2.74E-11 2.22E-11 2.06E-11 3.94E-11 73 15 1.25 15.36 3.87E-11 2.75E-11 2.17E-11 2.04E-11 3.88E-11 5
16 1.29 16.65 4.52E-11 3.22E-11 2.53E-11 2.09E-11 4.54E-11 3
17 1.47 18.12 4.50E-11 3.28E-11 2.49E-11 2.00E-11 4.52E-11 5
18 1.55 19.68 4.11E-11 2.84E-11 2.12E-11 1.90E-11 4.12E-11 5
19 1.21 20.89 3.79E-11 2.98E-11 2.56E-11 2.22E-11 3.81E-11 67 20 1.61 22.50 4.51E-11 3.22E-11 2.46E-11 2.04E-11 4.53E-11 73 21 1.60 24.09 4.81E-11 3.04E-11 2.19E-11 1.93E-11 4.84E-11 5
22 1.72 25.81 4.70E-11 3.16E-11 2.45E-11 2.18E-11 4.72E-11 67 23 1.36 27.17 5.18E-11 3.43E-11 2.53E-11 2.26E-11 5.20E-11 67 24 1.61 28.78 4.52E-11 3.04E-11 2.24E-11 1.97E-11 4.54E-11 67 25 1.43 30.21 4.47E-11 3.06E-11 2.44E-11 2.11E-11 4.49E-11 5
26 1.42 31.63 4.64E-11 3.10E-11 2.30E-11 2.01E-11 4.66E-11
-73 27 1.33 32.96 4.68E-11 3.01E-11 2.11E-11 1.90E-11 4.70E-11
-73 28 1.74 34.71 4.56E-11 2.97E-11 2.12E-11 1.91E-11 4.57E-11
-73 29 1.69 36.40 4.79E-11 3.08E-11 2.30E-11 1.99E-11 4.82E-11
-73 30 1.79 38.19 4.47E-11 3.02E-11 2.42E-11 2.22E-11 4.49E-11
-3 31 1.87 40.06 4.81E-11 3.08E-11 2.26E-11 1.98E-11 4.84E-11
-73 32 2.03 42.08 4.58E-11 2.97E-11 2.29E-11 2.10E-11 4.60E-11 61
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-16 WCAP-18746-NP January 2023 Revision 2 Table 2-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 1 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Displacements (dpa) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.40 1.40 3.34E-03 2.01E-03 1.32E-03 1.15E-03 3.35E-03
-73 2
0.78 2.18 5.15E-03 3.14E-03 2.10E-03 1.81E-03 5.17E-03
-73 3
0.86 3.03 7.33E-03 4.49E-03 2.94E-03 2.55E-03 7.36E-03
-3 4
0.89 3.92 9.54E-03 5.87E-03 3.88E-03 3.35E-03 9.58E-03
-3 5
0.99 4.91 1.21E-02 7.44E-03 4.89E-03 4.25E-03 1.22E-02
-1 6
0.87 5.79 1.44E-02 8.85E-03 5.81E-03 5.04E-03 1.45E-02
-1 7
1.01 6.80 1.66E-02 1.03E-02 6.90E-03 6.04E-03 1.66E-02
-1 8
0.89 7.69 1.91E-02 1.18E-02 7.92E-03 6.87E-03 1.92E-02
-1 9
0.94 8.63 2.14E-02 1.32E-02 8.94E-03 7.70E-03 2.14E-02
-1 10 0.92 9.55 2.41E-02 1.48E-02 9.88E-03 8.58E-03 2.42E-02 3
11 0.93 10.48 2.70E-02 1.66E-02 1.10E-02 9.41E-03 2.71E-02 3
12 1.18 11.65 2.97E-02 1.83E-02 1.23E-02 1.04E-02 2.98E-02 3
13 1.24 12.89 3.16E-02 1.96E-02 1.33E-02 1.13E-02 3.17E-02 3
14 1.21 14.11 3.31E-02 2.06E-02 1.42E-02 1.21E-02 3.32E-02 3
15 1.25 15.36 3.46E-02 2.17E-02 1.51E-02 1.29E-02 3.48E-02 3
16 1.29 16.65 3.65E-02 2.30E-02 1.61E-02 1.38E-02 3.66E-02 3
17 1.47 18.12 3.85E-02 2.45E-02 1.72E-02 1.47E-02 3.87E-02 5
18 1.55 19.68 4.06E-02 2.59E-02 1.83E-02 1.56E-02 4.07E-02 5
19 1.21 20.89 4.20E-02 2.71E-02 1.93E-02 1.65E-02 4.22E-02 5
20 1.61 22.50 4.42E-02 2.87E-02 2.05E-02 1.75E-02 4.44E-02 5
21 1.60 24.09 4.66E-02 3.02E-02 2.16E-02 1.85E-02 4.68E-02 5
22 1.72 25.81 4.92E-02 3.19E-02 2.29E-02 1.96E-02 4.94E-02 5
23 1.36 27.17 5.13E-02 3.33E-02 2.40E-02 2.06E-02 5.16E-02 5
24 1.61 28.78 5.36E-02 3.49E-02 2.51E-02 2.16E-02 5.38E-02 5
25 1.43 30.21 5.56E-02 3.63E-02 2.62E-02 2.25E-02 5.59E-02 5
26 1.42 31.63 5.77E-02 3.76E-02 2.72E-02 2.34E-02 5.79E-02 5
27 1.33 32.96 5.96E-02 3.89E-02 2.81E-02 2.42E-02 5.99E-02 5
28 1.74 34.71 6.21E-02 4.05E-02 2.93E-02 2.53E-02 6.24E-02 5
29 1.69 36.40 6.47E-02 4.22E-02 3.05E-02 2.63E-02 6.50E-02 5
30 1.79 38.19 6.72E-02 4.39E-02 3.18E-02 2.76E-02 6.75E-02 5
31 1.87 40.06 7.00E-02 4.57E-02 3.32E-02 2.87E-02 7.03E-02 5
32 2.03 42.08 7.30E-02 4.76E-02 3.46E-02 3.01E-02 7.33E-02 5
48(a) 8.23E-02 5.37E-02 3.93E-02 3.43E-02 8.27E-02 3
51 8.71E-02 5.67E-02 4.17E-02 3.65E-02 8.74E-02 3
54 9.18E-02 5.98E-02 4.41E-02 3.87E-02 9.22E-02 3
60 1.01E-01 6.60E-02 4.88E-02 4.30E-02 1.02E-01 3
Notes:
(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-17 WCAP-18746-NP January 2023 Revision 2 Table 2-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 1 Material Fast Fluence (n/cm2) 42.1 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 2.63E+19 3.00E+19 3.37E+19 3.74E+19 Intermediate Shell Forging 4.47E+19 5.05E+19 5.63E+19 6.22E+19 Lower Shell Forging 4.37E+19 4.95E+19 5.53E+19 6.12E+19 Inlet Nozzle to Nozzle Shell Weld
- Lowest Extent(a) 2.56E+16 2.96E+16 3.36E+16 3.76E+16 Upper to Intermediate Shell Weld 2.84E+19 3.23E+19 3.63E+19 4.02E+19 Intermediate to Lower Shell Weld 4.38E+19 4.95E+19 5.53E+19 6.12E+19 Lower Shell to Lower Closure Head Weld 1.49E+16 1.70E+16 1.92E+16 2.14E+16 Notes:
(a) The outlet nozzle weld fluence is bounded by the inlet nozzle location fluence.
(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-18 WCAP-18746-NP January 2023 Revision 2 Table 2-9 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Material for Unit 1 Material Iron Atom Displacements (dpa) 42.1 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 4.33E-02 4.93E-02 5.54E-02 6.15E-02 Intermediate Shell Forging 7.33E-02 8.27E-02 9.22E-02 1.02E-01 Lower Shell Forging 7.16E-02 8.10E-02 9.05E-02 1.00E-01 Inlet Nozzle to Nozzle Shell Weld -
Lowest Extent(a) 1.25E-04 1.43E-04 1.62E-04 1.80E-04 Upper to Intermediate Shell Weld 4.67E-02 5.31E-02 5.96E-02 6.61E-02 Intermediate to Lower Shell Weld 7.17E-02 8.10E-02 9.06E-02 1.00E-01 Lower Shell to Lower Closure Head Weld 7.60E-05 8.70E-05 9.81E-05 1.09E-04 Notes:
(a) The outlet nozzle weld dpa is bounded by the inlet nozzle location dpa.
(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-19 WCAP-18746-NP January 2023 Revision 2 Table 2-10 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence Rate (n/cm2-s) 13° 23° 33° 1
1.39 1.39 1.36E+11 7.79E+10 7.48E+10 2
0.87 2.26 1.45E+11 8.68E+10 8.40E+10 3
0.89 3.15 1.50E+11 8.91E+10 8.71E+10 4
0.98 4.13 1.51E+11 8.84E+10 8.46E+10 5
0.92 5.05 1.58E+11 9.03E+10 8.74E+10 6
0.99 6.04 1.49E+11 7.93E+10 7.12E+10 7
1.01 7.05 1.43E+11 8.20E+10 7.62E+10 8
0.90 7.95 1.25E+11 7.61E+10 7.06E+10 9
0.85 8.80 1.82E+11 9.43E+10 8.63E+10 10 0.92 9.71 1.58E+11 9.42E+10 9.12E+10 11 1.09 10.80 1.48E+11 8.63E+10 8.25E+10 12 1.08 11.88 1.26E+11 8.51E+10 8.57E+10 13 1.26 13.15 9.40E+10 6.46E+10 6.25E+10 14 1.33 14.47 8.76E+10 6.11E+10 6.11E+10 15 1.38 15.86 8.99E+10 5.67E+10 5.52E+10 16 1.38 17.24 9.43E+10 7.13E+10 6.69E+10 17 1.55 18.78 8.79E+10 6.56E+10 5.87E+10 18 1.48 20.27 8.37E+10 5.78E+10 5.47E+10 19 1.30 21.57 8.10E+10 6.50E+10 6.10E+10 20 1.56 23.13 9.53E+10 6.84E+10 6.04E+10 21 1.52 24.64 8.87E+10 6.30E+10 5.54E+10 22 1.48 26.12 8.86E+10 6.17E+10 5.81E+10 23 1.37 27.49 8.95E+10 6.00E+10 5.77E+10 24 1.72 29.21 8.69E+10 5.77E+10 5.48E+10 25 1.44 30.66 8.60E+10 5.94E+10 5.99E+10 26 1.68 32.34 8.87E+10 5.82E+10 5.74E+10 27 1.24 33.58 9.32E+10 6.10E+10 6.06E+10 28 1.65 35.24 9.35E+10 6.04E+10 5.70E+10 29 1.64 36.88 9.36E+10 6.14E+10 6.10E+10 30 1.85 38.73 8.86E+10 6.10E+10 6.21E+10 31 1.91 40.64 9.20E+10 5.91E+10 5.77E+10 32 1.95 42.59 9.15E+10 6.04E+10 6.01E+10
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-20 WCAP-18746-NP January 2023 Revision 2 Table 2-10 (continued) Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Location for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence (n/cm2)
V (13°)
T (23°)
R (13°)
P (23°)
N (33°)
S (33°)
1 1.39 1.39 5.98E+18 3.42E+18 5.98E+18 3.42E+18 3.29E+18 3.29E+18 2
0.87 2.26 5.81E+18 9.97E+18 5.81E+18 5.59E+18 5.59E+18 3
0.89 3.15 8.30E+18 1.42E+19 8.30E+18 8.03E+18 8.03E+18 4
0.98 4.13 1.10E+19 1.88E+19 1.10E+19 1.06E+19 1.06E+19 5
0.92 5.05 2.34E+19 1.37E+19 1.32E+19 1.32E+19 6
0.99 6.04 2.81E+19 1.61E+19 1.54E+19 1.54E+19 7
1.01 7.05 3.26E+19 1.88E+19 1.78E+19 1.78E+19 8
0.90 7.95 3.62E+19 2.09E+19 1.98E+19 1.98E+19 9
0.85 8.80 4.11E+19 2.34E+19 2.22E+19 2.22E+19 10 0.92 9.71 2.62E+19 2.48E+19 2.48E+19 11 1.09 10.80 2.91E+19 2.76E+19 2.76E+19 12 1.08 11.88 3.20E+19 3.05E+19 3.05E+19 13 1.26 13.15 3.46E+19 3.30E+19 3.30E+19 14 1.33 14.47 3.72E+19 3.56E+19 3.56E+19 15 1.38 15.86 3.96E+19 3.80E+19 3.80E+19 16 1.38 17.24 4.27E+19 4.09E+19 4.09E+19 17 1.55 18.78 4.38E+19 4.38E+19 18 1.48 20.27 4.64E+19 4.64E+19 19 1.30 21.57 4.89E+19 4.89E+19 20 1.56 23.13 5.18E+19 5.18E+19 21 1.52 24.64 5.45E+19 5.45E+19 22 1.48 26.12 5.72E+19 5.72E+19 23 1.37 27.49 5.97E+19 5.97E+19 24 1.72 29.21 6.27E+19 6.27E+19 25 1.44 30.66 6.54E+19 6.54E+19 26 1.68 32.34 6.84E+19 6.84E+19 27 1.24 33.58 7.08E+19 7.08E+19 28 1.65 35.24 7.38E+19 7.38E+19 29 1.64 36.88 7.69E+19 7.69E+19 30 1.85 38.73 8.06E+19 8.06E+19 31 1.91 40.64 8.41E+19 8.41E+19 32 1.95 42.59 8.77E+19 48 9.90E+19 51 1.05E+20 54 1.11E+20 60 1.24E+20
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-21 WCAP-18746-NP January 2023 Revision 2 Table 2-11 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Iron Atom Displacement Rate (dpa/s) 13° 23° 33° 1
1.39 1.39 2.49E-10 1.37E-10 1.32E-10 2
0.87 2.26 2.66E-10 1.52E-10 1.48E-10 3
0.89 3.15 2.75E-10 1.56E-10 1.54E-10 4
0.98 4.13 2.77E-10 1.55E-10 1.49E-10 5
0.92 5.05 2.90E-10 1.58E-10 1.54E-10 6
0.99 6.04 2.73E-10 1.39E-10 1.25E-10 7
1.01 7.05 2.62E-10 1.44E-10 1.34E-10 8
0.90 7.95 2.29E-10 1.33E-10 1.24E-10 9
0.85 8.80 3.34E-10 1.66E-10 1.52E-10 10 0.92 9.71 2.89E-10 1.65E-10 1.61E-10 11 1.09 10.80 2.70E-10 1.51E-10 1.45E-10 12 1.08 11.88 2.30E-10 1.49E-10 1.51E-10 13 1.26 13.15 1.71E-10 1.13E-10 1.10E-10 14 1.33 14.47 1.59E-10 1.06E-10 1.07E-10 15 1.38 15.86 1.63E-10 9.88E-11 9.69E-11 16 1.38 17.24 1.71E-10 1.24E-10 1.17E-10 17 1.55 18.78 1.59E-10 1.14E-10 1.03E-10 18 1.48 20.27 1.52E-10 1.00E-10 9.59E-11 19 1.30 21.57 1.47E-10 1.13E-10 1.07E-10 20 1.56 23.13 1.73E-10 1.19E-10 1.06E-10 21 1.52 24.64 1.61E-10 1.09E-10 9.72E-11 22 1.48 26.12 1.61E-10 1.07E-10 1.02E-10 23 1.37 27.49 1.62E-10 1.04E-10 1.01E-10 24 1.72 29.21 1.58E-10 1.00E-10 9.61E-11 25 1.44 30.66 1.56E-10 1.03E-10 1.05E-10 26 1.68 32.34 1.61E-10 1.01E-10 1.01E-10 27 1.24 33.58 1.69E-10 1.06E-10 1.06E-10 28 1.65 35.24 1.70E-10 1.05E-10 9.99E-11 29 1.64 36.88 1.70E-10 1.07E-10 1.07E-10 30 1.85 38.73 1.61E-10 1.06E-10 1.09E-10 31 1.91 40.64 1.67E-10 1.03E-10 1.01E-10 32 1.95 42.59 1.66E-10 1.05E-10 1.05E-10
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-22 WCAP-18746-NP January 2023 Revision 2 Table 2-11 (continued) Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Displacements (dpa)
V (13°)
T (23°)
R (13°)
P (23°)
N (33°)
S (33°)
1 1.39 1.39 1.10E-02 6.00E-03 1.10E-02 6.00E-03 5.79E-03 5.79E-03 2
0.87 2.26 1.02E-02 1.83E-02 1.02E-02 9.86E-03 9.86E-03 3
0.89 3.15 1.46E-02 2.60E-02 1.46E-02 1.42E-02 1.42E-02 4
0.98 4.13 1.93E-02 3.45E-02 1.93E-02 1.88E-02 1.88E-02 5
0.92 5.05 4.29E-02 2.39E-02 2.32E-02 2.32E-02 6
0.99 6.04 5.15E-02 2.83E-02 2.72E-02 2.72E-02 7
1.01 7.05 5.98E-02 3.29E-02 3.14E-02 3.14E-02 8
0.90 7.95 6.63E-02 3.66E-02 3.50E-02 3.50E-02 9
0.85 8.80 7.53E-02 4.11E-02 3.90E-02 3.90E-02 10 0.92 9.71 4.58E-02 4.37E-02 4.37E-02 11 1.09 10.80 5.10E-02 4.87E-02 4.87E-02 12 1.08 11.88 5.61E-02 5.38E-02 5.38E-02 13 1.26 13.15 6.06E-02 5.82E-02 5.82E-02 14 1.33 14.47 6.50E-02 6.27E-02 6.27E-02 15 1.38 15.86 6.93E-02 6.69E-02 6.69E-02 16 1.38 17.24 7.47E-02 7.20E-02 7.20E-02 17 1.55 18.78 7.71E-02 7.71E-02 18 1.48 20.27 8.16E-02 8.16E-02 19 1.30 21.57 8.60E-02 8.60E-02 20 1.56 23.13 9.12E-02 9.12E-02 21 1.52 24.64 9.58E-02 9.58E-02 22 1.48 26.12 1.01E-01 1.01E-01 23 1.37 27.49 1.05E-01 1.05E-01 24 1.72 29.21 1.10E-01 1.10E-01 25 1.44 30.66 1.15E-01 1.15E-01 26 1.68 32.34 1.20E-01 1.20E-01 27 1.24 33.58 1.24E-01 1.24E-01 28 1.65 35.24 1.30E-01 1.30E-01 29 1.64 36.88 1.35E-01 1.35E-01 30 1.85 38.73 1.42E-01 1.42E-01 31 1.91 40.64 1.48E-01 1.48E-01 32 1.95 42.59 1.54E-01 48 1.74E-01 51 1.85E-01 54 1.96E-01 60 2.18E-01
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-23 WCAP-18746-NP January 2023 Revision 2 Table 2-12 Calculated Surveillance Capsule Lead Factors for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Lead Factor 13° 23° 33° 1
1.39 1.39 3.03(a) 1.73 1.66 2
0.87 2.26 3.06 1.78 1.72 3
0.89 3.15 3.08 1.80 1.74 4
0.98 4.13 3.08 1.80(b) 1.74 5
0.92 5.05 3.07 1.79 1.73 6
0.99 6.04 3.07 1.76 1.68 7
1.01 7.05 3.11 1.78 1.70 8
0.90 7.95 3.09 1.78 1.69 9
0.85 8.80 3.08(c) 1.76 1.66 10 0.92 9.71 3.08 1.77 1.68 11 1.09 10.80 3.07 1.76 1.67 12 1.08 11.88 3.08 1.80 1.71 13 1.26 13.15 3.09 1.82 1.74 14 1.33 14.47 3.09 1.84 1.76 15 1.38 15.86 3.09 1.85 1.77 16 1.38 17.24 3.11 1.88(d) 1.80 17 1.55 18.78 3.12 1.92 1.83 18 1.48 20.27 3.13 1.93 1.84 19 1.30 21.57 3.14 1.97 1.87 20 1.56 23.13 3.15 1.99 1.89 21 1.52 24.64 3.16 2.01 1.89 22 1.48 26.12 3.16 2.02 1.90 23 1.37 27.49 3.16 2.02 1.91 24 1.72 29.21 3.16 2.02 1.91 25 1.44 30.66 3.16 2.02 1.92 26 1.68 32.34 3.15 2.02 1.92 27 1.24 33.58 3.15 2.02 1.92 28 1.65 35.24 3.15 2.02 1.92 29 1.64 36.88 3.15 2.03 1.93 30 1.85 38.73 3.16 2.04 1.94 31 1.91 40.64 3.15 2.03 1.94(e) 32 1.95 42.59 3.15 2.04 1.95 48(f) 3.15 2.04 1.96 51 3.15 2.04 1.97 54 3.15 2.04 1.97 60 3.15 2.04 1.98 Notes:
(a) Capsule V was removed after Cycle 1.
(b) Capsule T was removed after Cycle 4.
(c) Capsule R was removed after Cycle 9.
(d) Capsule P was removed after Cycle 16.
(e) Capsule N was removed after Cycle 31.
(f) The projections beyond Cycle 32 are based on Cycle 32 with a 10%
bias on peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-24 WCAP-18746-NP January 2023 Revision 2 Table 2-13 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence Rate (n/cm2-s) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.39 1.39 4.49E+10 2.69E+10 1.77E+10 1.55E+10 4.49E+10
-9 2
0.87 2.26 4.75E+10 2.93E+10 2.01E+10 1.73E+10 4.75E+10 73 3
0.89 3.15 4.83E+10 2.97E+10 2.05E+10 1.80E+10 4.83E+10
-1 4
0.98 4.13 4.90E+10 3.00E+10 2.01E+10 1.72E+10 4.90E+10 5
5 0.92 5.05 5.19E+10 3.11E+10 2.06E+10 1.84E+10 5.19E+10
-3 6
0.99 6.04 5.03E+10 2.98E+10 1.74E+10 1.62E+10 5.03E+10 59 7
1.01 7.05 4.24E+10 2.84E+10 1.82E+10 1.75E+10 4.24E+10
-3 8
0.90 7.95 4.28E+10 2.53E+10 1.70E+10 1.47E+10 4.28E+10
-3 9
0.85 8.80 5.98E+10 3.53E+10 2.06E+10 1.92E+10 5.98E+10
-3 10 0.92 9.71 5.07E+10 3.13E+10 2.15E+10 1.85E+10 5.07E+10 5
11 1.09 10.80 5.06E+10 2.96E+10 1.95E+10 1.89E+10 5.06E+10 5
12 1.08 11.88 3.87E+10 2.60E+10 2.01E+10 1.89E+10 3.87E+10 5
13 1.26 13.15 2.96E+10 1.97E+10 1.50E+10 1.33E+10 2.96E+10 5
14 1.33 14.47 2.79E+10 1.84E+10 1.45E+10 1.29E+10 2.79E+10 73 15 1.38 15.86 2.95E+10 1.85E+10 1.32E+10 1.26E+10 2.95E+10 5
16 1.38 17.24 2.78E+10 2.02E+10 1.62E+10 1.41E+10 2.78E+10 5
17 1.55 18.78 2.61E+10 1.89E+10 1.45E+10 1.21E+10 2.61E+10 5
18 1.48 20.27 2.56E+10 1.76E+10 1.32E+10 1.19E+10 2.56E+10 5
19 1.30 21.57 2.27E+10 1.76E+10 1.48E+10 1.26E+10 2.27E+10 5
20 1.56 23.13 2.87E+10 2.03E+10 1.50E+10 1.19E+10 2.87E+10 67 21 1.52 24.64 2.69E+10 1.89E+10 1.38E+10 1.16E+10 2.69E+10 67 22 1.48 26.12 2.75E+10 1.86E+10 1.41E+10 1.24E+10 2.76E+10 67 23 1.37 27.49 2.87E+10 1.86E+10 1.38E+10 1.26E+10 2.87E+10 67 24 1.72 29.21 2.86E+10 1.80E+10 1.31E+10 1.14E+10 2.86E+10 3
25 1.44 30.66 2.76E+10 1.79E+10 1.41E+10 1.28E+10 2.76E+10
-73 26 1.68 32.34 2.93E+10 1.83E+10 1.36E+10 1.28E+10 2.93E+10
-73 27 1.24 33.58 3.02E+10 1.92E+10 1.44E+10 1.32E+10 3.02E+10
-73 28 1.65 35.24 2.97E+10 1.91E+10 1.36E+10 1.22E+10 2.97E+10
-3 29 1.64 36.88 2.88E+10 1.91E+10 1.44E+10 1.31E+10 2.88E+10
-3 30 1.85 38.73 2.76E+10 1.83E+10 1.46E+10 1.34E+10 2.76E+10
-73 31 1.91 40.64 2.96E+10 1.88E+10 1.37E+10 1.21E+10 2.96E+10
-73 32 1.95 42.59 2.93E+10 1.88E+10 1.42E+10 1.29E+10 2.93E+10
-3
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-25 WCAP-18746-NP January 2023 Revision 2 Table 2-14 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Fluence (n/cm2) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.39 1.39 1.98E+18 1.18E+18 7.78E+17 6.81E+17 1.98E+18
-9 2
0.87 2.26 3.25E+18 1.97E+18 1.32E+18 1.15E+18 3.25E+18
-3 3
0.89 3.15 4.60E+18 2.81E+18 1.90E+18 1.65E+18 4.60E+18
-3 4
0.98 4.13 6.12E+18 3.73E+18 2.52E+18 2.18E+18 6.12E+18
-1 5
0.92 5.05 7.62E+18 4.63E+18 3.11E+18 2.72E+18 7.62E+18
-3 6
0.99 6.04 9.15E+18 5.55E+18 3.65E+18 3.22E+18 9.15E+18
-1 7
1.01 7.05 1.05E+19 6.45E+18 4.23E+18 3.78E+18 1.05E+19
-1 8
0.9 7.95 1.17E+19 7.17E+18 4.71E+18 4.19E+18 1.17E+19
-1 9
0.85 8.8 1.33E+19 8.12E+18 5.26E+18 4.71E+18 1.33E+19
-1 10 0.92 9.71 1.48E+19 9.02E+18 5.89E+18 5.24E+18 1.48E+19
-1 11 1.09 10.8 1.65E+19 1.00E+19 6.56E+18 5.89E+18 1.65E+19
-1 12 1.08 11.88 1.78E+19 1.09E+19 7.24E+18 6.54E+18 1.78E+19 1
13 1.26 13.15 1.90E+19 1.17E+19 7.84E+18 7.07E+18 1.90E+19 3
14 1.33 14.47 2.02E+19 1.25E+19 8.45E+18 7.61E+18 2.02E+19 3
15 1.38 15.86 2.15E+19 1.33E+19 9.02E+18 8.16E+18 2.15E+19 3
16 1.38 17.24 2.27E+19 1.42E+19 9.73E+18 8.77E+18 2.27E+19 3
17 1.55 18.78 2.39E+19 1.51E+19 1.04E+19 9.36E+18 2.39E+19 3
18 1.48 20.27 2.51E+19 1.59E+19 1.11E+19 9.92E+18 2.51E+19 3
19 1.3 21.57 2.61E+19 1.66E+19 1.17E+19 1.04E+19 2.61E+19 3
20 1.56 23.13 2.75E+19 1.76E+19 1.24E+19 1.10E+19 2.75E+19 5
21 1.52 24.64 2.88E+19 1.85E+19 1.31E+19 1.16E+19 2.88E+19 5
22 1.48 26.12 3.00E+19 1.94E+19 1.37E+19 1.22E+19 3.00E+19 5
23 1.37 27.49 3.13E+19 2.02E+19 1.43E+19 1.27E+19 3.13E+19 5
24 1.72 29.21 3.28E+19 2.12E+19 1.50E+19 1.33E+19 3.28E+19 5
25 1.44 30.66 3.41E+19 2.20E+19 1.57E+19 1.39E+19 3.41E+19 5
26 1.68 32.34 3.56E+19 2.29E+19 1.64E+19 1.46E+19 3.56E+19 5
27 1.24 33.58 3.68E+19 2.37E+19 1.69E+19 1.51E+19 3.68E+19 5
28 1.65 35.24 3.83E+19 2.47E+19 1.77E+19 1.57E+19 3.83E+19 3
29 1.64 36.88 3.98E+19 2.57E+19 1.84E+19 1.64E+19 3.98E+19 3
30 1.85 38.73 4.15E+19 2.68E+19 1.92E+19 1.72E+19 4.15E+19 3
31 1.91 40.64 4.32E+19 2.79E+19 2.01E+19 1.79E+19 4.32E+19 3
32 1.95 42.59 4.50E+19 2.90E+19 2.09E+19 1.87E+19 4.50E+19 3
48(a) 5.05E+19 3.25E+19 2.36E+19 2.11E+19 5.05E+19 3
51 5.35E+19 3.45E+19 2.50E+19 2.24E+19 5.35E+19 3
54 5.66E+19 3.64E+19 2.65E+19 2.38E+19 5.66E+19 3
60 6.26E+19 4.03E+19 2.94E+19 2.64E+19 6.26E+19 3
Notes:
(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-26 WCAP-18746-NP January 2023 Revision 2 Table 2-15 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Displacement Rate (dpa/s) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.39 1.39 7.37E-11 4.44E-11 2.89E-11 2.52E-11 7.37E-11
-9 2
0.87 2.26 7.79E-11 4.84E-11 3.28E-11 2.82E-11 7.79E-11 73 3
0.89 3.15 7.92E-11 4.91E-11 3.35E-11 2.93E-11 7.92E-11
-1 4
0.98 4.13 8.04E-11 4.95E-11 3.27E-11 2.80E-11 8.04E-11 5
5 0.92 5.05 8.51E-11 5.14E-11 3.36E-11 3.00E-11 8.51E-11
-3 6
0.99 6.04 8.25E-11 4.93E-11 2.84E-11 2.64E-11 8.25E-11 57 7
1.01 7.05 6.95E-11 4.68E-11 2.97E-11 2.84E-11 6.95E-11
-3 8
0.90 7.95 7.01E-11 4.18E-11 2.77E-11 2.39E-11 7.01E-11
-3 9
0.85 8.80 9.82E-11 5.83E-11 3.37E-11 3.13E-11 9.82E-11
-3 10 0.92 9.71 8.32E-11 5.17E-11 3.51E-11 3.02E-11 8.32E-11 5
11 1.09 10.80 8.28E-11 4.88E-11 3.19E-11 3.07E-11 8.28E-11 5
12 1.08 11.88 6.34E-11 4.28E-11 3.28E-11 3.08E-11 6.34E-11 9
13 1.26 13.15 4.85E-11 3.23E-11 2.44E-11 2.16E-11 4.85E-11 5
14 1.33 14.47 4.56E-11 3.02E-11 2.36E-11 2.10E-11 4.56E-11 73 15 1.38 15.86 4.83E-11 3.05E-11 2.14E-11 2.05E-11 4.83E-11 9
16 1.38 17.24 4.55E-11 3.32E-11 2.64E-11 2.29E-11 4.55E-11 5
17 1.55 18.78 4.27E-11 3.10E-11 2.36E-11 1.97E-11 4.27E-11 5
18 1.48 20.27 4.18E-11 2.89E-11 2.15E-11 1.93E-11 4.18E-11 5
19 1.30 21.57 3.72E-11 2.89E-11 2.41E-11 2.05E-11 3.72E-11 5
20 1.56 23.13 4.70E-11 3.33E-11 2.44E-11 1.94E-11 4.70E-11 65 21 1.52 24.64 4.40E-11 3.10E-11 2.24E-11 1.89E-11 4.40E-11 65 22 1.48 26.12 4.50E-11 3.06E-11 2.29E-11 2.02E-11 4.52E-11 67 23 1.37 27.49 4.69E-11 3.05E-11 2.24E-11 2.05E-11 4.69E-11 67 24 1.72 29.21 4.67E-11 2.95E-11 2.13E-11 1.86E-11 4.67E-11 3
25 1.44 30.66 4.51E-11 2.94E-11 2.30E-11 2.08E-11 4.51E-11
-73 26 1.68 32.34 4.78E-11 3.01E-11 2.22E-11 2.07E-11 4.78E-11
-73 27 1.24 33.58 4.93E-11 3.16E-11 2.34E-11 2.15E-11 4.93E-11
-71 28 1.65 35.24 4.86E-11 3.14E-11 2.22E-11 1.98E-11 4.86E-11 61 29 1.64 36.88 4.71E-11 3.14E-11 2.34E-11 2.13E-11 4.71E-11
-3 30 1.85 38.73 4.51E-11 3.01E-11 2.37E-11 2.17E-11 4.51E-11 65 31 1.91 40.64 4.84E-11 3.09E-11 2.23E-11 1.97E-11 4.84E-11
-73 32 1.95 42.59 4.80E-11 3.09E-11 2.30E-11 2.09E-11 4.80E-11 63
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-27 WCAP-18746-NP January 2023 Revision 2 Table 2-16 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface for Unit 2 Cycle Cycle Length (EFPY)
Total Time (EFPY)
Displacements (dpa) 0° 15° 30° 45° Maximum Elevation of Max.
(cm) 1 1.39 1.39 3.24E-03 1.95E-03 1.27E-03 1.11E-03 3.24E-03
-9 2
0.87 2.26 5.33E-03 3.26E-03 2.16E-03 1.87E-03 5.33E-03
-3 3
0.89 3.15 7.55E-03 4.63E-03 3.09E-03 2.69E-03 7.55E-03
-3 4
0.98 4.13 1.00E-02 6.16E-03 4.11E-03 3.56E-03 1.00E-02
-1 5
0.92 5.05 1.25E-02 7.65E-03 5.08E-03 4.43E-03 1.25E-02
-3 6
0.99 6.04 1.50E-02 9.16E-03 5.95E-03 5.24E-03 1.50E-02
-1 7
1.01 7.05 1.72E-02 1.07E-02 6.90E-03 6.15E-03 1.72E-02
-1 8
0.90 7.95 1.92E-02 1.18E-02 7.69E-03 6.83E-03 1.92E-02
-1 9
0.85 8.80 2.19E-02 1.34E-02 8.59E-03 7.67E-03 2.19E-02
-1 10 0.92 9.71 2.43E-02 1.49E-02 9.60E-03 8.54E-03 2.43E-02
-1 11 1.09 10.80 2.71E-02 1.66E-02 1.07E-02 9.59E-03 2.71E-02
-1 12 1.08 11.88 2.92E-02 1.80E-02 1.18E-02 1.06E-02 2.92E-02
-1 13 1.26 13.15 3.12E-02 1.93E-02 1.28E-02 1.15E-02 3.12E-02 3
14 1.33 14.47 3.31E-02 2.06E-02 1.38E-02 1.24E-02 3.31E-02 3
15 1.38 15.86 3.52E-02 2.19E-02 1.47E-02 1.33E-02 3.52E-02 3
16 1.38 17.24 3.72E-02 2.33E-02 1.59E-02 1.43E-02 3.72E-02 3
17 1.55 18.78 3.93E-02 2.49E-02 1.70E-02 1.52E-02 3.93E-02 3
18 1.48 20.27 4.12E-02 2.62E-02 1.80E-02 1.61E-02 4.12E-02 5
19 1.30 21.57 4.27E-02 2.74E-02 1.90E-02 1.70E-02 4.27E-02 5
20 1.56 23.13 4.50E-02 2.90E-02 2.02E-02 1.79E-02 4.50E-02 5
21 1.52 24.64 4.71E-02 3.05E-02 2.13E-02 1.88E-02 4.71E-02 5
22 1.48 26.12 4.92E-02 3.19E-02 2.24E-02 1.98E-02 4.92E-02 5
23 1.37 27.49 5.13E-02 3.33E-02 2.33E-02 2.07E-02 5.13E-02 5
24 1.72 29.21 5.38E-02 3.49E-02 2.45E-02 2.17E-02 5.38E-02 5
25 1.44 30.66 5.58E-02 3.62E-02 2.55E-02 2.26E-02 5.58E-02 5
26 1.68 32.34 5.84E-02 3.78E-02 2.67E-02 2.37E-02 5.84E-02 5
27 1.24 33.58 6.03E-02 3.90E-02 2.76E-02 2.45E-02 6.03E-02 5
28 1.65 35.24 6.28E-02 4.06E-02 2.88E-02 2.56E-02 6.28E-02 5
29 1.64 36.88 6.52E-02 4.23E-02 3.00E-02 2.67E-02 6.52E-02 5
30 1.85 38.73 6.79E-02 4.40E-02 3.13E-02 2.79E-02 6.79E-02 3
31 1.91 40.64 7.08E-02 4.59E-02 3.27E-02 2.91E-02 7.08E-02 3
32 1.95 42.59 7.37E-02 4.78E-02 3.41E-02 3.04E-02 7.37E-02 3
48(a) 8.27E-02 5.35E-02 3.84E-02 3.43E-02 8.27E-02 3
51 8.76E-02 5.67E-02 4.08E-02 3.65E-02 8.76E-02 3
54 9.26E-02 5.99E-02 4.32E-02 3.86E-02 9.26E-02 3
60 1.02E-01 6.63E-02 4.79E-02 4.30E-02 1.02E-01
-1 Notes:
(a) Values beyond Cycle 32 are projected based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-28 WCAP-18746-NP January 2023 Revision 2 Table 2-17 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Beltline and Extended Beltline Materials and Shells for Unit 2 Material Fast Fluence (n/cm2) 42.6 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 2.63E+19 2.98E+19 3.37E+19 3.76E+19 Intermediate Shell Forging 4.50E+19 5.05E+19 5.66E+19 6.26E+19 Lower Shell Forging 4.43E+19 4.98E+19 5.58E+19 6.19E+19 Inlet Nozzle to Nozzle Shell Weld
- Lowest Extent(a) 2.64E+16 3.01E+16 3.42E+16 3.82E+16 Upper to Intermediate Shell Weld 2.82E+19 3.20E+19 3.61E+19 4.03E+19 Intermediate to Lower Shell Weld 4.43E+19 4.98E+19 5.58E+19 6.19E+19 Lower Shell to Lower Closure Head Weld 1.52E+16 1.73E+16 1.95E+16 2.18E+16 Notes:
(a) The outlet nozzle weld fluence is bounded by the inlet nozzle location fluence.
(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-29 WCAP-18746-NP January 2023 Revision 2 Table 2-18 Calculated Maximum Iron Atom Displacements for Pressure Vessel Beltline and Extended Beltline Materials for Unit 2 Material Iron Atom Displacements (dpa) 42.6 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 4.32E-02 4.90E-02 5.54E-02 6.17E-02 Intermediate Shell Forging 7.37E-02 8.27E-02 9.26E-02 1.03E-01 Lower Shell Forging 7.25E-02 8.14E-02 9.13E-02 1.01E-01 Inlet Nozzle to Nozzle Shell Weld -
Lowest Extent(a) 1.26E-04 1.43E-04 1.62E-04 1.80E-04 Upper to Intermediate Shell Weld 4.65E-02 5.26E-02 5.94E-02 6.62E-02 Intermediate to Lower Shell Weld 7.25E-02 8.14E-02 9.13E-02 1.01E-01 Lower Shell to Lower Closure Head Weld 7.78E-05 8.82E-05 9.97E-05 1.11E-04 Notes:
(a) Outlet nozzle weld dpa is bounded by the inlet nozzle location dpa.
(b) Projections are based on Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-30 WCAP-18746-NP January 2023 Revision 2 Figure 2-1 Arrangement of Surveillance Capsules in the Prairie Island Units 1 and 2 Reactor Vessels CAPSULE (TYP) 1so*
270° go*
REACTOR VESSEL
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-31 WCAP-18746-NP January 2023 Revision 2 Figure 2-2 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Core Midplane 250 200 100 50 50 100 150 X-Axis 200
-Air l-Retlective Insulation ioshield Liner Plate 250
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-32 WCAP-18746-NP January 2023 Revision 2 Figure 2-3 Prairie Island Units 1 and 2 Plan View of the Reactor Geometry at the Nozzle Centerline 250 200 100 50 50 100 150 X-Axis 200 Core Barrel - SS Downcomer Water RPV Clad-SS RPV - CS Reflective Insulation Bi oshield Liner Plate Bioshield Concrete Outlet Plenum (Reg N Outlet Plenum (Reg B) 250
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 2-33 WCAP-18746-NP January 2023 Revision 2 Figure 2-4 Prairie Island Units 1 and 2 Section View of the Reactor Geometry at =33° Ill
- --t
~
s:c I
~
400 300 200 100 0
-100
-200
-300
-400 50 100 150 200 250 R-Axis Core - Zone l Core -Zone 2 Core -Zone 3 Core -Zone 4 Core -Zone 5
- Core - Zone 6 Core Baffle - SS Bypass Water Formers - SS Core Barrel - SS
-Thermal Shield - SS Downcomer Water Pressure Vessel Clad Pressure Vessel
-Air
-Reflective Insulation Bioshield Liner Plate Bioshield - Concrete Surveillance Capsule Holder Surveillance Capsule Bottom End Plugs Bottom Water Gap Bottom Nozzle Plate Bottom Nozzle Leg Lower Core Plate (Reg A)
Lower Core Plate (Reg B)
Top Water Gap Top Nozzle Top Water Gap Top Nozzle Upper Core Plate (Reg A)
Upper Core Plate (Reg B)
Outlet Plenum (Reg A)
Outlet Plenum (Reg B)
Lower Core Supports (Reg A)
-Lower Core Supports (Reg B)
Inlet Plenum Core - Lower Blanket Core - Upper Blanket
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 3-1 WCAP-18746-NP January 2023 Revision 2 3
FRACTURE TOUGHNESS PROPERTIES The requirements for reactor vessel integrity and P-T limit curve development are specified in 10 CFR 50, Appendix G [4]. The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:
the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.
The Prairie Island Unit 1 beltline materials traditionally included Nozzle (a.k.a. Upper) Shell Forging B, Intermediate Shell Forging C, Lower Shell Forging D, Nozzle Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2, and Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. The Prairie Island Unit 2 beltline materials traditionally included Upper Shell Forging B, Intermediate Shell Forging C, Lower Shell Forging D, Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2, and Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. However, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [6], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period should be considered to experience neutron fluence sufficient to cause embrittlement. The additional materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met. Since no additional materials are expected to reach this threshold by the EOLE per Table 2-8 (Unit 1) and Table 2-17 (Unit 2), no extended beltline materials are identified.
A summary of the best-estimate copper (Cu) and nickel (Ni), contents, in units of weight percent (wt. %),
as well as initial RTNDT and initial USE values for the reactor vessel beltline materials are provided in Table 3-1 and Table 3-2 for Prairie Island Unit 1 and 2, respectively. Table 3-3 contains a summary of the initial RTNDT values of the reactor vessel flange and reactor vessel closure head flange. These flange initial RTNDT values serve as input to the P-T limit curves flange-notch per Appendix G of 10 CFR 50 - See Section 6.3 for details.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 3-2 WCAP-18746-NP January 2023 Revision 2 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 1 Reactor Vessel Materials(a)
Reactor Vessel Material and Identification Number Heat Number Flux Type (Lot)
Chemical Composition Fracture Toughness Property Wt.
% Cu Wt.
Ni Initial RTNDT
(°F)
I
(°F)
Initial Upper-Shelf Energy (ft-lb)
Nozzle Shell Forging B 21744/38384 0.08 0.68
-4 0
84 Intermediate Shell Forging C 21918/38566 0.07 0.80 14 0
143 Lower Shell Forging D 21887/38530 0.07 0.66
-4 0
88 Nozzle Shell to Intermediate Shell Circumferential Weld Seam W2 2269 UM 89 (1180) 0.15 0.15 0(b) 17(b) 84 Intermediate Shell to Lower Shell Circumferential Weld Seam W3 1752 UM 89 (1230) 0.13 0.13
-13 0
78.5 Surveillance Materials Intermediate Shell Forging C 21918/38566 Surveillance Weld 1752 UM 89 (1230) 0.14 0.11 Notes:
(a) Values are measured values from the CMTRs or other Prairie Island Unit 1 fabrication records unless otherwise noted.
(b) The initial RTNDT of weld seam W2 is estimated based on Branch Technical Position 5-3 [7]; therefore, I = 17°F.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 3-3 WCAP-18746-NP January 2023 Revision 2 Table 3-2 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Prairie Island Unit 2 Reactor Vessel Materials(a)
Reactor Vessel Material and Identification Number Heat Number Flux Type (Lot)
Chemical Composition Fracture Toughness Property Wt.
% Cu Wt.
Ni Initial RTNDT
(°F)
I
(°F)
Initial Upper-Shelf Energy (ft-lb)
Upper Shell Forging B 22231/39088 0.07 0.73
-13 0
85 Intermediate Shell Forging C 22829 0.07 0.75 14 0
112 Lower Shell Forging D 22642 0.08 0.67
-4 0
108 Upper Shell to Intermediate Shell Circumferential Weld Seam W2 1752 UM 89 (1263) 0.13(b) 0.13(b)
-13(b) 0 78.5(b)
Intermediate Shell to Lower Shell Circumferential Weld Seam W3 2721 UM 89 (1263) 0.09 0.11
-31 0
103 Surveillance Materials Lower Shell Forging D 22642 Surveillance Weld 2721 UM 89 (1263) 0.08 0.08 Surveillance Weld Material from Prairie Island Unit 1 1752 UM 89 (1230) 0.14 0.11 Notes:
(a) Values are measured values from the CMTRs or other Prairie Island Unit 2 fabrication records unless otherwise noted.
(b) Consistent with Prairie Island Unit 1 Intermediate Shell to Lower Shell Circumferential Weld Seam W3.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 3-4 WCAP-18746-NP January 2023 Revision 2 Table 3-3 Summary of Prairie Island Units 1 and 2 Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values Reactor Vessel Material Unit 1 Initial RTNDT
(°F)
Unit 2 Initial RTNDT
(°F)
Replacement Reactor Vessel Closure Head
-70
-50 Vessel Flange
-4
-22
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 4-1 WCAP-18746-NP January 2023 Revision 2 4
SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2 [1], calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, surveillance data generated from surveillance programs at other plants which include Prairie Island Unit 1 or 2 reactor vessel material should also be considered when calculating Position 2.1 chemistry factors. Note that material from Heat # 1752 from the Prairie Island Unit 1 surveillance program in used in the Prairie Island Unit 2 reactor vessel.
The surveillance capsule forging material for Prairie Island Unit 1 is from Intermediate Shell Forging C.
Per WCAP-18660-NP [17], the surveillance forging data are deemed non-credible. The Prairie Island Unit 1 surveillance weld specimens were fabricated from weld material Heat # 1752 flux type UM 89, Lot # 1230.
This weld material is applicable to the Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. Per WCAP-18660-NP, the surveillance weld data are deemed non-credible. Therefore, a reduced margin term cannot be utilized with the Position 2.1 chemistry factors in the ART or PTS calculations for these materials contained in Section 7 and Appendix D, respectively.
The surveillance capsule forging material for Prairie Island Unit 2 is from Lower Shell Forging D. Per Appendix D of WCAP-18795-NP [20] and as restated in Appendix E, the surveillance forging data are deemed non-credible; therefore, a reduced margin term cannot be utilized with the Position 2.1 chemistry factor in the Lower Shell Forging D ART or PTS calculations contained in Section 7 and Appendix D, respectively. The Prairie Island Unit 2 surveillance weld specimens were fabricated from weld material Heat # 2721 flux type UM 89, Lot # 1263. This weld material is applicable to the Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3. The surveillance weld data are deemed credible in Appendix E; therefore, a reduced margin term will be utilized with the Position 2.1 chemistry factor in the ART and PTS calculations for these welds contained in Section 7 and Appendix D, respectively.
Table 4-1 and Table 4-2 summarize the surveillance data available for the Prairie Island Units 1 and 2 forging and weld materials that will be used in the calculation of the Position 2.1 chemistry factor values, respectively.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 4-2 WCAP-18746-NP January 2023 Revision 2 Table 4-1 Prairie Island Unit 1 Surveillance Capsule Data(a)
Material Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV)
Measured 30 ft-lb Transition Temperature Shift
(°F)
Measured Upper-Shelf Energy Decrease
(%)
Intermediate Shell Forging C (Tangential)
V 0.609 31.9 2
P 1.31 23.2 10 R
4.02 96.0 9
S 4.39 101.6 10 N
8.45 165.5 13 Intermediate Shell Forging C (Axial)
V 0.609 48.8 0(b)
P 1.31 33.9 5
R 4.02 84.0 10 S
4.39 74.0 6
N 8.45 147.5 16 Surveillance Weld (Heat # 1752)
V 0.609 35.2 0(b)
P 1.31 45.6 0(b)
R 4.02 123.0 9
S 4.39 161.0 0(b)
N 8.45 219.0 4
Notes:
(a) Surveillance data from WCAP-18660-NP [17].
(b) An increase in USE was measured, which physically should not occur after irradiation. Therefore, a conservative 0%
decrease value is used.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 4-3 WCAP-18746-NP January 2023 Revision 2 Table 4-2 Prairie Island Unit 2 Surveillance Capsule Data(a)
Material Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV)
Measured 30 ft-lb Transition Temperature Shift
(°F)
Measured Upper-Shelf Energy Decrease
(%)
Lower Shell Forging D (Axial)
V 0.598 35.0 0
T 1.10 27.9 13 R
4.11 84.3 7
P 4.27 103.5 11 N
8.41 154.1 16 Lower Shell Forging D (Tangential)
V 0.598 33.8 0
T 1.10 54.4 10 R
4.11 89.6 14 P
4.27 99.6 13 N
8.41 176.5 15 Surveillance Weld (Heat # 2721)
V 0.598 69.3 6
T 1.10 57.7 8
R 4.11 100.3 12 P
4.27 96.2 5
N 8.41 135.6 8
Note:
(a) Surveillance data from WCAP-18795-NP [20].
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-1 WCAP-18746-NP January 2023 Revision 2 5
CHEMISTRY FACTORS The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2, Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.
The best-estimate copper and nickel weight percent values for the Prairie Island Units 1 and 2 reactor vessel materials are provided in Table 3-1 and Table 3-2 of this report, respectively.
The Position 2.1 chemistry factor calculation is presented in Table 5-1 and Table 5-2 for the Prairie Island Units 1 and 2 surveillance materials, respectively. These values were calculated using the surveillance data summarized in Section 4 of this report. The calculations are performed using the method described in Regulatory Guide 1.99, Revision 2. All of the surveillance weld data considers the chemical composition differences between the surveillance weld and the weld being evaluated, in accordance with the guidance presented at an industry meeting held by the NRC on February 12 and 13, 1998 [11]. In addition to the chemical composition differences, temperature adjustments are considered for the surveillance material from other plants. In this case, the Unit 1 surveillance weld data includes a temperature consideration when applied to the Unit 2 reactor vessel weld. Margin will be applied to the ART calculations in Section 7 and PTS calculations in Appendix D according to the conclusions of the credibility evaluation for the surveillance material, as documented in Section 4.
The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-3 and Table 5-4 for Prairie Island Units 1 and 2, respectively.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-2 WCAP-18746-NP January 2023 Revision 2 Table 5-1 Calculation of Prairie Island Unit 1 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule Fluence(a)
(x 1019 n/cm2, E > 1.0 MeV)
FF(b)
RTNDT(c)
(°F)
FF*RTNDT
(°F)
FF2 Intermediate Shell Forging C (Tangential)
V 0.609 0.861 31.9 27.5 0.741 P
1.31 1.075 23.2 24.9 1.156 R
4.02 1.357 96.0 130.3 1.842 S
4.39 1.376 101.6 139.8 1.893 N
8.45 1.491 165.5 246.8 2.224 Intermediate Shell Forging C (Axial)
V 0.609 0.861 48.8 42.0 0.741 P
1.31 1.075 33.9 36.4 1.156 R
4.02 1.357 84.0 114.0 1.842 S
4.39 1.376 74.0 101.8 1.893 N
8.45 1.491 147.5 220.0 2.224 SUM:
1083.6 15.715 CF IS Forging C = (FF
V 0.609 0.861 34.5(d)
(35.2) 29.7 0.741 P
1.31 1.075 44.7(d)
(45.6) 48.0 1.156 R
4.02 1.357 120.5(d)
(123.0) 163.6 1.842 S
4.39 1.376 157.8(d)
(161.0) 217.1 1.893 N
8.45 1.491 214.6(d)
(219.0) 320.1 2.224 SUM:
778.6 7.857 CF Surv. Weld = (FF
- RTNDT) ÷ (FF2) = (778.6) ÷ (7.857) = 99.1°F Notes:
(a) Data taken from Table 4-1.
(b) FF = fluence factor = f(0.28 - 0.10*log f).
(c) RTNDT values taken from Table 4-1.
(d) The surveillance weld RTNDT values have been adjusted by a factor of 0.98. The calculated adjustment is the ratio of the Position 1.1 CFs of the vessel weld to the surveillance weld (CFVessel Weld / CFSurv. Weld), which is (69.7°F / 70.9°F). The measured (unadjusted) RTNDT values are shown in parenthesis.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-3 WCAP-18746-NP January 2023 Revision 2 Table 5-2 Calculation of Prairie Island Unit 2 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule Fluence(a)
(x 1019 n/cm2, E > 1.0 MeV)
FF(b)
RTNDT(c)
(°F)
FF*RTNDT
(°F)
FF2 Lower Shell Forging D
(Axial)
V 0.598 0.856 35.0 30.0 0.733 T
1.10 1.027 27.9 28.6 1.054 R
4.11 1.362 84.3 114.8 1.855 P
4.27 1.370 103.5 141.8 1.877 N
8.41 1.491 154.1 229.7 2.222 Lower Shell Forging D
(Tangential)
V 0.598 0.856 33.8 28.9 0.733 T
1.10 1.027 54.4 55.8 1.054 R
4.11 1.362 89.6 122.0 1.855 P
4.27 1.370 99.6 136.5 1.877 N
8.41 1.491 176.5 263.1 2.222 SUM:
1151.4 15.484 CF LS Forging D = (FF
V 0.598 0.856 79.7(d)
(69.3) 68.2 0.733 T
1.10 1.027 66.4(d)
(57.7) 68.1 1.054 R
4.11 1.362 115.3(d)
(100.3) 157.1 1.855 P
4.27 1.370 110.6(d)
(96.2) 151.6 1.877 N
8.41 1.491 155.9(d)
(135.6) 232.5 2.222 SUM:
677.5 7.742 CF Surv. Weld = (FF
- RTNDT) ÷ (FF2) = (677.5) ÷ (7.742) = 87.5°F Notes:
(a) Data taken from Table 4-2.
(b) FF = fluence factor = f(0.28 - 0.10*log f).
(c) RTNDT values taken from Table 4-2.
(d) The surveillance weld RTNDT values have been adjusted by a factor of 1.15. The calculated adjustment is the ratio of the Position 1.1 CFs of the vessel weld to the surveillance weld (CFVessel Weld / CFSurv. Weld), which is (51.6°F / 44.8°F). The measured (unadjusted) RTNDT values are shown in parenthesis.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-4 WCAP-18746-NP January 2023 Revision 2 Table 5-3 Summary of Prairie Island Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material and Identification Number Heat Number (Lot)
Chemistry Factor (°F)
Position 1.1(a)
Position 2.1(b)
Nozzle Shell Forging B 21744/38384 51.0 Intermediate Shell Forging C 21918/38566 44.0 69.0 Lower Shell Forging D 21887/38530 44.0 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 2269 (1180) 79.5 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 1752 (1230) 69.7 99.1 Reactor Vessel Surveillance Material Surveillance Program Weld Metal 1752 (1230) 70.9 Notes:
(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of this report and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 [1].
(b) Position 2.1 chemistry factors were taken from Table 5-1 of this report. As discussed in Section 4, both the surveillance forging data and the surveillance weld data were deemed non-credible.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-5 WCAP-18746-NP January 2023 Revision 2 Table 5-4 Summary of Prairie Island Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material and Identification Number Heat Number (Lot)
Chemistry Factor (°F)
Position 1.1(a)
Position 2.1(b)
Upper Shell Forging B 22231/39088 44.0 Intermediate Shell Forging C 22829 44.0 Lower Shell Forging D 22642 51.0 74.4 Upper Shell to Intermediate Shell Circumferential Weld Seam W2 1752 (1263) 69.7 101.4(c)
Intermediate Shell to Lower Shell Circumferential Weld Seam W3 2721 (1263) 51.6 87.5 Reactor Vessel Surveillance Material Prairie Island Unit 2 Surveillance Program Weld Metal 2721 (1263) 44.8 Prairie Island Unit 1 Surveillance Program Weld Metal 1752 (1230) 70.9 Notes:
(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of this report and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 [1].
(b) Position 2.1 chemistry factors were taken from Table 5-2 of this report. As discussed in Section 4, the surveillance forging data was deemed non-credible and the surveillance weld data was deemed credible.
(c) Value is determined by recalculating the surveillance weld CF from Table 5-1 and adding the 3°F temperature adjustment before multiplying by the 0.98 CF ratio. This temperature adjustment is meant to account for the Prairie Island Unit 1 surveillance weld material irradiation temperature of 536°F and the time-weighted average inlet temperature for Prairie Island Unit 2 over 54 EFPY of 533°F.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 6-1 WCAP-18746-NP January 2023 Revision 2 6
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALL APPROACH The American Society of Mechanical Engineers (ASME) approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIc, for the metal temperature at that time. KIc is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [3]. The KIc curve is given by the following equation:
)]
(
02
.0
[
734 20 2.
Ic e
K
+
=
(1)
- where, KIc (ksiin.)
=
reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This KIc curve is based on the lower bound of static critical KI values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.
6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C* KIm + KIt < KIc (2)
- where, KIm
=
stress intensity factor caused by membrane (pressure) stress KIt
=
stress intensity factor caused by the thermal gradients KIc
=
reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT C
=
2.0 for Level A and Level B service limits C
=
1.5 for hydrostatic and leak test conditions during which the reactor core is not critical
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 6-2 WCAP-18746-NP January 2023 Revision 2 For membrane tension, the corresponding KI for the postulated defect is:
)
/
(
Im t
pR M
K i
m
=
(3)
Axial Flaw Methodology where, Mm for an inside axial surface flaw is given by:
Mm
=
1.85 for t < 2, Mm
=
0.926 t for 2
t 3464 Mm
=
3.21 for t > 3.464 and, Mm for an outside axial surface flaw is given by:
Mm
=
1.77 for t < 2, Mm
=
0.893 t for 2 464
.3
t Mm
=
3.09 for t > 3.464 Circumferential Flaw Methodology Similarly, Mm for an inside or an outside circumferential surface flaw is given by:
Mm
=
0.89 for t < 2, Mm
=
0.443 t for 2 464
.3
t Mm
=
1.53 for t > 3.464 where, p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).
For bending stress, the corresponding KI for the postulated axial or circumferential defect is:
KIb = Mb
- Maximum Bending Stress, where Mb is two-thirds of Mm (4)
The maximum KI produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:
KIt = 0.953 x 10-3 x CR x t2.5 (5) where CR is the cooldown rate in F/hr., or for a postulated axial or circumferential outside surface defect
_f J
_f
_f
_f J
_f
_f
_f J
_f
_f
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 6-3 WCAP-18746-NP January 2023 Revision 2 KIt = 0.753 x 10-3 x HU x t2.5 (6) where HU is the heatup rate in F/hr.
The through-wall temperature difference associated with the maximum thermal KI can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal KI:
(a) The maximum thermal KI relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).
(b) Alternatively, the KI for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness axial or circumferential inside surface defect using the relationship:
K C
C C
C a
It =
+
+
+
(.
)*
10359 06322 04753 03855 0
1 2
3 (7) or similarly, KIt during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:
a C
C C
C KIt
)
401
.0 481
.0 630
.0 043
.1(
3 2
1 0
+
+
+
=
(8) where the coefficients C0, C1, C2, and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
( )
( / )
( / )
( / )
x C
C x a C x a C x a
=
+
+
+
0 1
2 2
3 3
(9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).
Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves [2] Section 2.6 (Equations 2.6.2-4 and 2.6.3-1). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (Equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft2/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-°F.
At any time during the heatup or cooldown transient, KIc is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, Paragraph G-2120), the appropriate value for RTNDT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature
_f J
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 6-4 WCAP-18746-NP January 2023 Revision 2 gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable P-T curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the T (temperature) across the vessel wall developed during cooldown results in a higher value of KIc at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIc exceeds KIt, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable P-T relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIc for the inside 1/4T flaw during heatup is lower than the KIc for the flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIc values do not offset each other, and the P-T curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The third portion of the heatup analysis concerns the calculation of the P-T limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 6-5 WCAP-18746-NP January 2023 Revision 2 ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of P-T curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20% of the preservice hydrostatic test pressure, which is calculated to be 621 psig. The initial RTNDT values of the reactor vessel closure head and vessel flange are documented in Table 3-3. The limiting unirradiated RTNDT of -4°F for Prairie Island Units 1 and 2 is associated with the vessel flange of the Prairie Island Unit 1 vessel, so the minimum allowable temperature of this region is 116°F at pressures greater than 621 psig without margins for instrument uncertainties. This limit is shown in Figure 8-1 and Figure 8-2.
6.4 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RTNDT, in the closure flange region. This requirement is established in Appendix G to 10 CFR 50 [4]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [2], the minimum boltup temperature should be 60°F or the limiting unirradiated RTNDT of the closure flange region, whichever is higher. Since the limiting unirradiated RTNDT of this region is below 60F per Table 3-3, the minimum boltup temperature for the Prairie Island Units 1 and 2 reactor vessels is 60°F without margins for instrument uncertainties. This limit is shown in Figure 8-1 and Figure 8-2.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-1 WCAP-18746-NP January 2023 Revision 2 7
CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [1], the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = Initial RTNDT + RTNDT + Margin (10)
Initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [8]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.
RTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
- f (0.28 - 0.10 log f)
(11)
To calculate RTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth:
f(depth x) = fsurface
- e (-0.24x)
(12) where x inches (reactor vessel cylindrical shell beltline thickness is 6.692 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the RTNDT at the specific depth.
The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-18124-NP-A.
Table 7-1 and Table 7-2 contain the surface fluence values at 54 EFPY for Prairie Island Units 1 and 2, respectively, which were used for the development of the P-T limit curves contained in this report. Table 7-1 and Table 7-2 also contain the 1/4T and 3/4T calculated fluence values and fluence factors (FFs), per Regulatory Guide 1.99, Revision 2. The values in this table will be used to calculate the 54 EFPY ART values for the Prairie Island Units 1 and 2 reactor vessel materials.
Margin is calculated as M = 2 2
2
+
I
. The standard deviation for the initial RTNDT margin term (I) is 0F when the initial RTNDT is a measured value, and 17F when a generic value is available. The standard deviation for the RTNDT margin term,, is 17F for plates or forgings when surveillance data is not used or is non-credible, and 8.5F (half the value) for plates or forgings when credible surveillance data is used.
For welds, is equal to 28F when surveillance capsule data is not used or is non-credible and is 14F (half the value) when credible surveillance capsule data is used. The value for need not exceed 0.5 times the mean value of RTNDT.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-2 WCAP-18746-NP January 2023 Revision 2 Contained in Table 7-3 and Table 7-4 are the 54 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Prairie Island Unit 1 heatup and cooldown curves. Contained in Table 7-5 and Table 7-6 are the 54 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Prairie Island Unit 2 heatup and cooldown curves.
The inlet/outlet nozzle forging weld materials for Prairie Island Units 1 and 2 have projected fluence values that do not exceed the 1 x 1017 n/cm2 fluence threshold of RIS 2014-11 [6] at 54 EFPY per Table 2-8 (Unit 1) and Table 2-17 (Unit 2). Note that neither Table 2-8 nor Table 2-17 provide fluence values for the outlet nozzle, which are at a higher elevation than the inlet nozzles. The projected fluence value for the inlet nozzle forging weld material provides a conservative estimate of the fluence values of the outlet nozzles. Therefore, neutron radiation embrittlement need not be considered herein for the nozzle forging or weld materials.
Thus, ART calculations for the inlet and outlet nozzle forging and weld materials are excluded from Table 7-3 through Table 7-6.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-3 WCAP-18746-NP January 2023 Revision 2 Table 7-1 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 1 Reactor Vessel at 54 EFPY Reactor Vessel Material Surface Fluence(a)
(n/cm2, E > 1.0 MeV)
Surface FF(b) 1/4T Fluence(a)
(n/cm2, E > 1.0 MeV) 1/4T FF(b) 3/4T Fluence(a)
(n/cm2, E > 1.0 MeV) 3/4T FF(b)
Nozzle Shell Forging B 3.37 1.318 2.26 1.220 1.01 1.003 Intermediate Shell Forging C 5.63 1.425 3.77 1.343 1.69 1.144 Lower Shell Forging D 5.53 1.422 3.70 1.339 1.66 1.139 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 3.63 1.335 2.43 1.239 1.09 1.024 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 5.53 1.422 3.70 1.339 1.66 1.139 Notes:
(a) 54 EFPY fluence values are documented in Table 2-8. 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (6.692 inches) and equation f = fsurf
- e-0.24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).
(b) FF = fluence factor = f(0.28 - 0.10*log (f)).
Table 7-2 Fluence Values and Fluence Factors for the Beltline Materials at the Vessel Surface, 1/4T and 3/4T Locations for the Prairie Island Unit 2 Reactor Vessel at 54 EFPY Reactor Vessel Material Surface Fluence(a)
(n/cm2, E > 1.0 MeV)
Surface FF(b) 1/4T Fluence(a)
(n/cm2, E > 1.0 MeV) 1/4T FF(b) 3/4T Fluence(a)
(n/cm2, E > 1.0 MeV) 3/4T FF(b)
Upper Shell Forging B 3.37 1.318 2.26 1.220 1.01 1.003 Intermediate Shell Forging C 5.66 1.426 3.79 1.344 1.70 1.146 Lower Shell Forging D 5.58 1.423 3.73 1.341 1.67 1.142 Upper Shell to Intermediate Shell Circumferential Weld - Seam W2 3.61 1.334 2.42 1.238 1.08 1.022 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 5.58 1.423 3.73 1.341 1.67 1.142 Notes:
(a) 54 EFPY fluence values are documented in Table 2-17. 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (6.692 inches) and equation f = fsurf
- e-0.24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).
(b) FF = fluence factor = f(0.28 - 0.10*log (f)).
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-4 WCAP-18746-NP January 2023 Revision 2 Table 7-3 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)
(°F) 1/4T Fluence(b)
(n/cm2, E > 1.0 MeV) 1/4T FF(b)
RTNDT(U)(c)
(°F)
(°F)
I
(°F)
(d)
(°F)
Margin
(°F)
(°F)
Nozzle Shell Forging B 1.1 51.0 2.26 1.220
-4 62.2 0
17.0 34.0 92.2 Intermediate Shell Forging C 1.1 44.0 3.77 1.343 14 59.1 0
17.0 34.0 107.1 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 69.0 3.77 1.343 14 92.7 0
17.0 34.0 140.7 Lower Shell Forging D 1.1 44.0 3.70 1.339
-4 58.9 0
17.0 34.0 88.9 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 1.1 79.5 2.43 1.239 0
98.5 17 28.0 65.5 164.0 Intermediate to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 1.1 69.7 3.70 1.339
-13 93.3 0
28.0 56.0 136.3 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 99.1 3.70 1.339
-13 132.7 0
28.0 56.0 175.7 Notes:
(a) Data is from Table 5-3.
(b) Data is from Table 7-1.
(c) Data is from Table 3-1.
(d) As discussed in Section 4, the intermediate shell forging material surveillance data and the weld Heat # 1752 surveillance data were both determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or Position 2.1 with non-credible surveillance data and
= 8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-5 WCAP-18746-NP January 2023 Revision 2 Table 7-4 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)
(°F) 3/4T Fluence(b)
(n/cm2, E > 1.0 MeV) 3/4T FF(b)
RTNDT(U)(c)
(°F)
(°F)
I
(°F)
(d)
(°F)
Margin
(°F)
(°F)
Nozzle Shell Forging B 1.1 51.0 1.01 1.003
-4 51.1 0
17.0 34.0 81.1 Intermediate Shell Forging C 1.1 44.0 1.69 1.144 14 50.3 0
17.0 34.0 98.3 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 69.0 1.69 1.144 14 78.9 0
17.0 34.0 126.9 Lower Shell Forging D 1.1 44.0 1.66 1.139
-4 50.1 0
17.0 34.0 80.1 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 1.1 79.5 1.09 1.024 0
81.4 17 28.0 65.5 146.9 Intermediate to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 1.1 69.7 1.66 1.139
-13 79.4 0
28.0 56.0 122.4 Using Non-credible Prairie Island Unit 1 Surveillance Data 2.1 99.1 1.66 1.139
-13 112.9 0
28.0 56.0 155.9 Notes:
(a) Data is from Table 5-3.
(b) Data is from Table 7-1.
(c) Data is from Table 3-1.
(d) As discussed in Section 4, the intermediate shell forging material surveillance data and the weld Heat # 1752 surveillance data were both determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or Position 2.1 with non-credible surveillance data and = 8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-6 WCAP-18746-NP January 2023 Revision 2 Table 7-5 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)
(°F) 1/4T Fluence(b)
(n/cm2, E > 1.0 MeV) 1/4T FF(b)
RTNDT(U)(c)
(°F)
(°F)
I
(°F)
(d)
(°F)
Margin
(°F)
(°F)
Upper Shell Forging B 1.1 44.0 2.26 1.220
-13 53.7 0
17.0 34.0 74.7 Intermediate Shell Forging C 1.1 44.0 3.79 1.344 14 59.2 0
17.0 34.0 107.2 Lower Shell Forging D 1.1 51.0 3.73 1.341
-4 68.4 0
17.0 34.0 98.4 Using Non-credible Prairie Island Unit 2 Surveillance Data 2.1 74.4 3.73 1.341
-4 99.8 0
17.0 34.0 129.8 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 1.1 69.7 2.42 1.238
-13 86.3 0
28.0 56.0 129.3 Using Non-credible Prairie Island Unit 1 Surveillance Data (Heat # 1752) 2.1 101.4 2.42 1.238
-13 125.5 0
28.0 56.0 168.5 Intermediate to Lower Shell Circumferential Weld - Seam W3 1.1 51.6 3.73 1.341
-31 69.2 0
28.0 56.0 94.2 Using Credible Prairie Island Unit 2 Surveillance Data (Heat # 2721) 2.1 87.5 3.73 1.341
-31 117.3 0
14.0 28.0 114.3 Notes:
(a) Data is from Table 5-4.
(b) Data is from Table 7-2.
(c) Data is from Table 3-2.
(d) As discussed in Section 4, the lower shell forging material surveillance data was determined to be non-credible, while the weld Heat # 2721 surveillance data was determined to be credible. In addition, as discussed in Section 4, the weld Heat # 1752 surveillance data was determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or Position 2.1 with non-credible surveillance data and = 8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-7 WCAP-18746-NP January 2023 Revision 2 Table 7-6 Adjusted Reference Temperature Evaluation for the Prairie Island Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF(a)
(°F) 3/4T Fluence(b)
(n/cm2, E > 1.0 MeV) 3/4T FF(b)
RTNDT(U)(c)
(°F)
(°F)
I
(°F)
(d)
(°F)
Margin
(°F)
(°F)
Upper Shell Forging B 1.1 44.0 1.01 1.003
-13 44.1 0
17.0 34.0 65.1 Intermediate Shell Forging C 1.1 44.0 1.70 1.146 14 50.4 0
17.0 34.0 98.4 Lower Shell Forging D 1.1 51.0 1.67 1.142
-4 58.2 0
17.0 34.0 88.2 Using Non-credible Prairie Island Unit 2 Surveillance Data 2.1 74.4 1.67 1.142
-4 85.0 0
17.0 34.0 115.0 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 1.1 69.7 1.08 1.022
-13 71.2 0
28.0 56.0 114.2 Using Non-credible Prairie Island Unit 1 Surveillance Data (Heat # 1752) 2.1 101.4 1.08 1.022
-13 103.6 0
28.0 56.0 146.6 Intermediate to Lower Shell Circumferential Weld - Seam W3 1.1 51.6 1.67 1.142
-31 58.9 0
28.0 56.0 83.9 Using Credible Prairie Island Unit 2 Surveillance Data (Heat # 2721) 2.1 87.5 1.67 1.142
-31 99.9 0
14.0 28.0 96.9 Notes:
(a) Data is from Table 5-4.
(b) Data is from Table 7-2.
(c) Data is from Table 3-2.
(d) As discussed in Section 4, the lower shell forging material surveillance data was determined to be non-credible, while the weld Heat # 2721 surveillance data was determined to be credible. In addition, as discussed in Section 4, the weld Heat # 1752 surveillance data was determined to be non-credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or Position 2.1 with non-credible surveillance data and = 8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-1 WCAP-18746-NP January 2023 Revision 2 8
HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES P-T limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4 [2].
The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figure 8-1 through Figure 8-2. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed in the following paragraphs.
The reactor must not be made critical until P-T combinations are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:
1.5 KIm < KIc (13)
- where, KIm is the stress intensity factor covered by membrane (pressure) stress [see Equation (3) in Section 6],
KIc = 33.2 + 20.734 e [0.02 (T - RTNDT)] [see Equation (1) in Section 6],
T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.
The criticality limit curve specifies P-T limits for core operation in order to provide additional margin during actual power production. The P-T limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding P-T curve for heatup and cooldown calculated as described in Section 6 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic leak tests for the Prairie Island Units 1 and 2 reactor vessels at 54 EFPY is 219F without uncertainties. This temperature is the minimum permissible temperature at which design pressure can be reached during a hydrostatic test per Equation (13). The vertical line drawn from these points on the P-T curve, intersecting a curve 40°F higher than the P-T limit curve, constitutes the limit for core operation for the reactor vessel.
The ART values were calculated in Table 7-3 through Table 7-6. The limiting ART values for Prairie Island Units 1 and 2 are for Unit 1 Intermediate Shell to Lower Shell Circumferential Weld Seam W3 (Heat
- 1752). However, since this material is a Circumferential Flaw material, the applied membrane
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-2 WCAP-18746-NP January 2023 Revision 2 (pressure) stress and resulting stress intensity factor at the postulated flaw location are much lower than for the most limiting Axial Flaw material. Consequently, this material may not produce the most limiting P-T limit curves. This is illustrated in the figures in Section 9. The ART values considered for circumferential flaws in the generation of the 54 EFPY P-T limit curves are provided in Table 8-1, and bound the ART values in Table 7-3 through Table 7-6. Therefore, the most limiting material with an axially oriented flaw must also be considered. The limiting material with an axially oriented flaw is Unit 1 Intermediate Shell Forging C. It is noted that the 1/4T and 3/4T ART values for Unit 1 Intermediate Shell Forging C at 54 EFPY (140.7°F and 126.9°F, respectively) are below those used to generate the P-T curves in WCAP-14780 and currently implemented in the PTLR (154°F and 136°F, respectively); thereby, validating the PTLR curve for the material with a postulated axial flaw. However, the ART values for Unit 1 Intermediate Shell Forging C were conservatively increased to account for future perturbations such as the Unit 2 Surveillance Capsule N results. The ART values, considered in the P-T limit curves development are summarized in Table 8-1.
Table 8-1 Summary of the ART Values Used in the Generation of the Prairie Island Units 1 and 2 Heatup and Cooldown Curves at 54 EFPY Flaw Orientation 1/4T Limiting ART 3/4T Limiting ART Axial 170°F 160°F Unit 1 Intermediate Shell Forging C Circumferential 179°F 159°F Unit 1 Intermediate Shell to Lower Shell Circumferential Weld Seam W3 (Heat # 1752)
Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, -20, -40, -60, and -100°F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. The data points used for developing the heatup and cooldown P-T limit curves shown in Figure 8-1 and Figure 8-2 are presented in Table 8-2 and Table 8-3. Vacuum refill limits for the Reactor Coolant System (RCS) are displayed on Figure 8-1 and Figure 8-2 by showing a minimum pressure of 0 psia.
Inlet and Outlet Nozzles P-T Limit Curves NRC Regulatory Issue Summary (RIS) 2014-11 [6] requires that the P-T limit curves account for the higher stresses in the nozzle corner region due to the potential for more restrictive P-T limits, even if the RTNDT for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.
PWROG-15109-NP-A [16] addresses this concern generically for the U.S. pressurized water reactor (PWR) operating fleet. The results of PWROG-15109-NP-A demonstrate that P-T limit curves developed with current NRC-approved methods (e.g. WCAP-14040-A) bound the generic nozzle P-T limit curves. The
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-3 WCAP-18746-NP January 2023 Revision 2 results and conclusions of PWROG-15109-NP-A are applicable as long as the plant-specific Prairie Island Units 1 and 2 fluence of the nozzle corners remains less than the screening criterion of 4.28 x 1017 n/cm2, as described in PWROG-15109-NP-A. Table 2-8 and Table 2-17 demonstrate Prairie Island Units 1 and 2 adherence to this screening criterion; thus, PWROG-15109-NP-A is applicable.
In conclusion, PWROG-15109-NP-A demonstrates that the nozzles will not be limiting with respect to the P-T limit curves at Prairie Island Units 1 and 2. Therefore, the concerns the concerns of RIS 2014-11 are adequately addressed.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-4 WCAP-18746-NP January 2023 Revision 2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Unit 1, Intermediate Shell Forging C LIMITING ART VALUES AT 54 EFPY:
1/4T, 170F (Axial Flaw) 3/4T, 160F (Axial Flaw)
Figure 8-1 Prairie Island Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100oF/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/KIc)
-250 0
250 500 750 1000 1250 1500 1750 2000 2250 2500 0
50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (PSIG)
Moderator Temperature (Deg. F)
OperlimAnalysis Version:5.4 Run:27388 Operlim.xlsm Version: 5.4.1 Unacceptable Operation Acceptable Operation Criticality Limit based on inservice hydrostatic test temperature (219ºF) for the service period up to 54 EFPY Heatup Rate 60 F/Hr Heatup Rate 100 F/Hr Critical Limit 60 F/Hr Critical Limit 100 F/Hr Leak Test Limit Boltup Temp.
Prairie Island Unit 1 54 EFPY Curves using K1c, Axial Flaw, No instrumentation errors, ART 170_160, No Flange Notch Lower Limit for RCS pressure is 0 psia
~
r---r---r-----.a.
~
~
r------....
~
I I
/)"'
I I
~J 1/
_ _J /7 I
~
I I
I I :-[jj I
I
~ I I
I I
r I/ II I
I I
j
~ j I~
I
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-5 WCAP-18746-NP January 2023 Revision 2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Unit 1, Intermediate Shell Forging C LIMITING ART VALUES AT 54 EFPY:
1/4T, 170F (Axial Flaw) 3/4T, 160F (Axial Flaw)
Figure 8-2 Prairie Island Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)
-250 0
250 500 750 1000 1250 1500 1750 2000 2250 2500 0
50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (PSIG)
Moderator Temperature (Deg. F)
OperlimAnalysis Version:5.4 Run:27388 Operlim.xlsm Version: 5.4.1 Unacceptable Operation Acceptable Operation Cooldown Rates 0 F/hr
-20 F/hr
-40 F/hr
-60 F/hr
-100 F/hr Prairie Island Unit 1 54 EFPY Curves using K1c, Axial Flaw, No instrumentation errors, ART 170_160, No Flange Notch Steady State and Cooldown Curves Lower Limit for RCS pressure is 0 psia I
I I
I I I J
J I
I I
I
~
1,..-
~
I I
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-6 WCAP-18746-NP January 2023 Revision 2 Table 8-2 Prairie Island Units 1 and 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 60
-14.7 219
-14.7 60
-14.7 219
-14.7 60 621 219 1101 60 621 219 954 65 621 220 1111 65 621 220 961 70 621 225 1159 70 621 225 998 75 621 230 1212 75 621 230 1039 80 621 235 1270 80 621 235 1084 85 621 240 1335 85 621 240 1134 90 621 245 1406 90 621 245 1189 95 621 250 1485 95 621 250 1250 100 621 255 1572 100 621 255 1317 105 621 260 1668 105 621 260 1391 110 621 265 1774 110 621 265 1472 115 621 270 1892 115 621 270 1563 116 621 275 2021 116 621 275 1662 116 783 280 2164 116 721 280 1772 120 793 285 2322 120 727 285 1894 125 807 125 736 290 2028 130 823 130 747 295 2175 135 840 135 759 300 2338 140 859 140 772 145 881 145 788 150 905 150 805 155 931 155 825 160 960 160 847 165 992 165 871 170 1028 170 898 175 1067 175 928 180 1111 180 961 185 1159 185 998 190 1212 190 1039 195 1270 195 1084 200 1335 200 1134 205 1406 205 1189
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-7 WCAP-18746-NP January 2023 Revision 2 Table 8-2 Prairie Island Units 1 and 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 210 1485 210 1250 215 1572 215 1317 220 1668 220 1391 225 1774 225 1472 230 1892 230 1563 235 2021 235 1662 240 2164 240 1772 245 2322 245 1894 250 2028 255 2175 260 2338 Leak Test Limit T (°F)
P (psig) 200 2000 219 2485
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-8 WCAP-18746-NP January 2023 Revision 2 Table 8-3 Prairie Island Units 1 and 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)
Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 60
-14.7 60
-14.7 60
-14.7 60
-14.7 60
-14.7 60 621 60 621 60 621 60 621 60 601 65 621 65 621 65 621 65 621 65 607 70 621 70 621 70 621 70 621 70 614 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 116 621 116 621 116 621 116 621 116 621 116 849 116 825 116 800 116 777 116 729 120 861 120 838 120 814 120 791 120 745 125 878 125 856 125 833 125 811 125 768 130 897 130 875 130 854 130 833 130 793 135 918 135 897 135 877 135 857 135 820 140 941 140 921 140 903 140 885 140 851 145 966 145 948 145 931 145 915 145 885 150 994 150 978 150 962 150 948 150 922 155 1025 155 1010 155 997 155 985 155 964 160 1059 160 1046 160 1035 160 1026 160 1010 165 1096 165 1086 165 1078 165 1071 165 1061 170 1138 170 1131 170 1125 170 1121 170 1118 175 1184 175 1179 175 1177 175 1176 175 1176 180 1235 180 1233 180 1233 180 1233 180 1233 185 1291 185 1291 185 1291 185 1291 185 1291 190 1353 190 1353 190 1353 190 1353 190 1353 195 1422 195 1422 195 1422 195 1422 195 1422 200 1498 200 1498 200 1498 200 1498 200 1498
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-9 WCAP-18746-NP January 2023 Revision 2 Table 8-3 Prairie Island Units 1 and 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)
Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 205 1581 205 1581 205 1581 205 1581 205 1581 210 1674 210 1674 210 1674 210 1674 210 1674 215 1777 215 1777 215 1777 215 1777 215 1777 220 1890 220 1890 220 1890 220 1890 220 1890 225 2015 225 2015 225 2015 225 2015 225 2015 230 2153 230 2153 230 2153 230 2153 230 2153 235 2306 235 2306 235 2306 235 2306 235 2306 240 2475 240 2475 240 2475 240 2475 240 2475
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 9-1 WCAP-18746-NP January 2023 Revision 2 9
HEATUP AMD COOLDOWN LIMITS APPLICABILITY AND MARGIN ASSESSMENT This section provides a comparison of the heatup and cooldown P-T limit curves currently implemented in the Prairie Island Units 1 and 2 PTLR and the heatup and cooldown P-T limit curves generated in this report.
The curves contained in the PTLR were generated in WCAP-14780 [9] and were generated consistent with WCAP-14040-A, Revision 2. WCAP-14780 used KIa stress intensity factors and an Axial-Flaw methodology even though the limiting materials are circumferential welds. KIa is a toughness based on the lower bound of crack arrest toughness, whereas KIc is a toughness based on the lower bound of crack initiation toughness. This was consistent with the requirements of Revision 2 of WCAP-14040-NP-A.
Since the time the curves from WCAP-14780 were generated, WCAP-14040-A has been revised to Revision 4 [2] and approved for general use by the NRC. As discussed in Section 6, WCAP-14040-A, Revision 4 utilizes the 1998 version of the ASME Code through the Summer 2000 Addenda, which allows the use of the less restrictive KIc stress intensity factors and allows the use of the less restrictive Circ-Flaw methodology (formerly known as ASME Code Cases N-640 and N-588, respectively). Because less restrictive methodologies are used in this report; the current PTLR curves may remain valid, despite an increase in the maximum ART values.
The curves generated in this report are compared to the P-T limit curves from WCAP-14780 in Figure 9-1 and Figure 9-2. The curves generated with an axial flaw and KIc are shown as dashed blue lines. The curves generated with a circumferential flaw and KIc are shown as orange dotted lines. The curves from WCAP-14780 are shown as solid green lines. Note that the curves do not have margins for instrumentation errors, but they do have the Appendix G flange notch requirements included with the exception of the circumferential flaw curves which do not have the flange notch included for visual clarity.
In the flange region, the minimum pressure difference, at a constant temperature, between the current P-T limit curves and the new curves developed in this report is 0 psid for the cooldown curves. This is driven by the flange notch requirements of 10 CFR 50, Appendix G [4] being identical for each curve. The pressure margins at temperatures past the flange notch region show greater margin, with the minimum margin being 158 psid at a temperature of 116°F during the 100°F/hr heatup.
P-T Limit Curve Applicability Conclusion As shown in Figure 9-1 and Figure 9-2, the margins between the curves developed in this report and the current P-T limit curves illustrate that the current P-T limit curves remain applicable through 54 EPFY.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 9-2 WCAP-18746-NP January 2023 Revision 2 Figure 9-1 Heatup Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws
-250 0
250 500 750 1000 1250 1500 1750 2000 2250 2500 0
50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (psig)
Moderator Temperature
(°F)
Axial Flaw - ART Values (170, 160)
Circ Flaw - ART Values (179, 159)
Acceptable Operation Unacceptable Operation I
I I I
I......-
...
- I
.. : I I
I I I
I I I '
. : I I I
. : I I I
.. I I
.. I I
I I I I I II I I I I I *
- I
'/
I I I I I I I I
! I
. I I
- I
~......
I: I j
I I
I I
I lj/
I I
I I
j I
I I
I i
I 1/
I I
,','!t
" 1/
I
. -~
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 9-3 WCAP-18746-NP January 2023 Revision 2 Figure 9-2 Cooldown Curves Comparison between the Prairie Island Units 1 and 2 PTLR KIa Curves with New KIc Curves with Axial and Circumferential Flaws
-250 0
250 500 750 1000 1250 1500 1750 2000 2250 2500 0
50 100 150 200 250 300 350 400 450 500 550 Calculated Pressure (psig)
Moderator Temperature
(°F)
Axial Flaw - ART Values (170, 160)
Circ Flaw - ART Values (179, 159)
Acceptable Operation Unacceptable Operation I
r I
H I
I j
~
j
..-:. ~:
I
.-~....
ie:
- ~.:
I le ***
J J I
I ir J
~1
,,i I t~'
~ w
~
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 10-1 WCAP-18746-NP January 2023 Revision 2 10 REFERENCES
- 1.
U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988. [Agencywide Document Management System (ADAMS) Accession Number ML003740284]
- 2.
Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004. [ADAMS Accession Number ML050120209]
- 3.
Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.
- 4.
Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.
- 5.
RSICC Data Library Collection DLC-185, BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, July 1999.
- 6.
NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S.
Nuclear Regulatory Commission, October 2014. [ADAMS Accession Number ML14149A165]
- 7.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, Fracture Toughness Requirements, Revision 4, U.S. Nuclear Regulatory Commission, March 2019. [ADAMS Accession Number ML18338A516]
- 8.
ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Class 1 Components.
- 9.
Westinghouse Report WCAP-14780, Revision 3, Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, February 1998.
- 10. Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.
- 11. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.
[ADAMS Accession Number ML110070570]
- 12. ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results, 2018.
- 13. ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706 (ID), 1994.
- 14. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession Number ML010890301]
- 15. Westinghouse Report WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018. [ADAMS Accession Number ML18204A010]
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 10-2 WCAP-18746-NP January 2023 Revision 2
- 16. Pressurized Water Reactor Owners Group (PWROG) Report PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]
- 17. Westinghouse Report WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, November 2021.
- 18. Westinghouse Report WCAP-14613, Revision 2, Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, February 1998.
- 19. Westinghouse Report WCAP-18124-NP-A, Revision 0, Supplement 1-P, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, December 2020. [ADAMS Accession Number ML20344A388]
- 20. Westinghouse Report, WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, December 2022.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 A-1 WCAP-18746-NP January 2023 Revision 2 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)
Table A-1 and Table A-2 contain the thermal stress intensity factors (KIt) for the maximum heatup and cooldown rates at 54 EFPY for Prairie Island Units 1 and 2. The reactor vessel cylindrical shell radii to the 1/4T and 3/4T locations are as follows:
1/4T Radius = 67.869 inches 3/4T Radius = 71.215 inches
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 A-2 WCAP-18746-NP January 2023 Revision 2 Table A-1 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Heatup Curves Water Temp.
(F)
Vessel Temperature at 1/4T Location for 100F/hr Heatup (F) 1/4T Thermal Stress Intensity Factor (ksi in.)
Vessel Temperature at 3/4T Location for 100F/hr Heatup (F) 3/4T Thermal Stress Intensity Factor (ksi in.)
60 56.404
-0.958 55.128 0.520 65 59.579
-2.202 55.759 1.442 70 63.008
-3.125 57.148 2.208 75 66.730
-3.950 59.188 2.848 80 70.703
-4.586 61.743 3.369 85 74.816
-5.138 64.738 3.800 90 79.119
-5.574 68.083 4.154 95 83.508
-5.952 71.717 4.448 100 88.030
-6.253 75.584 4.691 105 92.606
-6.518 79.638 4.897 110 97.271
-6.730 83.845 5.068 115 101.971
-6.919 88.178 5.215 120 106.730
-7.072 92.613 5.338 125 111.513
-7.211 97.131 5.445 130 116.335
-7.324 101.719 5.536 135 121.173
-7.429 106.362 5.616 140 126.037
-7.515 111.050 5.686 145 130.912
-7.597 115.777 5.748 150 135.804
-7.666 120.534 5.803 155 140.704
-7.733 125.316 5.854 160 145.615
-7.790 130.119 5.899 165 150.532
-7.846 134.938 5.942 170 155.456
-7.896 139.772 5.981 175 160.384
-7.945 144.617 6.018 180 165.316
-7.989 149.471 6.052 185 170.252
-8.033 154.333 6.086 190 175.190
-8.074 159.201 6.117 195 180.131
-8.115 164.074 6.148 200 185.072
-8.153 168.951 6.178 205 190.017
-8.193 173.832 6.207 210 194.961
-8.229 178.716 6.236
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 A-3 WCAP-18746-NP January 2023 Revision 2 Table A-2 KIt Values for Prairie Island Units 1 and 2 at 54 EFPY 100F/hr Cooldown Curves Water Temp.
(F)
Vessel Temperature at 1/4T Location for 100F/hr Cooldown (F) 100F/hr Cooldown 1/4T Thermal Stress Intensity Factor (ksi in.)
210 226.075 8.848 205 221.021 8.811 200 215.968 8.774 195 210.914 8.737 190 205.861 8.701 185 200.808 8.664 180 195.755 8.627 175 190.702 8.591 170 185.649 8.554 165 180.596 8.518 160 175.543 8.482 155 170.491 8.445 150 165.438 8.409 145 160.386 8.373 140 155.334 8.337 135 150.281 8.301 130 145.229 8.265 125 140.177 8.229 120 135.126 8.194 115 130.074 8.158 110 125.022 8.123 105 119.970 8.087 100 114.919 8.052 95 109.867 8.017 90 104.816 7.982 85 99.765 7.947 80 94.714 7.912 75 89.663 7.877 70 84.612 7.842 65 79.561 7.807 60 74.512 7.772
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 B-1 WCAP-18746-NP January 2023 Revision 2 APPENDIX B OTHER RCPB FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [B-1], requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement (LST) for all RCPB components, which is specified in NB-3210 and NB-2332(b) of the Section III ASME Code [B-2], is the relevant requirement that would affect the P-T limits. This requirement is applicable to ferritic materials outside of the reactor vessel with a nominal wall thickness greater than 2 1/2 inches, such as piping, pumps and valves. The Prairie Island Units 1 and 2 reactor coolant system components do not contain ferritic materials in the Class 1 piping, pumps and valves. Therefore, the LST requirements of NB-2332(b) and NB-3210 are not applicable to the Prairie Island Units 1 and 2 P-T limits.
The other ferritic RCPB components that are not part of the reactor vessel beltline consist of the reactor vessel closure head, pressurizer, and steam generators.
RIS 2014-11 also addresses other ferritic components of the reactor coolant system relative to P-T limits, and states the following:
As specified in Sections I and IV.A of 10 CFR Part 50, Appendix G, ferritic RCPB components outside of the reactor vessel must meet the applicable requirements of ASME Code,Section III, Rules for Construction of Nuclear Facility Components.
The reactor vessel closure head flange materials have been considered in the development of P-T limits, see Section 6.3 of this report for further detail. Furthermore, the replacement reactor vessel closure heads were constructed to the 1998 Edition through 2000 Addenda of Section III of the ASME Code and has met all applicable requirements at the time of construction. The steam generators were constructed to the 1995 Edition through 1996 Addenda of Section III of the ASME Code and have met all applicable requirements.
The Unit 1 pressurizer was constructed to the 1965 Edition through 1966 Summer Addenda of Section III of the ASME Code, and the Unit 2 pressurizer was constructed to the 1965 Edition through 1966 Winter Addenda of Section III of the ASME Code. These Prairie Island Units 1 and 2 primary system components are analyzed to the identified ASME Code Section III Editions and met all applicable requirements at the time of construction. In addition, these components have not undergone neutron embrittlement. Therefore, no further consideration is necessary for these components with regard to P-T limits.
B.1 REFERENCES B-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.
B-2 ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Class 1 Components.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 C-1 WCAP-18746-NP January 2023 Revision 2 APPENDIX C UPPER-SHELF ENERGY EVALUATION Charpy upper-shelf energy (USE) is associated with the determination of acceptable reactor pressure vessel (RPV) toughness during the licensed operating period.
The requirements on USE are included in 10 CFR 50, Appendix G [C-1]. 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the USE of any RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.
There are two methods that can be used to predict the decrease in USE with irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide 1.99, Revision 2 [C-2].
For vessel materials that are not in the surveillance program or have non-credible data, the Charpy USE (Position 1.2) is assumed to decrease as a function of fluence and copper content, as indicated in Figure 2 of Regulatory Guide 1.99, Revision 2.
When two or more credible surveillance sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material. The surveillance data are then used in conjunction with the Regulatory Guide to predict the change in USE (Position 2.2) of the RPV material due to irradiation by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure C-1 and Figure C-2 of this report for Prairie Island Units 1 and 2, respectively) and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.
The 54 EFPY (EOLE) Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2.
The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection.
The projected USE values were calculated to determine if the Prairie Island Units 1 and 2 beltline materials remain above the 50 ft-lb criterion at 54 EFPY (EOLE). These calculations are summarized in Table C-1 and Table C-2 for Prairie Island Units 1 and 2, respectively. Note, even though some surveillance data for both units is deemed non-credible, it is still used in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2. The data from the surveillance materials are determined to be non-credible by Credibility Criterion 3. Credibility Criterion 3 indicates that even if the surveillance data are not considered credible for determination of RTNDT, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82.
USE Conclusion For Prairie Island Units 1 and 2, all of the beltline materials are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G) through EOLE (54 EFPY) with the exception of the Unit 1 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 without using surveillance data (Position 1.2). However, taking into account the surveillance data (Position 2.2), which should be used in preference to Position 1.2 per Regulatory Guide 1.99, Revision 2, the USE value for the Intermediate
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 C-2 WCAP-18746-NP January 2023 Revision 2 Shell to Lower Shell Circumferential Weld - Seam W3 is projected to remain above the 50 ft-lb screening criterion value through EOLE. Therefore, all of the Prairie Island Units 1 and 2 reactor vessel materials are projected to remain above the 10 CFR 50, Appendix USE screening criterion value of 50 ft-lb at EOLE (54 EFPY).
C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.
C-2 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 C-3 WCAP-18746-NP January 2023 Revision 2 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 1 Beltline Materials Reactor Vessel Material Wt %
Cu(a)
EOLE 1/4T Fluence(b)
(x 1019 n/cm2, E > 1.0 MeV)
Initial USE(a)
(ft-lb)
Projected USE Decrease
(%)
Projected EOLE USE (ft-lb)
Position 1.2(c)
Nozzle Shell Forging B 0.08 2.26 84 23 65 Intermediate Shell Forging C 0.07 3.77 143 26 106 Lower Shell Forging D 0.07 3.70 88 26 65 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 0.15 2.43 84 36 54 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 0.13 3.70 78.5 38 49 Position 2.2(d)
Intermediate Shell Forging C 3.77 143 14 123 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 3.70 78.5 9(e) 71 Notes:
(a)
Data taken from Table 3-1 of this report. If the base metal and weld Cu weight percentages are below the minimum value presented in Figure 2 of [C-2] (0.1 for base metal and 0.05 for welds), then the Cu weight percentages were conservatively rounded up to the minimum value.
(b) Values taken from Table 7-1. Fluence values above 1017 n/cm2 (E > 1.0 MeV) but below 2 x 1017 n/cm2 (E > 1.0 MeV) were rounded to 2 x 1017 n/cm2 (E > 1.0 MeV) when determining the % decrease because 2 x 1017 n/cm2 is the lowest fluence displayed in Figure 2 of Regulatory Guide 1.99, Revision 2 [C-2].
(c)
Percentage USE decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2 [C-2] and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide and using the material-specific Cu wt. % values. The percent-loss lines were extended into the low fluence area of Regulatory Guide 1.99, Revision 2, Figure 2, i.e., below 1018 n/cm2, in order to determine the USE % decrease as needed.
(d) Percentage USE decrease values are based on Position 2.2 of [C-2]. Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of [C-2]) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.
(e)
This Regulatory Guide 1.99, Revision 2, Position 2.2 USE %-decrease determination is based on a Capsule R Heat #
1752 surveillance weld %-decrease of 9%, which is based on an unirradiated USE of 82.5 ft-lb compared to the unirradiated USE for the reactor vessel Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat #
1752) unirradiated USE of 78.5 ft-lb. The 82.5 ft-lb value considers an additional data point with 97% shear. Whereas the reactor vessel weld unirradiated USE value of 78.5 ft-lb is consistent with FSAR Table 4.7-2 and only considers data with 100% shear. Each value is conservative for their respective use. A higher unirradiated USE value, when evaluating surveillance data, will result in a higher measured %-decrease, and a lower unirradiated USE value, when performing USE projection, will result in a lower USE projection.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 C-4 WCAP-18746-NP January 2023 Revision 2 Table C-2 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for the Prairie Island Unit 2 Beltline Materials Reactor Vessel Material Wt %
Cu(a)
EOLE 1/4T Fluence(b)
(x 1019 n/cm2, E > 1.0 MeV)
Initial USE(a)
(ft-lb)
Projected USE Decrease
(%)
Projected EOLE USE (ft-lb)
Position 1.2(c)
Upper Shell Forging B 0.07 2.26 85 23 65 Intermediate Shell Forging C 0.07 3.79 112 26 83 Lower Shell Forging D 0.08 3.73 108 26 80 Upper Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 1752) 0.13 2.42 78.5 33 53 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 2721) 0.09 3.73 103 32 70 Position 2.2(d)
Lower Shell Forging D 3.73 108 18 89 Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 2721) 3.73 103 12 91 Notes:
(a) Data taken from Table 3-2 of this report. If the base metal and weld Cu weight percentages are below the minimum value presented in Figure 2 of [C-2] (0.1 for base metal and 0.05 for welds), then the Cu weight percentages were conservatively rounded up to the minimum value.
(b) Values taken from Table 7-2. Fluence values above 1017 n/cm2 (E > 1.0 MeV) but below 2 x 1017 n/cm2 (E > 1.0 MeV) were rounded to 2 x 1017 n/cm2 (E > 1.0 MeV) when determining the % decrease because 2 x 1017 n/cm2 is the lowest fluence displayed in Figure 2 of Regulatory Guide 1.99, Revision 2 [C-2].
(c)
Percentage USE decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2 [C-2] and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide and using the material-specific Cu wt. % values. The percent-loss lines were extended into the low fluence area of Regulatory Guide 1.99, Revision 2, Figure 2, i.e., below 1018 n/cm2, in order to determine the USE % decrease as needed.
(d) Percentage USE decrease values are based on Position 2.2 of [C-2]. Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of [C-2]) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.
11----------------------------i
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 C-5 WCAP-18746-NP January 2023 Revision 2 Figure C-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 1 1.0 10.0 100.0 1.00E+17 1.00E+18 1.00E+19 1.00E+20 Percentage Drop in USE Neutron Fluence, n/cm2 (E > 1 MeV)
Surveillance Material: Intermediate Shell Forging C Surveillance Material: Weld Heat # 1752
% Copper Base Metal Weld 0.35 0.30 0.30 0.25 0.25 0.20 0.20 0.15 0.15 0.10 0.10 0.05 Upper Limit Forging Line Weld Line Limiting Forging Percent USE Decrease 16% from Capsule N (Axial Orientation)
Limiting Weld Percent USE Decrease 9% from Capsule R Nozzle Shell Forging B 1/4T Fluence = 2.26x1019 n/cm2 Nozzle Shell to Intermediate Shell Circumferential Weld Seam - W2 1/4T Fluence = 2.43x1019 n/cm2 Lower Shell Forging D and Intermediate Shell to Lower Shell Circumferential Weld Seam - W3 1/4T Fluence = 3.70x1019 n/cm2 Intermediate Shell Forging C 1/4T Fluence = 3.77x1019 n/cm2
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 C-6 WCAP-18746-NP January 2023 Revision 2 Figure C-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Prairie Island Unit 2 1100.0
% Cop pe;r Base Me,tal w *eld 0.35 o.30
_ 0.30
\\
0.25 0.25 t I 0.20 llimitin.g Forging Percent USE I limiting Weld Percent USE. I Decrease _ 13% trom Capsule T \\
Decrease 12% from Cansule R (Axial OnentatIonl I \\
I I
~-~ _ __,,....__..... _......... ___.___,,___..........,_1
_ 0.20
\\ t \\ 0.15 0.15 I \\ 0.10
- 0. 10 1
\\
I. 0.05 lntennediate Shell Forging c _-_.j..-----1-*f.-:...-:..."t.i+---c.:.I-1---1---1--+--+--+--,1
......_......_ _____ -+-------+-- 1i/4T Fluence = 3.7'9x1I019 n/cm 2':...__
-.__-+-_--+-.,.\\-1__ ~___,JHr ~--_--.....t_--_j_--_--....,t_--.....t...,+/--t-----.===+/-===+/-==::1::=~c::t::1::=+/-;--t-tir----t-11--_-+-"'---'_;114-1,1_,,1...-----1,_----1--1I
,_I I
Upper Snell Forgtng 1B I
\\ \\
I Upper limit r-i 1/4T Ffuence. 2-26x10 19 n/cm2\\
\\ \\
\\
=---
ll!~==------.1r, _--1---=-*-~--=----l--+----l----l---l-+~-i.---
1---,==----
~ =---1---1---,-+-----1-i-
-+_-::::1,...+~""':::i:-:..;_1,----+- ____ __,, _ _..... :._......... ___._,___.,,._.
~
_i,_._
~i.--
~
_ i.-i-"""".... -
1-----+---+----=-"-IF-+-+-+-+-l------.-.__ __.___.,____._.,__..__..__........ _......,. __ --lr- --~ _ __,,_....... _......... ___._,___..........,_,,
1...--""i---
I Upper Snell to lntem1ediaite Sh.ell
'=-...,e---
=--+---+--+----ll-+-+++-1---,----1 Ctrcumferenlial W eld Seam - W2 t="--::.....=--+- *. ---1----1--........ --1---1-....j.....,'I 1I/4T Fluence = 2.42x1I019 n/cm2
--....... NI Slurve11Iance Materi : Lowe;r Shell Forging D I
.i.. Survealance Materi : We d eat # 272 1 Lower Shell Forgtn.g D and Intermediate Shell to Lower Shell C trcumferenlial Weld Seam - W2 I
H4T Ruence = 3.73x1019 nlcm2 1.0 l-----L-__JL__JL...l..._J__j_j__L,_-1-__
___..!!:::,,,,:i::::,;,,;,,,::::::::,E:,::::::!,.,,,,,,,::::i:!::=i::!:,;i"""":::::i=i=:=!...,_--.L~- -L-1.L-L__JL_LJL.i-!
1.00E+H 1.00E3/4-18 1.00E+19 1.00E+20 Neutmn Fl uence, n/cm2 {E > 1 MeV)
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 D-1 WCAP-18746-NP January 2023 Revision 2 APPENDIX D PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION D.1 PRESSURIZED THERMAL SHOCK Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the reactor pressure vessel (RPV) under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high pressurization, significant degradation of vessel material toughness caused by radiation embrittlement, and the presence of a critical-size defect anywhere within the vessel wall.
In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [D-1]) that established screening criteria on Pressurized Water Reactor (PWR) vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS.
RTPTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license.
The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTPTS) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 [D-2].
These accepted methods were used with the clad/base metal interface fluence values of Section 2 to calculate the following RTPTS values for the Prairie Island Units 1 and 2 RPV materials at 54 EFPY (EOLE).
The EOLE RTPTS calculations are summarized in Table D-1 and Table D-2.
PTS Conclusion The Prairie Island Units 1 and 2 limiting RTPTS value for base metal at 54 EFPY is 146.3°F (see Tables D-1 and D-2), which corresponds to Unit 1 Intermediate Shell Forging C when considering non-credible surveillance data (using Position 2.1). The Prairie Island Units 1 and 2 limiting RTPTS value for circumferentially oriented welds at 54 EFPY is 183.9°F, which corresponds to the Unit 1 Intermediate to Lower Shell Circumferential Weld Seam W3, Heat # 1752 considering non-credible surveillance data (using Position 2.1).
Therefore, all materials in the Prairie Island Units 1 and 2 reactor vessels are below the RTPTS screening criteria of 270F for base metal and/or longitudinal welds, and 300F for circumferentially oriented welds through EOLE (54 EFPY).
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 D-2 WCAP-18746-NP January 2023 Revision 2 Table D-1 RTPTS Calculations for the Prairie Island Unit 1 Reactor Vessel Materials at 54 EFPY Reactor Vessel Material CF(a)
(°F)
Surface Fluence(b)
(x 1019 n/cm2, E > 1.0 MeV)
Surface FF(b)
RTNDT(U)(c)
(°F)
(°F)
U
(°F)
(d)
(°F)
Margin
(°F)
RTPTS
(°F)
Nozzle Shell Forging B 51.0 3.37 1.318
-4 67.2 0
17.0 34.0 97.2 Intermediate Shell Forging C 44.0 5.63 1.425 14 62.7 0
17.0 34.0 110.7 Using Non-credible Prairie Island Unit 1 Surveillance Data 69.0 5.63 1.425 14 98.3 0
17.0 34.0 146.3 Lower Shell Forging D 44.0 5.53 1.422
-4 62.6 0
17.0 34.0 92.6 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 2269) 79.5 3.63 1.335 0
106.1 17 28.0 65.5 171.6 Intermediate to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752) 69.7 5.53 1.422
-13 99.1 0
28.0 56.0 142.1 Using Non-credible Prairie Island Unit 1 Surveillance Data 99.1 5.53 1.422
-13 140.9 0
28.0 56.0 183.9 Notes:
(a) Data is from Table 5-3.
(b) Data is from Table 7-1.
(c) Data is from Table 3-1.
(d) The credibility conclusion for the surveillance material is discussed in Section 4. The intermediate shell forging material surveillance data and the weld Heat
- 1752 surveillance data were both determined to be non-credible. Per the guidance of 10 CFR 50.61 [D-1], the base metal = 17°F when surveillance data is non-credible or not used to determine the CF, and the weld metal = 28°F when surveillance data is not used to determine the CF and = 14°F when credible surveillance data is used to determine the CF. However, need not exceed 0.5*RTNDT per regulatory guidance in [D-1].
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 D-3 WCAP-18746-NP January 2023 Revision 2 Table D-2 RTPTS Calculations for the Prairie Island Unit 2 Reactor Vessel Materials at 54 EFPY Reactor Vessel Material CF(a)
(°F)
Surface Fluence(b)
(x 1019 n/cm2, E > 1.0 MeV)
Surface FF(b)
RTNDT(U)(c)
(°F)
(°F)
U
(°F)
(d)
(°F)
Margin
(°F)
RTPTS
(°F)
Upper Shell Forging B 44.0 3.37 1.318
-13 58.0 0
17.0 34.0 79.0 Intermediate Shell Forging C 44.0 5.66 1.426 14 62.7 0
17.0 34.0 110.7 Lower Shell Forging D 51.0 5.58 1.423
-4 72.6 0
17.0 34.0 102.6 Using Non-credible Prairie Island Unit 2 Surveillance Data 74.4 5.58 1.423
-4 105.9 0
17.0 34.0 135.9 Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 69.7 3.61 1.334
-13 93.0 0
28.0 56.0 136.0 Using Non-credible Prairie Island Unit 1 Surveillance Data (Heat # 1752) 101.4 3.61 1.334
-13 135.2 0
28.0 56.0 178.2 Intermediate to Lower Shell Circumferential Weld - Seam W3 51.6 5.58 1.423
-31 73.4 0
28.0 56.0 98.4 Using Credible Prairie Island Unit 2 Surveillance Data (Heat # 2721) 87.5 5.58 1.423
-31 124.5 0
14.0 28.0 121.5 Notes:
(a) Data is from Table 5-4.
(b) Data is from Table 7-2.
(c) Data is from Table 3-2.
(d) The credibility conclusion for the surveillance material is discussed in Section 4. The lower shell forging material surveillance data was determined to be non-credible, while the weld Heat # 2721 surveillance data was determined to be credible. Additionally, as discussed in Section 4, the weld Heat # 1752 surveillance data was determined to be non-credible. Per the guidance of 10 CFR 50.61 [D-1], the base metal = 17°F when surveillance data is non-credible or not used to determine the CF, and the weld metal = 28°F when surveillance data is not used to determine the CF and = 14°F when credible surveillance data is used to determine the CF. However, need not exceed 0.5*RTNDT per regulatory guidance in [D-1].
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 D-4 WCAP-18746-NP January 2023 Revision 2 D.2 EMERGENCY RESPONSE GUIDELINE LIMITS The Emergency Response Guideline (ERG) limits were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event [D-3]. Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.
The highest value of RTNDT for which the generic category ERG limits were developed is 250F for a longitudinal flaw and 300F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250F for a longitudinal flaw or 300F for a circumferential flaw, plant-specific ERG P-T limits must be developed.
The ERG category is determined by the magnitude of the limiting RTNDT value, which is calculated the same way as the RTPTS values are calculated in Section D.1 of this report. The material with the highest RTPTS defines the limiting material. Table D-3 and Table D-4 identify ERG category limits and the limiting material RTNDT values at 54 EFPY for Prairie Island Valley Units 1 and 2, respectively.
Table D-3 Evaluation of Prairie Island Unit 1 ERG Limit Category ERG Pressure-Temperature Limits [D-3]
Applicable RTNDT Value(a)
ERG P-T Limit Category RTNDT < 200F Category I 200F < RTNDT < 250F Category II 250F < RTNDT < 300F Category IIIb Limiting RTNDT Value(b)
Reactor Vessel Material RTNDT Value @ 54 EFPY Intermediate Shell to Lower Shell Circumferential Weld - Seam W3 (Heat # 1752)
Using Non-credible Surveillance Data 183.9 Notes:
(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.
(b) Value taken from Table D-1.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 D-5 WCAP-18746-NP January 2023 Revision 2 Table D-4 Evaluation of Prairie Island Unit 2 ERG Limit Category ERG Pressure-Temperature Limits [D-3]
Applicable RTNDT Value(a)
ERG P-T Limit Category RTNDT < 200F Category I 200F < RTNDT < 250F Category II 250F < RTNDT < 300F Category IIIb Limiting RTNDT Value(b)
Reactor Vessel Material RTNDT Value @ 54 EFPY Nozzle Shell to Intermediate Shell Circumferential Weld - Seam W2 (Heat # 1752)
Using Non-credible Surveillance Data 178.2 Notes:
(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.
(b) Value taken from Table D-2.
Per the ERG limit guidance document [D-3], some vessels do not change categories for operation through the end of license. However, when a vessel does change ERG categories between the beginning and end of operation, a plant-specific assessment must be performed to determine at what operating time the category changes. Thus, the ERG classification need not be changed until the operating cycle during which the maximum vessel value of actual or estimated real-time RTNDT exceeds the limit on its current ERG category.
Per Table D-3 and Table D-4, the limiting material for Prairie Island Units 1 and 2 have an RTNDT less than 200°F through 54 EFPY. Therefore, Prairie Island Units 1 and 2 remain in ERG Category I through EOLE (54 EFPY).
Conclusion of ERG P-T Limit Categorization As summarized above, Prairie Island Units 1 and 2 will remain in ERG Category I through EOLE (54 EFPY).
D.3 REFERENCES D-1 Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Federal Register, November 29, 2019.
D-2 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.
D-3 Westinghouse Owners Group Document HF04BG, Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 3, March 31, 2014.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-1 WCAP-18746-NP January 2023 Revision 2 APPENDIX E CREDIBILITY EVALUATION OF THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE PROGRAM E.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [E-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date there have been five surveillance capsules removed and tested from the Prairie Island Unit 2 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.
The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Prairie Island Unit 2 reactor vessel surveillance data and determine if that surveillance data is credible.
E.2 EVALUATION Criterion 1:
Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.
The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," [E-2] as follows:
"the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."
In addition, the Upper Shell Forging B and the Intermediate Shell Forging to Upper Shell Forging Weld Seam W2 will be considered to be part of the beltline region. Hence, the Prairie Island Unit 2 reactor vessel beltline region consists of the following materials:
- 1. Upper Shell Forging B (Heat # 22231/39088).
- 2. Intermediate Shell Forging C (Heat # 22829).
- 3. Lower Shell Forging D (Heat # 22642).
- 4. Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2 (Weld Wire Type UM40/Heat # 1752, Flux Type UM 89/Lot # 1263).
- 5. Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld Seam W3 (Weld Wire Type UM40/Heat # 2721, Flux Type UM 89/Lot # 1263).
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-2 WCAP-18746-NP January 2023 Revision 2 Per WCAP-8193, Revision 0 [E-3], the Prairie Island Unit 2 surveillance program was based on ASTM E185-70 [E-4], Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Per Section 3.1.2 of ASTM E185-70, A minimum test program shall consist of specimens taken from the following locations: (1) base metal of one heat, incorporated in the highest flux location of the reactor vessel, that has the highest initial ductile-brittle transition temperature, (2) weld metal fully representative of fabrication practice used for the welds in the highest flux location of the reactor vessel (weld wire or rod, and flux must come from one of the heats used in the highest flux region of the reactor vessel), and (3) the heat-affected zone of the weldments noted above.
It should be noted here that the Upper Shell Forging B and Weld Seam W2 were not considered when the surveillance program was developed. Therefore, at the time the Prairie Island Unit 2 surveillance capsule program was developed, Lower Shell Forging D was judged to be most limiting. This was based on the fact that both the Intermediate Shell Forging C and Lower Shell Forging D initial RTNDT values were within 2F of each other and Lower Shell Forging D had a lower USE value. As for the weld, the Prairie Island Unit 2 vessel has only one weld in the highest flux region (Weld Seam W3, Weld Control # PS-011, Type UM40, Heat # 2721, Flux Type UM89, Flux Lot No. 1263). The same weld was used in the surveillance program.
Therefore, the materials selected for use in the Prairie Island Unit 2 surveillance program were those judged to be most likely limiting with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was developed.
Based on the discussion, Criterion 1 is met for the Prairie Island Unit 2 surveillance program.
Criterion 2:
Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.
Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions are presented in Section 5 and Appendix C of the latest surveillance capsule report, WCAP-14613 [E-5].
Based on engineering judgment, the scatter in the data presented in these plots, as documented in Appendix A of this calculation note, is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Prairie Island Unit 2 surveillance materials unambiguously.
Hence, Criterion 2 is met for the Prairie Island Unit 2 surveillance program.
Criterion 3:
When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values.
Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [E-6].
The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these RTNDT values about this line is less than 28°F for welds and less than 17°F for the forging.
Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [E-7]. At this meeting the NRC presented five cases. Of the five cases, Case 1
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-3 WCAP-18746-NP January 2023 Revision 2 (Surveillance data available from plant but no other source) most closely represents the situation listed above for Prairie Island Unit 2 surveillance weld metal and forging materials. It is noted that the Unit 2 Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld Seam W2 utilizes surveillance data for Heat # 1752 from the Prairie Island Unit 1 surveillance program, consistent with Case 5 (Surveillance data from other sources only); however, creditability of this data was evaluated in WCAP-18660-NP [E-8].
Table E-1 contains the calculation of chemistry factors for the Prairie Island Unit 2 reactor vessel beltline materials contained in the surveillance program.
Table E-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Prairie Island Unit 2 Surveillance Data Material Capsule Capsule Fluence(a)
(x 1019 n/cm2, E > 1.0 MeV)
FF(b)
RTNDT(c)
(°F)
FF*RTNDT
(°F)
FF(b)
Lower Shell Forging D (Axial)
V 0.598 0.856 35.0 30.0 0.733 T
1.10 1.027 27.9 28.6 1.054 R
4.11 1.362 84.3 114.8 1.855 P
4.27 1.370 103.5 141.8 1.877 N
8.41 1.491 154.1 229.7 2.222 Lower Shell Forging D (Tangential)
V 0.598 0.856 33.8 28.9 0.733 T
1.10 1.027 54.4 55.8 1.054 R
4.11 1.362 89.6 122.0 1.855 P
4.27 1.370 99.6 136.5 1.877 N
8.41 1.491 176.5 263.1 2.222 SUM:
V 0.598 0.856 69.3 59.3 0.733 T
1.10 1.027 57.7 59.2 1.054 R
4.11 1.362 100.3 136.6 1.855 P
4.27 1.370 96.2 131.8 1.877 N
8.41 1.491 135.6 202.2 2.222 SUM:
589.1 7.742 CF Surv. Weld = (FF
- RTNDT) ÷ (FF2) = (589.1) ÷ (7.742) = 76.1°F Notes:
(a) Fluence taken from Table 2-10.
(b) FF = fluence factor = f(0.28 - 0.10*log (f)).
(c) RTNDT values do not include the adjustment ratio procedure of Reg. Guide 1.99 Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only Prairie Island Unit 2 reactor vessel is being considered; therefore, no temperature adjustment is required.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-4 WCAP-18746-NP January 2023 Revision 2 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table E-2.
Table E-2 Best-Fit Evaluation for Prairie Island Unit 2 Surveillance Materials Material Capsule CF(a)
(Slopebest-fit)
(°F)
Capsule Fluence(b)
(x 1019 n/cm2)
FF(c)
Measured RTNDT(d)
(°F)
Predicted RTNDT(e)
(°F)
Scatter RTNDT(f)
(°F)
<17°F (Base Metal)
<28°F (Weld)
Lower Shell Forging D (Axial)
V 74.4 0.598 0.856 35.0 63.7 28.7 No T
1.10 1.027 27.9 76.3 48.4 No R
4.11 1.362 84.3 101.3 17.0 No P
4.27 1.370 103.5 101.9 1.6 Yes N
8.41 1.491 154.1 110.9 43.2 No Lower Shell Forging D (Tangential)
V 0.598 0.856 33.8 63.7 29.9 No T
1.10 1.027 54.4 76.4 22.0 Yes R
4.11 1.362 89.6 101.3 11.7 Yes P
4.27 1.370 99.6 101.9 2.3 Yes N
8.41 1.491 176.5 110.9 65.6 No Surveillance Weld Metal (Heat # 2721)
V 76.1 0.598 0.856 69.3 65.1 4.2 Yes T
1.10 1.027 57.7 78.1 20.4 Yes R
4.11 1.362 100.3 103.7 3.4 Yes P
4.27 1.370 96.2 104.3 8.1 Yes N
8.41 1.491 135.6 113.4 22.2 Yes Notes:
(a) CF calculated in Table E-1.
(b) Fluence taken from Table 2-10.
(c) FF = fluence factor = f(0.28 - 0.10*log (f)).
(d) Measured RTNDT taken from Table E-1.
(e) Predicted RTNDT = CF x FF.
(f) Scatter RTNDT = Absolute Value [Predicted RTNDT - Adjusted RTNDT].
The scatter of RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 17°F for base metal and 28°F for weld metal. From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table E-2 indicates that four of the ten surveillance data points fall inside the +/- 1 of 17°F scatter band for surveillance base metals, which is 40% of the data (4/10 x 100). In addition, Table E-2 indicates that zero of the five surveillance data points fall outside the +/- 1 of 28°F scatter band for surveillance weld materials.
Therefore, the Lower Shell Forging D is deemed not-credible, while the Surveillance weld is deemed credible. Although Lower Shell Forging D did not meet Criterion 3, both materials may still be used in determining the upper-shelf energy decrease in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2.
Criterion 4:
The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-5 WCAP-18746-NP January 2023 Revision 2 The Prairie Island Unit 2 capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25F.
The weld metal used in Weld Seam W2 is contained in the Unit 1 surveillance program. Both Prairie Island Unit 1 and Unit 2 operate at the same temperature. Hence, the Unit 1 surveillance program weld metal was irradiated at a temperature within +/- 25F of weld seam W2 in Unit 2. Therefore, the Unit 1 surveillance weld will be used to project the fracture toughness properties of the Unit 2 Weld Seam W2.
Hence, Criterion 4 is met for the Prairie Island Unit 2 surveillance program.
Criterion 5:
The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.
The Prairie Island Unit 2 surveillance program does contain correlation monitor material, which was supplied by Oak Ridge National Laboratory from plate material used in the AEC-Sponsored Heavy Section Steel Technology (HSST) Program. This material was obtained from a 12-inch-thick A533, Grade B, Class 1 plate (HSST Plate 02), which was provided to Subcommittee II of ASTM Committee E10 on Radioisotopes and Radiation Effects to serve as correlation monitor material in reactor vessel surveillance programs. The plate was produced by the Lukens Steel Company and heat treated by Combustion Engineering, Inc.
This criterion was met per the surveillance capsule fluence values evaluated in WCAP-18795-NP [E-10].
NUREG/CR-6413, ORNL/TM-13133 [E-9], contains a plot of Residual vs. Fast Fluence for the HSST01 and HSST02 correlation monitor material (Figure 11 of the report). The figure shows a 2 uncertainty of 50°F. The data used for this plot is contained in Tables 13, 14, and 15 (in the NUREG Report). However, the data in the NUREG report does not consider the recalculated fluence values documented herein. Thus, Table E-3 below presents an updated calculation of Residual vs. Fast Fluence for Prairie Island Unit 2.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-6 WCAP-18746-NP January 2023 Revision 2 Table E-3 Calculation of Residual vs. Fast Fluence for Prairie Island Unit 2 Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV)
FF Measured Shift(a)
(°F)
RG 1.99 Shift(b)
(°F)
Residual(c)
V 0.598 0.856 123.5 109.6 13.9 T
1.10 1.027 158.0 131.4 26.6 R
4.11 1.362 183.2 174.3 8.9 P
4.27 1.370 196.8 175.4 21.4 N
8.41 1.491 212.9 190.8 22.1 Notes:
(a) Measured T30 values for the correlation monitor material were taken from Table B-17 of WCAP-14613 [E-5] for HSST02 material.
(b) Per NUREG/CR-6413, ORNL/TM-13133 [E-9], the Cu and Ni values for the correlation monitor material (HSST Plate 02) are 0.17 and 0.64, respectively. This equates to a chemistry factor value of 128F based on Regulatory Guide 1.99, Revision 2, Position 1.1. The calculated shift is thus equal to CF
- FF.
(c) Residual = Measured Shift - RG 1.99 Shift.
Table E-3 shows a 2 uncertainty of less than 50F, which is the allowable scatter in NUREG/CR-6413, ORNL/TM-13133.
Hence, Criterion 5 is met for the Prairie Island Unit 2 surveillance program.
E.3 CONCLUSION Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Prairie Island Unit 2 surveillance forging material is deemed non-credible and the Prairie Island Unit 2 surveillance weld material is deemed credible.
- This record was final approved on 1/12/2023, 11:29:55 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 E-7 WCAP-18746-NP January 2023 Revision 2 E.4 REFERENCES E-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.
[ADAMS Accession Number ML003740284]
E-2 Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S.
Nuclear Regulatory Commission, Federal Register, November 29, 2019.
E-3 Westinghouse Report WCAP-8193, Revision 0, Northern States Power Co. Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program, September 1973.
E-4 ASTM E185-70, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, American Society of Testing and Materials, Philadelphia, PA, 1970.
E-5 Westinghouse Report WCAP-14613, Revision 2, Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, February 1998.
E-6 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society of Testing and Materials, Philadelphia, PA, 1982.
E-7 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC Presentation, Generic Letter 92-01 and RPV Integrity Assessment, Status, Schedule, and Issues, NRC/Industry Workshop on RPV Integrity Issues, February 1998. [ADAMS Accession Number ML110070570]
E-8 Westinghouse Report WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, November 2021.
E-9 NUREG/CR-6413; ORNL/TM-13133, Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials, J. A. Wang, Oak Ridge National Laboratory, Oak Ridge, TN, April 1996. [ADAMS Accession Number ML20112B397].
E-10 Westinghouse Report, WCAP-18795-NP, Revision 0-A, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, December 2022.