RA-23-0245, License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits

License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits

ML24170A696
Person / Time
Site:	Catawba
Issue date:	06/18/2024
From:	Flippin N Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-23-0245
Download: ML24170A696 (1)

Text

{{#Wiki_filter:Nicole L. Flippin Vice President Catawba Nuclear Station Duke Energy 4800 Concord Rd York, SC 29745 803-701-3340 Nicole.Flippin@duke-energy.com Serial: RA-23-0245 10 CFR 50.90 June 18, 2024 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Catawba Nuclear Station, Unit Nos. 1 and 2 Docket Nos. 50-413, 50-414 Renewed License Nos. NPF-35 and NPF-52

SUBJECT:

LICENSE AMENDMENT REQUEST PROPOSING TO REVISE TECHNICAL SPECIFICATION 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Ladies and Gentlemen: Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) hereby requests an amendment to the Catawba Nuclear Station (CNS) Unit 2 Technical Specifications (TS). The proposed change will revise CNS TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect an update to the P/T limit curves in Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY). The proposed change will also reflect that the revised CNS Unit 2 P/T limit curves in TS 3.4.3 are applicable until 54 effective full power years EFPY. The proposed change is necessary because the existing CNS Unit 2 P/T limit curves in TS 3.4.3 are only applicable up to 34 EFPY. CNS Unit 2 is expected to reach 34 EFPY in August 2025. Note that although the proposed change only impacts CNS Unit 2, the amendment request is being docketed for both CNS Units 1 and 2 since the TS are common to both units. The Enclosure provides a description and assessment of the proposed change. Attachment 1 provides the existing CNS TS pages marked up to show the proposed change. Attachment 2 provides a non-proprietary version of Westinghouse report WCAP-18843-NP, Revision 0, Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation. provides a non-proprietary version of Westinghouse report WCAP-15285, Revision 0, Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation using Code Case N-640. The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no significant hazards consideration. The bases for these determinations are included in the Enclosure. Duke Energy requests NRC approval of the proposed amendment by May of 2025, in order to allow implementation prior to the CNS Unit 2 P/T limit curves expiring in August of 2025. ( ~ DUKE ENERGY

U.S. Nuclear Regulatory Commission RA-23-0245 Page 2 There are no newregulatory commitments contained in this submittal. In accordance with 10 CFR 50.91, Duke Energy is notifying the State of South Carolina of this license amendment request by transmitting a copy.of this letter and enclosure to the designated State Official. Should you have any question concerning this letter ahd its enclosure, please contact Ryan Treadway, Director-Nuclear Fleet Licensing at (980) 373-5873, I declare under penalty of perjury that the foregoing is true and correct. Executed on June 18, 2024. Sincerely, Nicole L. Flippin Vice President, Catawba Nuclear Station

Enclosure:

Description and Assessment of the Proposed Change Attachments:

1. Technical Specifications Markup

2. Westinghouse Report WCAP-18843-NP, Revision 0, i'Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation" 3.

Westinghouse Report WCAP-15285, Revision 0, "Catawba Unit 2 Heatup and Cooldown Limit Curves fot Normal Operation using Code Case N-640;"

U.S. Nuclear Regulatory Commission RA-23-0245 Page 3 cc: L. Dudes, USNRC, Region II Regional Administrator S. Williams, USNRC NRR Project Manager for CNS D. Rivard, USNRC Senior Resident Inspector for CNS

U.S. Nuclear Regulatory Commission RA-23-0245 Page 1 ENCLOSURE Description and Assessment of the Proposed Change

Subject:

License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3.

TECHNICAL EVALUATION

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS:

1. Technical Specifications Markup

2. Westinghouse Report WCAP-18843-NP, Revision 0, Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

3. Westinghouse Report WCAP-15285, Revision 0, Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation using Code Case N-640.

U.S. Nuclear Regulatory Commission RA-23-0245 Page 2

1.

SUMMARY

DESCRIPTION Duke Energy Carolinas, LLC (Duke Energy) hereby request an amendment to the Catawba Nuclear Station (CNS) Unit 2 Technical Specifications (TS). The proposed change will revise CNS TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY). The proposed change will also reflect that the revised CNS Unit 2 P/T limit curves in TS 3.4.3 are applicable until 54 effective full power years (EFPY).

2.

DETAILED DESCRIPTION 2.1 System Design and Operation All components of the CNS Reactor Coolant System (RCS) are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients and reactor trips. CNS is required to limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation. The CNS TS contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The typical use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. Operating limits are established that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). CNS Updated Final Safety Analysis Report (UFSAR) Section 5.3.2 provides additional details regarding the methodology used to develop the P/T limit curves that are contained in the CNS TS. 2.2 Current Technical Specifications Requirements CNS Limiting Condition for Operation (LCO) 3.4.3 requires that RCS pressure, RCS temperature and RCS heatup and cooldown rates be maintained within the limits specified in Figures 3.4.3-1 and 3.4.3-2 (P/T curves) that are associated with this proposed change. The two elements of LCO 3.4.3 are:

• The limit curves for heatup, cooldown and ISLH testing; and Limits on the rate of change of temperature.

LCO 3.4.3 limits apply to all components of the RCS, except the pressurizer and define allowable operating regions and permit a large number of operating cycles while providing a wide margin to non-ductile failure. Violating LCO 3.4.3 limits would result in placing the reactor vessel outside of the bounds of the stress analyses and could increase stresses in other RCPB components.

U.S. Nuclear Regulatory Commission RA-23-0245 Page 3 2.3 Reason for the Proposed Change The proposed change is necessary because the existing CNS Unit 2 P/T limit curves in TS 3.4.3 are only applicable up to 34 EFPY. CNS Unit 2 is expected to reach 34 EFPY in August of 2025. A new set of P/T limit curves with a longer term of applicability is required. 2.4 Description of the Propose Change TS 3.4.3, Figure 3.4.3-1, (UNIT 2 ONLY) RCS Heatup Limitations is revised as follows: The existing RCS heatup limitations curve is superseded entirely by a new curve applicable up to 54 EFPY. The words Limiting ART at 34 EFPY are replaced with Limiting ART at 54 EFPY. The 1/4-T value of 121°F is revised to state 126°F. The 3/4-T value of 106°F is revised to state 112°F. TS 3.4.3, Figure 3.4.3-2, (UNIT 2 ONLY) RCS Cooldown Limitations is revised as follows: The existing RCS cooldown limitations curve is superseded entirely by a new curve applicable up to 54 EFPY. The words Limiting ART at 34 EFPY are replaced with Limiting ART at 54 EFPY. The 1/4-T value of 121°F is revised to state 126°F. The 3/4-T value of 106°F is revised to state 112°F. The TS markup provided in Attachment 1 reflects the proposed change described above.

3.

TECHNICAL EVALUATION Pressure/Temperature Limit Curves Development and Applicability Term Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility transition temperature) corresponding to the limiting material in the beltline region of the reactor pressure vessel (RPV). The most limiting RTNDT of the material in the beltline region of the RPV was determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced shift ( RTNDT). RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactors life, RTNDT due to the radiation exposure associated with that time period was added to the original unirradiated RTNDT. Using the

U.S. Nuclear Regulatory Commission RA-23-0245 Page 4 Adjusted Reference Temperature (ART) values, P/T limit curves were determined in accordance with the requirements of 10 CFR 50, Appendix G. The existing P/T limit curves in CNS TS 3.4.3 for Unit 2 were originally developed in Westinghouse report WCAP-15285 (Attachment 3) and are valid through 34 EFPY. However, additional P/T limit curves were also originally developed in WCAP-15285 with an applicability term of 51 EFPY. The 51 EFPY P/T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material. 10 CFR 50, Appendix G requires that P/T limits be developed to bound all ferritic materials in the RPV. Regulatory Issue Summary (RIS) 2014-11 (Reference 1) clarifies that P/T limit calculations for ferritic RPV materials other than those with the highest reference temperature may define P/T limit curves that are more limiting because the consideration of stress levels from structural discontinuities (such as RPV inlet and outlet nozzles) may produce a lower allowable pressure. RIS 2014-11 specifically focused on those portions of the vessel known as the extended beltline, which are materials outside of the traditional RPV beltline which are predicted to exceed a fluence level of 1.0 x 1017 n/cm2. Since the 51 EFPY P/T limit curves from WCAP-15285 were originally developed prior to the issuance of RIS 2014-11, a re-evaluation of the 51 EFPY P/T limit curves was necessary. It is also important to note that since WCAP-15285 was developed, an end-of-life extension (EOLE) term of 54 EFPY was established for CNS rather than the 51 EFPY term considered within WCAP-15285. Therefore, an updated fluence analysis for the RPV was performed in Westinghouse report WCAP-18843-NP (Attachment 2) and new 54 EFPY ART values were calculated considering both the traditional beltline and extended beltline materials. The updated fluence analysis determined that the RPV inlet and outlet nozzles will not be limiting with respect to the P/T limit curves based on applicability to topical report PWROG-15109-NP-A (Reference 2). PWROG-15109-NP-A was developed to generically address RPV nozzle P/T limit curves for the U.S. pressurized water reactor (PWR) fleet. The evaluation within PWROG-15109-NP-A demonstrated that a plants P/T limit curves developed with NRC-approved methods bound the generic nozzle P/T limit curves as long as the plant-specific fluence of the RPV nozzle corners remain less than the screening criterion of 4.28 x 1017 n/cm2. PWROG-15109-NP-A was approved by the NRC (Reference 3) as an acceptable means to address the concerns of RIS 2014-11 for RPV nozzles. Since WCAP-18843-NP determined the RPV nozzle fluence at 54 EFPY remains below the screening criterion of 4.28 x 1017 n/cm2, the results and conclusions of PWROG-15109-NP-A apply to CNS Unit 2 and demonstrate the nozzles will not be limiting with respect to the P/T limit curves. To determine whether the 51 EFPY P/T limit curves developed in WCAP-15285 are applicable through 54 EFPY considering the new fluence values documented in WCAP-18843-NP, a comparison of the 54 EFPY limiting 1/4T and 3/4T ART values determined within WCAP-18843-NP was made with the limiting 1/4T and 3/4T ART values for the 51 EFPY P/T limit curves in WCAP-15285. Based on this comparison, it was determined the 54 EFPY limiting 1/4T and 3/4T ART values do not exceed the limiting 1/4T and 3/4T ART values for the 51 EFPY P/T limit curves in WCAP-15285. Therefore, the limiting ART values from the 51 EFPY P/T limit curves originally generated in WCAP-15285 remain conservative and bounding through 54 EFPY and are therefore maintained in TS Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY).

U.S. Nuclear Regulatory Commission RA-23-0245 Page 5 Low Temperature Overpressure (LTOP) Setpoint Considerations The effects of the newly revised P-T curves for Catawba Unit 2 were analyzed to verify the existing Limiting Conditions for Operation in TS 3.4.12 remain valid with no changes required. Inspection of the Unit 2 P-T Curves published in TS 3.4.3 to the proposed curves developed in WCAP-18843 show that the allowable pressures across the analyzed temperature range have been reduced through 54 EFPY. Comparison of the previously analyzed peak pressures for various LTOP events to the new P-T limit curves was performed. It was determined that the existing setpoints for the LTOP Pressure Relief Devices (Pressurizer Power Operated Relief Valves [PORVs] and Residual Heat Removal [RHR] pump suction relief valves) and Reactor Coolant Pump (RCP) operating restrictions as currently described in TS 3.4.12 remain adequate to ensure RCS pressure will not exceed limiting pressures illustrated in the new P-T limit curves through 54 EFPY. Implementation of the new P-T limit curves do not require revision to TS 3.4.12 to ensure LTOP operation can safely prevent overpressure events.

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements and guidance documents are applicable to the proposed change. 10 CFR 50.36 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, establishes the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36(c) TSs will include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) LCOs, (3) Surveillance Requirements (SRs), (4) design features; and (5) administrative controls. CNS LCO 3.4.3 limits the pressure and temperature changes during RCS heatup and cooldown (i.e., to the right and below the P/T curves in Figures 3.4.3-1 and 3.4.3-2), to prevent non-ductile RPV failure. The proposed change revises the CNS Unit 2 P/T limit curves in TS 3.4.3 and reflects that the curves are applicable until 54 EFPY. Based on the determination that the proposed TS Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY) are acceptable up to 54 EFPY (see Section 3 above), Duke Energy concludes that CNS LCO 3.4.3 will continue to meet 10 CFR 50.36(c)(2)(i) with the proposed change. 10 CFR 50.60 Section 50.60 of 10 CFR, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, imposes fracture toughness and material surveillance program requirements, which are set forth in 10 CFR 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements. With the proposed change, CNS meets the requirements set forth in 10 CFR 50 appendices G and H. Therefore, CNS also satisfies the requirements of 10 CFR 50.60 for the proposed change.

U.S. Nuclear Regulatory Commission RA-23-0245 Page 6 10 CFR 50, Appendix G Appendix G to 10 CFR 50 requires that the P/T limits for the facilitys reactor pressure vessel (RPV) be at least as conservative as those obtained by following the linear elastic fracture mechanics methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Using the ART values, P/T limit curves were determined in accordance with the requirements of 10 CFR 50, Appendix G. Therefore, Duke Energy concludes for the proposed change that the CNS RPV will continue to meet RPV integrity regulatory requirements through 54 EFPY. 10 CFR 50, Appendix H Appendix H to 10 CFR 50 establishes requirements for each facility related to its RPV material surveillance. The proposed change does not impact the CNS surveillance capsule removal schedule. Regulatory Guide (RG) 1.99, Revision 2 RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, contains guidance on methodologies the NRC considers acceptable for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. This RG was used along with the surface fluence in Section 2 of WCAP-18843-NP to calculate ART values for the CNS Unit 2 reactor vessel materials at 54 EFPY. As previously noted in Section 3 of this LAR above, the 54 EFPY ART values (as opposed to 51 EFPY values) were calculated to correspond with the applicable EOLE term for CNS. Therefore, the proposed change has no effect on how Duke Energy applies RG 1.99, Revision 2 for CNS Unit 2. RG 1.190 RG. 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001, describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence. As noted in WCAP-18843-NP, the neutron transport evaluation methodologies utilized for the CNS Unit 2 neutron fluence evaluation follow the guidance of RG 1.190. Regulatory Issue Summary (RIS) 2014-11 RIS 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may result in more limiting P/T curves because of higher stresses due to structural discontinuities, such as those in RPV inlet and outlet nozzles. Duke Energy appropriately considered RIS 2014-11 for the proposed change in the Technical Evaluation section of the LAR above (Section 3). The proposed change does not affect plant compliance with any of the above regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

U.S. Nuclear Regulatory Commission RA-23-0245 Page 7 4.2 Precedent The NRC has previously approved changes similar to the proposed change in this License Amendment Request for other nuclear power plants including: Catawba Nuclear Station, Unit No. 1: Application dated July 2, 2019 (ADAMS Accession No. ML19183A038); NRC Safety Evaluation dated August 4, 2020 (ADAMS Accession No. ML20174A045). H.B. Robinson Steam Electric Plant, Unit No. 2: Application dated November 2, 2015 (ADAMS Accession No. ML15307A069); NRC Safety Evaluation dated November 22, 2016 (ADAMS Accession No. ML16285A404). Browns Ferry Nuclear Plant, Unit 2: Application dated June 19, 2014 (ADAMS Accession No. ML14175A307); NRC Safety Evaluation dated June 2, 2015 (ADAMS Accession No. ML15065A049). 4.2 No Significant Hazards Consideration Determination Analysis Duke Energy requests an amendment to the Catawba Nuclear Station (CNS) Unit 2 Renewed Facility Operating License. The proposed change will revise CNS Technical Specification (TS) 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY). The proposed change will also reflect that the revised CNS Unit 2 P/T limit curves in TS 3.4.3 are applicable until 54 effective full power years (EFPY). Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises TS 3.4.3 to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY) that are applicable until 54 EFPY. The proposed change does not involve physical changes to the plant or alter the reactor coolant system (RCS) pressure boundary (i.e., there are no changes in operating pressure, materials or seismic loading). The proposed P/T limit curves and Adjusted Reference Temperature (ART) values for TS 3.4.3 with an applicability term of 54 EFPY provide continued assurance that the fracture toughness of the reactor pressure vessel (RPV) is consistent with analysis assumptions and NRC regulations. The methodology used to develop the proposed P/T limit curves provides assurance that the probability of a rapidly propagating failure will be minimized. The proposed P/T limit curves, with the applicability term of 54 EFPY, will continue to prohibit operation in regions where it is possible for brittle fracture of reactor vessel materials to occur, thereby assuring that the integrity of the RCS pressure boundary is maintained.

U.S. Nuclear Regulatory Commission RA-23-0245 Page 8 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change revises TS 3.4.3 to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY) that are applicable until 54 EFPY. The proposed change does not affect the design or assumed accident performance of any structure, system or component or introduce any new modes of system operation or failure modes. Compliance with the proposed P/T limit curves will provide sufficient protection against brittle fracture of reactor vessel materials to assure that the RCS pressure boundary performs as previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change revises TS 3.4.3 to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 2 ONLY) and 3.4.3-2 (UNIT 2 ONLY) that are applicable until 54 EFPY. CNS complies with applicable regulations (i.e., 10 CFR 50, Appendices G and H) and adheres to Nuclear Regulatory Commission (NRC)-approved methodologies (i.e., Regulatory Guides 1.99 and 1.190) with respect to the proposed P/T limit curves in TS 3.4.3 in order to provide an adequate margin of safety to the conditions at which brittle fracture may occur. The proposed P/T limit curves for CNS Unit 2, with an applicability term of 54 EFPY, will continue to provide assurance that the P/T limits are not exceeded. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified. 4.4 Conclusions In conclusion, based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commissions regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

U.S. Nuclear Regulatory Commission RA-23-0245 Page 9

5.

ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.

REFERENCES

1. 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 No. ML14149A165).

2. Pressurized Water Reactor Owners Group (PWROG) PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020 (ADAMS Accession No. ML20024E573).

3. NRC Safety Evaluation Final Safety Evaluation by the Office of Nuclear Reactor Owners Group Topical Report PWROG-15109-NP, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation EPID L-2018-TOP-0009, October 31, 2019 (ADAMS Accession Nos. ML19301D063 & ML19301D160).

to RA-23-0245 Technical Specifications Markup

RCS P/T Limits 3.4.3 Catawba Units 1 and 2 3.4.3-4 Amendment Nos. 212/206 Figure 3.4.3-1 (UNIT 2 ONLY) RCS Heatup Limitations 0 500 1000 1500 2000 2500 0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F) Indicated Pressure (psig) MATERIALS PROPERTY BASIS Limiting Material: Intermediate Shell, B8605-2 Limiting ART at 34 EFPY: 1/4-T, 121°F 3/4-T, 106°F Leak Test Limit Unacceptable Operation Heatup Rate Up To 60 °F/hr Acceptable Operation Closure Head & Vessel Flange Limit Criticality Limits for Service Period Up To 34 EFPY REPLACE WITH INSERT 1 ~:1 I I I I I ~ I '\\ I I '\\. I I I ~~ ~ I ~~ ~~ ~~ I I ~~ I " ' '\\ I '\\. I I ~~ ~~ I I ~~ ~ D ii::°llli / ~~ I J I I'\\ / J I/ I " I / ' '\\. '\\. 1.. 1'.. ' \\ I'\\ I\\. ' '\\.

RCS P/T Limits 3.4.3 Catawba Units 1 and 2 3.4.3-4 [NEXT AMENDMENT NUMBER] 0 500 1000 1500 2000 2500 0 50 100 150 200 250 300 350 400 450 500 PRESSURE (PSIG) TEMPERATURE (°F) Figure 3.4.3-1 (UNIT 2 ONLY) RCS Heatup Limitations (Without Margins for Instrument Errors) Leak Test Limit Unacceptable Operation Acceptable Operation Heatup Rate up to 60 °F/hour Criticality Limits for Service Period up to 54 EFPY Closure Head and Vessel Flange Limit MATERIALS PROPERTY BASIS Limiting Material: Intermediate Shell, B8605-2 Limiting ART at 54 EFPY: 1/4-T, 126°F 3/4-T, 112 °F INSERT 1 r T T I,.____I______, I I I t 1 1

RCS P/T Limits 3.4.3 Catawba Units 1 and 2 3.4.3-6 Amendment Nos. 212/206 Figure 3.4.3-2 (UNIT 2 ONLY) RCS Cooldown Limitations 0 500 1000 1500 2000 2500 0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F) Indicated Pressure (psig) MATERIALS PROPERTY BASIS Limiting Material: Intermediate Shell, B8605-2 Limiting ART at 34 EFPY: 1/4-T, 121°F 3/4-T, 106°F Acceptable Operation Unacceptable Operation Cooldown Rates (°F/hr) 0, 20, 40, 60, & 100 Closure Head & Vessel Flange Limit REPLACE WITH INSERT 2 I I I\\. ' 1, I ~- \\ ~- ~- I ' " I\\. I \\. '\\ ' '\\ I J \\: I\\. '\\ I\\. / \\ \\. \\ "" \\. '\\ '\\

RCS P/T Limits 3.4.3 Catawba Units 1 and 2 3.4.3-6 [NEXT AMENDMENT NUMBER] 0 500 1000 1500 2000 2500 0 50 100 150 200 250 300 PRESSURE (PSIG) TEMPERATURE (°F) Unacceptable Operation Acceptable Operation Closure Head and Vessel Flange Limit Cooldown Rate up to 100°F/hour Figure 3.4.3-2 (UNIT 2 ONLY) RCS Cooldown Limitations (Without Margins for Instrument Errors) MATERIALS PROPERTY BASIS Limiting Material: Intermediate Shell, B8605-2 Limiting ART at 54 EFPY: 1/4-T, 126°F 3/4-T, 112 °F INSERT 2 T T to RA-23-0245 Westinghouse Report WCAP-18843-NP, Revision 0, Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

Westinghouse Non-Proprietary Class 3 WCAP-18843-NP August 2023 Revision 0 Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

@Westinghouse

Westinghouse Non-Proprietary Class 3

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA © 2023 Westinghouse Electric Company LLC All Rights Reserved WCAP-18843-NP Revision 0 Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation Tyler C. Ziegler* RV/CV Design and Analysis Arturo Mirallas Ferrete* Nuclear Operations August 2023 Reviewer: Margaret. L Long* RV/CV Design and Analysis Donald M. McNutt III* RV/CV Design and Analysis Jared L. Geer* Radiation Engineering & Analysis Approved: Lynn A. Patterson*, Manager RV/CV Design and Analysis Jesse J Klingensmith*, Manager Radiation Engineering & Analysis

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii WCAP-18843-NP August 2023 Revision 0 RECORD OF REVISION Revision Description Completed 0 Original Issue August 2023

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii WCAP-18843-NP August 2023 Revision 0 TABLE OF CONTENTS LIST OF TABLES....................................................................................................................................... iv LIST OF FIGURES..................................................................................................................................... vi EXECUTIVE

SUMMARY

........................................................................................................................ VII 1 INTRODUCTION........................................................................................................................ 1-1 2 CALCULATED NEUTRON FLUENCE..................................................................................... 2-1

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS........................................................................... 2-2 2.3 NEUTRON DOSIMETRY.............................................................................................. 2-4 2.4 CALCULATIONAL UNCERTAINTIES........................................................................ 2-7 3 FRACTURE TOUGHNESS PROPERTIES................................................................................. 3-1 4 SURVEILLANCE DATA............................................................................................................. 4-1 5 CHEMISTRY FACTORS............................................................................................................. 5-1 6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE.......................................... 6-1 7 P-T LIMIT CURVE APPLICABILITY EVALUATION.............................................................. 7-1 8 PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDLIENS........... 8-1 8.1 PRESSURIZED THERMAL SHOCK............................................................................ 8-1 8.2 EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION............................. 8-3 9 UPPER SHELF ENERGY EVALUATION.................................................................................. 9-1 10 REFERENCES........................................................................................................................... 10-1 APPENDIX A CREDIBILITY RE-EVALUATION OF THE CATAWBA UNIT 2 SURVEILLANCE PROGRAM..................................................................................................................... A-1

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv WCAP-18843-NP August 2023 Revision 0 LIST OF TABLES Table 2-1 Reactor Pressure Vessel (RPV) Material Locations......................................................... 2-8 Table 2-2 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Reactor Pressure Vessel Clad/Base Metal Interface.................................................................................... 2-9 Table 2-3 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Reactor Pressure Vessel Clad/Base Metal Interface.............................................................................................. 2-10 Table 2-4 Calculated Maximum Iron Displacement Rate at the Reactor Pressure Vessel Clad/Base Metal Interface............................................................................................................... 2-11 Table 2-5 Calculated Maximum Iron Displacement at the Reactor Pressure Vessel Clad/Base Metal Interface......................................................................................................................... 2-12 Table 2-6 Projection of Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Reactor Pressure Vessel Materials in the Beltline and Extended Beltline................................... 2-13 Table 2-7 Projection of Iron Atom Displacements (dpa) Experienced by Reactor Pressure Vessel Materials in the Beltline and Extended Beltline............................................................ 2-14 Table 2-8 Calculated Fast Neutron Fluence Rate at the Geometric Center of the Surveillance Capsules......................................................................................................................... 2-15 Table 2-9 Calculated Fast Neutron Fluence at the Geometric Center of the Surveillance Capsules.... ....................................................................................................................................... 2-16 Table 2-10 Calculated Iron Atom Displacement Rate at the Geometric Center of the Surveillance Capsules......................................................................................................................... 2-17 Table 2-11 Calculated Iron Atom Displacement Rate at the Geometric Center of the Surveillance Capsules......................................................................................................................... 2-18 Table 2-12 Surveillance Capsule Lead Factors................................................................................ 2-19 Table 2-13 In-Vessel Surveillance Capsule Exposure Values......................................................... 2-20 Table 2-14 Calculational Uncertainties........................................................................................... 2-20 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Reactor Vessel Materials.................................................................................................. 3-2 Table 3-2 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Reactor Vessel Inlet and Outlet Nozzle Materials(a)(d)...................................................... 3-3 Table 4-1 Surveillance Capsule Data............................................................................................... 4-2 Table 5-1 Calculation of Chemistry Factors Using Surveillance Capsule Data............................... 5-2 Table 5-2 Summary of Position 1.1 and 2.1 Chemistry Factors...................................................... 5-3 Table 6-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Reactor Vessel Materials at 54 EFPY........................................................................ 6-2

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Westinghouse Non-Proprietary Class 3 v WCAP-18843-NP August 2023 Revision 0 Table 6-2 Adjusted Reference Temperature Evaluation for the Reactor Vessel Materials Through 54 EFPY at the 1/4T Location.............................................................................................. 6-3 Table 6-3 Adjusted Reference Temperature Evaluation for the Reactor Vessel Materials Through 54 EFPY at the 3/4T Location.............................................................................................. 6-5 Table 6-4 Limiting ART Values at 54 EFPY.................................................................................... 6-7 Table 7-1 Summary of the Limiting ART Values............................................................................. 7-1 Table 7-2 54 EFPY Heatup Curve Data Points using the 1996 Appendix G (with Flange Requirements and without Margins for Instrumentation Errors)..................................... 7-5 Table 7-3 54 EFPY Cooldown Curve Data Points using the 1996 Appendix G (with Flange Requirements and without Margins for Instrumentation Errors)..................................... 7-6 Table 8-1 RTPTS Calculations for Reactor Vessel Materials at 54 EFPY......................................... 8-2 Table 8-2 Evaluation of ERG Limit Category................................................................................. 8-3 Table 9-1 Predicted USE Values at 54 EFPY for the Reactor Vessel Beltline and Extended Beltline Materials.......................................................................................................................... 9-2 Table A-1 Identification of Credibility Criteria Affected by Fluence Change................................. A-1 Table A-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Catawba Unit 2 Surveillance Capsule Data Only.......................................................................... A-4 Table A-3 Surveillance Capsule Data Scatter about the Best-Fit Line Using Only Catawba Unit 2 Surveillance Data............................................................................................................ A-5

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi WCAP-18843-NP August 2023 Revision 0 LIST OF FIGURES Figure 2-1 Catawba Unit 2 Reactor Geometry in r,T at the Core Midplane with Single Surveillance Capsule.......................................................................................................................... 2-21 Figure 2-2 Plan View of the Reactor Geometry at the Core Midplane with Dual Surveillance Capsule ....................................................................................................................................... 2-22 Figure 2-3 Section View of the Reactor Geometry - 0q Azimuth................................................... 2-23 Figure 2-4 Section View of the Reactor Geometry - 31q Azimuth................................................. 2-24 Figure 2-5 Reactor Geometry in r,T at the Nozzles Centerline........................................................ 2-25 Figure 2-6 Reactor Geometry in r,z Section View at 67q Azimuthal Angle (Inlet Nozzle or Cold Leg) ....................................................................................................................................... 2-26 Figure 2-7 Reactor Geometry in r,z Section View at 22q Azimuthal Angle (Outlet Nozzle or Hot Leg)................................................................................................................................ 2-27 Figure 7-1 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60°, 80°, and 100°F/hr) Applicable for the First 54 EFPY (Without Margins for Instrumentation Errors).............................................................................................................................. 7-3 Figure 7-2 Catawba Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0°, 20°, 40°, 60° and 100°F/hr) Applicable for the First 54 EFPY (Without Margins for Instrumentation Errors).................................................................................................... 7-4 Figure 9-1 Regulatory Guide 1.99, Revision 2, Catawba Unit 2 Predicted Decrease in USE at 54 EFPY as a Function of Copper and Fluence.................................................................... 9-3

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Westinghouse Non-Proprietary Class 3 vii WCAP-18843-NP August 2023 Revision 0 EXECUTIVE

SUMMARY

This report provides an evaluation of several parameters for reactor vessel (RV) integrity in order to determine whether the applicable U.S. Nuclear Regulatory Commission (NRC) requirements are met and continued safe operation of Catawba Unit 2, with regards to RV integrity, can be justified through end-of-life extension (EOLE), i.e., 60 total years of operation. Specifically, this report documents the following Catawba Unit 2 RV integrity calculations/evaluations:

1. Perform an updated neutron fluence assessment for the Catawba Unit 2 pressure vessel materials.

Cycle-specific analyses for the past and current operating cycles, as well as projections for future operation through 54 effective full-power years (EFPY) are performed.

2. Determine the adjusted reference temperature (ART) values for the RV beltline and extended beltline materials at EOLE, 54 EFPY, i.e., 60 total years of operation for Catawba Unit 2.

3. Evaluate the applicability of the pressure-temperature (P-T) limit curves developed in WCAP-15285 [Ref. 1], with consideration of updated fluence values. The applicability of the Catawba Unit 2 P-T limit curves remains unchanged at 54 EFPY. The curves will be reprinted including a flange notch per 10 CFR 50, Appendix G [Ref. 8].

4. Determine the pressurized thermal shock (PTS) reference temperature (RTPTS) values for the beltline and extended beltline materials in the RV at 54 EFPY. The RTPTS values of all of the beltline and extended beltline materials in the Catawba Unit 2 RV are below the RTPTS screening criteria of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds (per 10 CFR 50.61.b.2), through EOLE (54 EFPY).

5. Determine the upper shelf energy (USE) values for the beltline and extended beltline materials in the RV at 54 EFPY. The USE values of all the beltline and extended beltline in the Catawba Unit 2 RV are above the USE screening criteria of 50 ft-lb per 10 CFR 50, Appendix G [Ref. 8] at 54 EFPY.

Appendix A contains a credibility evaluation for surveillance materials considering the updated fluence analysis.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 WCAP-18843-NP August 2023 Revision 0 1 INTRODUCTION The purpose of this report is to evaluate several parameters for reactor vessel (RV) integrity in order to determine whether the applicable U.S. Nuclear Regulatory Commission (NRC) requirements are met and continued safe operation of Catawba Unit 2, with regards to RV integrity, can be justified through end-of-life extension (EOLE), i.e., 60 total years of operation. As part of the time-limited aging analyses (TLAA) performed for the Catawba Unit 2 license renewal, a fluence analysis was completed, and new heatup and cooldown P-T limit curves were developed, applicable to 51 EFPY. These analyses are documented in WCAP-15285 [Ref. 1]. In this report, a new fluence analysis is performed, using the fully three-dimensional RAPTOR-M3G methodology. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) are established on a plant-and fuel cycle-specific basis for the first 24 cycles. Projections beyond Cycle 24 are based on average core power distributions and reactor operating conditions of Cycle 24. All of the neutron transport calculations performed in this analysis are based on the nuclear cross-section data derived from ENDF/B-VI and make use of the latest available calculation tools. The neutron transport methodology follows the guidance of Regulatory Guide 1.190 [Ref. 2], and the methods used to determine the RV neutron exposures are consistent with the NRC approved methodology described in WCAP-18124-NP-A [Ref. 3]. Heatup and cooldown P-T limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) of the limiting material of the reactor vessel. The adjusted reference temperature (ART) of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced RTNDT, and adding a margin. The unirradiated RTNDT (RTNDT(U)) is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F. RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, the RTNDT increases associated with the new fluence projections discussed above need to be evaluated to ensure that the limiting RTNDT used in the 51 EFPY P-T limit curves in WCAP-15285 remain bounding. To find the most limiting RTNDT at any time period in the reactors life, RTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The NRC has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. 4]. Regulatory Guide 1.99, Revision 2 is used for the calculation of ART values (RTNDT(U) + RTNDT + margin for uncertainties) at the quarter thickness (1/4T) and three-quarter thickness (3/4T) locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. These results are used to perform an evaluation to confirm the applicability of the P-T curves from WCAP-15285 with consideration of updated fluence values. This report documents the calculated ART values in Section 6. A description of the updated fluence analysis is provided in Section 2 of this report. Section 8 contains an evaluation for the RTPTS values for Catawba Unit 2 at 54 EFPY. Section 9 contains an evaluation for the USE values for Catawba Unit 2 at 54 EFPY. Appendix A provides a credibility evaluation of the Catawba Unit 2 surveillance data.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-1 WCAP-18843-NP August 2023 Revision 0 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

Discrete ordinates (SN) transport analyses were performed for the Catawba Unit 2 reactor to determine the neutron radiation environment within the reactor pressure vessel (RPV). In these analyses, radiation exposure parameters were established on a plant-and fuel-cycle-specific basis. The dosimetry analysis summarized in Section 2.3 shows that the +/-20% (1) acceptance criteria specified in Regulatory Guide 1.190 [Ref. 2] is met, based on the measurement-to-calculation (M/C) comparison results for the in-vessel surveillance capsules withdrawn and analyzed to-date. These validated calculations form the basis for providing projections of the neutron exposure of the RPV through EOLE. All of the calculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (specifically, ENDF/B-VI). Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET [Ref. 3]. The neutron transport evaluation methodology described in Reference 3 is based on the guidance of Regulatory Guide 1.190. Note, however, that the NRC-issued Safety Evaluation for WCAP-18124-NP-A appears in Section A of Reference 3. 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.

Regarding Limitation #1, 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 Catawba Unit 2 RPV extended beltline region is summarized in WCAP18124-NP-A, Revision 0, Supplement 1-P/NP [Ref. 18]. Note that the NRC has issued the Final Safety Evaluation Report for

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-2 WCAP-18843-NP August 2023 Revision 0 WCAP18124-NP-A, Revision 0, Supplement 1-P/NP. All RPV extended beltline calculations for Catawba Unit 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. Therefore, the application of the methodology used herein is approved by the NRC for both the traditional beltline and extended beltline regions. 2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the RPV, a series of fuel-cycle-specific forward transport calculations were performed using the three-dimensional discrete ordinates code, RAPTOR-M3G [Ref. 3], and the BUGLE-96 cross-section library [Ref. 5]. The BUGLE-96 library provides a coupled 47-neutron and 20-gamma-ray 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 the angular discretization was modeled with an S16 order of angular quadrature. Energy-and space-dependent core power distributions were treated on a fuel-cycle-specific basis. The Catawba Unit 2 reactor is a standard Westinghouse 4-loop design employing reactor internals that include 0.875-inch-thick baffle plates and a fully circumferential thermal shield. The model of the reactor (and reactor cavity) geometry used in the plant-specific evaluation is shown in Figure 2-2 through Figure 2-4. The model extends radially from the center of the core to 411.48 cm, azimuthally from 0° to 90° (taking advantage of the octant symmetry of the reactor configuration), and axially from -363.296 cm to 343.26 cm with respect to the midplane of the active core. Elevations of key RPV materials relative to the model geometry are provided in Table 2-1. A plan view of the model geometry at the core midplane is shown in Figure 2-2. In this figure, a single quarter is depicted showing the arrangement of the core, reactor internals, core barrel, thermal shield, downcomer, cladding, RPV, reactor cavity, reflective insulation, and bioshield. Depictions of the in-vessel surveillance capsules, including their associated support structures, are also shown. From a neutronics standpoint, the inclusion of the surveillance capsules and associated support structures in the geometric model is significant. Since the presence of the capsules and support structures has a marked impact on the magnitude of the neutron fluence rate and relative neutron and gamma ray spectra at dosimetry locations within the capsules, a meaningful evaluation of the radiation environment internal to the capsules can be made only when these perturbation effects are accounted for in the transport calculations. When developing the reactor model shown in Figure 2-2 through Figure 2-7, nominal design dimensions were employed for the various structural components. Likewise, water temperatures and, hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-3 WCAP-18843-NP August 2023 Revision 0 power operating conditions. These coolant temperatures were varied on a 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 and guide tubes. Note that the stainless-steel former plates located between the core baffle and barrel regions are shown in these figures. There are two models for RAPTOR-M3G [Ref. 3] to represent the in-vessel surveillance capsules: the single surveillance capsule is depicted in Figure 2-1 and the dual surveillance capsule is shown in Figure 2-2. The geometric mesh description of the reactor model (shown in Figure 2-1 through Figure 2-7) consisted of 222 radial by 257 azimuthal by 409 axial intervals. Mesh sizes were chosen to ensure sufficient resolution of the stair-step shaped baffle plates as well as an adequate number of meshes throughout the radial and axial regions of interest. The pointwise inner iteration convergence criterion utilized in the calculations was set at a value of 0.001. The core power distributions used in the plant-specific transport analysis were taken from nuclear design documentation. The data extracted included fuel assembly-specific initial enrichments, beginning-of-cycle burnups and end-of-cycle burnups. Appropriate axial power distributions were also obtained. For each fuel cycle of operation, fuel-assembly-specific enrichment and burnup data were used to generate the spatially dependent neutron source throughout the reactor core. This source description included the spatial variation of isotope-dependent (U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242) fission spectra, neutron emission rate per fission, and energy release per fission based on the burnup history of individual fuel assemblies. These fuel-assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin Cartesian coordinates to the spatial mesh arrays used in the discrete ordinates calculations. In Table 2-1, axial and azimuthal locations of the RPV materials are provided. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the midplane of the active fuel stack. Cycle-specific calculations were performed for Cycles 1-24. Note that future fluence projection data beyond Cycle 24 are based on Cycle 24, however it was identified that Catawba 2 has temporarily implemented a slight increase of RCS temperature (1.5°F Tavg increase) beginning with Cycle 26. Therefore, this temperature increase was conservatively considered to remain in place through remaining plant operation and entails a reduction of water density and an increase of the neutron flux out of the core. Maximum neutron fluence rates and fluences at the clad/base metal interface for the RPV are given in Table 2-2 and Table 2-3. Maximum iron atom displacements and iron atom displacement rates at the clad/base metal interface for the RPV are given in Table 2-4 and Table 2-5. Table 2-6 and Table 2-7 document the maximum fluence and iron atom displacement projections at the RPV material locations identified in Table 2-1. These projections were based on the spatial power distribution and reactor operating conditions of Cycles 24 but increasing of the RCS temperature. The projected results will remain valid as long as future plant operation is consistent with these assumptions, or if Catawba 2 reverts back to a lower RCS operating temperature. Note that RPV neutron exposure rates are dominated by neutron leakage from the peripheral fuel assemblies. Table 2-8 through Table 2-11 provide the results of the discrete ordinates transport analysis pertinent to the surveillance capsule evaluations. Table 2-8 and Table 2-9 provide the maximum fluence and fluence rates

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-4 WCAP-18843-NP August 2023 Revision 0 at the geometric center of the surveillance capsules. Table 2-10 and provide the maximum iron atom displacement and iron atom displacement rates at the geometric center of the surveillance capsules. Projections of future operation are based on the spatial power distributions and reactor operating conditions of Cycles 24 but increasing of the RCS temperature. The lead factors associated with the surveillance capsules are provided as a function of time in Table 2-12. The lead factor is defined as the ratio of the neutron fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule in Table 2-8 for both the single and the dual surveillance capsules to the maximum neutron fluence (E > 1.0 MeV) at the pressure vessel clad/base metal interface in Table 2-3. The exposure values for each capsule withdrawn from the reactor and analyzed (Capsules Z, X, V, and Y) are in Table 2-13. This table includes the fast fluence rate (E > 1.0 MeV), the iron atom displacement and the lead factor at each surveillance capsule location. 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 four (4) previously analyzed in-vessel surveillance capsules removed from each of Catawba Unit 2 (Capsules Z, X, V, and Y) and the eight (8) ex-vessel midplane capsules (Set 2S-1 removed at the end of Cycle 14 and Set 2S-2 removed at the end of Cycle 20) is provided in this section of the report. Another in-vessel capsule (capsule W) was withdrawn at the end of cycle 14 and it is stored without being analyzed. The irradiation parameters, position in the core and type of capsule are shown below: Parameter Capsule Z Capsule X Capsule V Capsule Y Capsule W Starting Cycle of Irradiation 1 1 1 1 1 Starting Date of Irradiation 8/18/1986 8/18/1986 8/18/1986 8/18/1986 8/18/1986 Ending Cycle of Irradiation 1 5 9 9 14 Ending Date of Irradiation 12/24/1987 1/29/1993 9/5/1998 9/5/1998 2006 Dosimetry Read Date 4/25/1988 6/3/1993 2/22/1999 2/22/1999 Stored Capsule Azimuth 31.5° 29° 29° 31.5° 31.5° Capsules-per-Octant (a) Single Dual Dual Dual Single Capsule Radius (cm) 207.315 207.315 207.315 207.315 207.315 Note(s) (a): Single Corresponds to Single Surveillance Capsule model (Figure 2-1) and Dual" corresponds to Dual Surveillance Capsule model (Figure 2-2).

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-5 WCAP-18843-NP August 2023 Revision 0 The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsules Z, X, V, and Y that were withdrawn from the reactor, is summarized below. Reaction Capsule Average % Std. Dev. Capsule Z (31.5° Single) Capsule X (29° Dual) Capsule V (29° Dual) Capsule Y (31.5° Dual) 63Cu (n,) 60Co 1.16 1.08 0.98 0.96 1.05 8.9 54Fe (n,p) 54Mn 0.99 0.87 0.86 0.83 0.89 7.9 58Ni (n,p) 58Co 0.94 0.84 0.88 0.83 0.87 5.7 238U(Cd) (n,f) 137Cs 1.13 1.06 1.04 0.97 1.05 6.3 237Np(Cd) (n,f) 137Cs 1.12 1.12 1.02 0.90 1.04 10.1 Average of M/C Results 0.98 11.1 The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsules A, B, C, and E for Set 2S-1 and Capsules G, H, I, and K for Set 2S-2, is summarized below. Capsule M/C Ratio Cu-63 (n,D) Fe-54 (n,p) Ni-58 (n,p) Ti-46 (n,p) Nb-93 (n,n) U-238 (n,f) Np-237 (n,f) 1-A 0.87 0.95 0.87 1.02 1.00 1.15 1-B 1.02 1.11 1.00 1.06 1.19 1.32 1-C 1.03 1.12 0.99 1.08 1.08 1.19 1-E 0.95 1.04 0.94 0.95 1.12 1.26 2-G 0.84 0.94 0.89 0.91 1.37 2-H 0.97 1.01 1.00 0.99 1.11 2-I 1.05 1.09 1.05 1.08 1.13 2-K 0.93 0.98 0.91 0.96 1.11 Average 0.96 1.03 0.96 1.01 1.18 1.10 1.23 Standard Deviation 7.9 6.9 6.6 6.4 10.8 7.2 6.1

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-6 WCAP-18843-NP August 2023 Revision 0 The measurement to calculation comparisons based on the individual sensor reactions without recourse to the least-squares adjustment procedure are summarized as follows: Reaction In-Vessel Database Midplane Ex-Vessel Database Combined Database Average M/C Standard Deviation (1) Average M/C Standard Deviation (1) Average M/C Standard Deviation (1) Cu-63 (n,) 1.05 8.9 0.96 7.9 1.00 6.0 Ti-46 (n,p) 1.01 6.4 Fe-54 (n,p) 0.89 7.9 1.03 6.9 0.96 5.2 Ni-58 (n,p) 0.87 5.7 0.96 6.6 0.91 4.4 U-238 (n,f) Cd 1.05 6.3 1.10 7.2 1.07 4.8 Np-237 (n,f) Cd 1.04 10.1 1.23 6.1 1.14 5.7 Nb-93 (n,n) 1.18 10.8 Linear Average 0.98 11.1 1.07 10.1 1.02 7.5 The average measured-to-calculated (M/C) reaction rate ratios for the threshold reactions of the four in-vessel previously withdrawn surveillance capsules and the eight midplane ex-vessel capsules for Catawba Unit 2 is 1.02 r 7.5% (1V). Each of these data comparisons shows that for both in-vessel and midplane ex-vessel locations the measurements and calculations agree well within the 20% criterion specified in Regulatory Guide 1.190. In fact, both the average M/C results and Best Estimate over Calculated (BE/C) results fall within the 13% (1V) uncertainty assigned to the fast neutron exposure of the absolute transport calculations. This comparison provides a validation of the plant-specific neutron transport calculations for Catawba Unit 2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-7 WCAP-18843-NP August 2023 Revision 0 2.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the RPV is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology used in the plant-specific neutron exposure evaluation is carried out in the following four stages:

1. Comparisons of calculations with benchmark measurements from the pool critical assembly (PCA) simulator (NUREG/CR-6454, Pool Critical Assembly Pressure Vessel Facility Benchmark [Ref. 6]) at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 experiment.

2. Comparison of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment (NUREG/CR-6453, H.B.

Robinson-2 Pressure Vessel Benchmark [Ref. 7]).

3. 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 (WCAP-18124-NP-A [Ref. 3]).

4. Comparison of the calculations with all available dosimetry results from the RPV measurement programs carried out at Catawba Unit 2.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations, nor did it address uncertainties in operational and geometric variables that impact power reactor calculations. The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties 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 method approximations as well as to a lack of knowledge relative to various plant-specific parameters. The overall calculational uncertainty applicable to the Catawba Unit 2 analyses were established from the results of these three phases of the methods qualification. The fourth phase of the uncertainty assessment (comparisons of plant-specific dosimetry measurements) was used solely to demonstrate the adequacy of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used to bias the final results in any way. Table 2-14 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in WCAP-18124-NP-A. 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 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-8 WCAP-18843-NP August 2023 Revision 0 Table 2-1 Reactor Pressure Vessel (RPV) Material Locations Material Azimuthal Location(a) [°] Axial Location(b) [cm] Min Max Min Max Outlet Nozzle to Nozzle Shell Welds - Nozzle 1 - Nozzle 2 - Nozzle 3 - Nozzle 4 22.0 158.0 202.0 338.0 274.88 274.88 274.88 274.88 Inlet Nozzle to Nozzle Shell Welds - Nozzle 1 - Nozzle 2 - Nozzle 3 - Nozzle 4 67.0 113.0 247.0 293.0 264.80 264.80 264.80 264.80 Nozzle Shell Longitudinal Welds(c) - Weld 1 - Weld 2 - Weld 3 42.0 162.0 282.0 236.14 236.14 236.14 300 300 300 Nozzle Shell(d) 0 360.0 236.14 300 Nozzle Shell to Intermediate Shell Circumferential Weld(g) 0 360.0 234.55 237.73 Intermediate Shell 0 360.0 -35.94 236.14 Intermediate Shell Longitudinal Welds(e) - Weld 1 - Weld 2 - Weld 3 0 120.0 240.0 -35.94 -35.94 -35.94 236.14 236.14 236.14 Intermediate Shell to Lower Shell Circumferential Weld(g) 0 360.0 -37.53 -34.35 Lower Shell 0 360.0 -310.02 -35.94 Lower Shell Longitudinal Welds(f) - Weld 1 - Weld 2 - Weld 3 60.0 180.0 300.0 -310.02 -310.02 -310.02 -35.94 -35.94 -35.94 Lower Shell to Lower Vessel Head Circumferential Weld(g) 0 360.0 -311.61 -308.43 Core Support Lugs(h) 0 90.0 -268.605 -268.605 Note(s): (a) Azimuthal locations are indexed to = 0.0, where on the reactor vessel the =0 position is indicated in Drawing E-8871-171-005 Rev.1 General Plan Arrangement [Ref 20] (b) Axial elevations are indexed to Z = 0.0 at the midplane of the active fuel stack. (c) Evaluated at 42, 72, and 12 degrees and assumed a maximum elevation of 300 cm. (d) Assumed a maximum elevation of 300 cm. (e) Evaluated at 0°, 30° and 60°, respectively. (f) Evaluated at 60°, 0°, 30° degrees respectively. (g) The circumferential weld, which is assumed to be 1.25 inches in height. (h) The core support lugs are evaluated from the inner radius to the RPV and from the whole azimuthal angle.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-9 WCAP-18843-NP August 2023 Revision 0 Table 2-2 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Reactor Pressure Vessel Clad/Base Metal Interface Cycle Cycle Length (EFPY) Total Time (EFPY) Maximum Fluence Rate (n/cm2-s) 1 0.86 0.86 2.94E+10 2 0.78 1.64 2.25E+10 3 0.91 2.55 2.34E+10 4 0.95 3.50 2.19E+10 5 1.02 4.52 2.12E+10 6 1.03 5.55 2.31E+10 7 1.17 6.72 2.09E+10 8 1.21 7.93 2.05E+10 9 1.31 9.24 1.77E+10 10 1.22 10.46 1.82E+10 11 1.41 11.87 2.02E+10 12 1.30 13.17 1.71E+10 13 1.45 14.62 1.59E+10 14 1.38 16.00 1.58E+10 15 1.35 17.35 1.46E+10 16 1.30 18.65 1.41E+10 17 1.41 20.06 1.42E+10 18 1.36 21.42 1.51E+10 19 1.40 22.82 1.49E+10 20 1.34 24.16 1.49E+10 21 1.43 25.59 1.50E+10 22 1.42 27.01 1.51E+10 23 1.41 28.42 1.47E+10 24 1.46 29.88 1.49E+10 Projection(a) 1.53E+10 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-10 WCAP-18843-NP August 2023 Revision 0 Table 2-3 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Reactor Pressure Vessel Clad/Base Metal Interface Cycle Cycle Length (EFPY) Total Time (EFPY) Maximum Fluence (n/cm2) 1 0.86 0.86 7.98E+17 2 0.78 1.64 1.30E+18 3 0.91 2.55 1.95E+18 4 0.95 3.50 2.59E+18 5 1.02 4.52 3.27E+18 6 1.03 5.55 4.02E+18 7 1.17 6.72 4.79E+18 8 1.21 7.93 5.56E+18 9 1.31 9.24 6.29E+18 10 1.22 10.46 6.92E+18 11 1.41 11.87 7.66E+18 12 1.30 13.17 8.30E+18 13 1.45 14.62 9.02E+18 14 1.38 16.00 9.71E+18 15 1.35 17.35 1.03E+19 16 1.30 18.65 1.09E+19 17 1.41 20.06 1.15E+19 18 1.36 21.42 1.22E+19 19 1.40 22.82 1.28E+19 20 1.34 24.16 1.35E+19 21 1.43 25.59 1.41E+19 22 1.42 27.01 1.48E+19 23 1.41 28.42 1.55E+19 24 1.46 29.88 1.61E+19 Projection(a) 30 1.62E+19 42 2.19E+19 50 2.58E+19 54 2.77E+19 72 3.64E+19 80 4.02E+19 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-11 WCAP-18843-NP August 2023 Revision 0 Table 2-4 Calculated Maximum Iron Displacement Rate at the Reactor Pressure Vessel Clad/Base Metal Interface(a) Cycle Cycle Length (EFPY) Total Time (EFPY) Iron Displacement Rate (dpa/s) 1 0.86 0.86 4.65E-11 2 0.78 1.64 3.44E-11 3 0.91 2.55 3.60E-11 4 0.95 3.50 3.34E-11 5 1.02 4.52 3.23E-11 6 1.03 5.55 3.52E-11 7 1.17 6.72 3.19E-11 8 1.21 7.93 3.12E-11 9 1.31 9.24 2.71E-11 10 1.22 10.46 2.88E-11 11 1.41 11.87 3.18E-11 12 1.30 13.17 2.69E-11 13 1.45 14.62 2.43E-11 14 1.38 16.00 2.41E-11 15 1.35 17.35 2.24E-11 16 1.30 18.65 2.15E-11 17 1.41 20.06 2.17E-11 18 1.36 21.42 2.31E-11 19 1.40 22.82 2.27E-11 20 1.34 24.16 2.28E-11 21 1.43 25.59 2.29E-11 22 1.42 27.01 2.31E-11 23 1.41 28.42 2.25E-11 24 1.46 29.88 2.27E-11 Projection(b) 2.34E-11 Note(s): (a) The fluence values provided are evaluated at the vessel clad/base metal interface. (b) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-12 WCAP-18843-NP August 2023 Revision 0 Table 2-5 Calculated Maximum Iron Displacement at the Reactor Pressure Vessel Clad/Base Metal Interface(a) Cycle Cycle Length (EFPY) Total Time (EFPY) Iron Displacement (dpa) 1 0.86 0.86 1.26E-03 2 0.78 1.64 2.06E-03 3 0.91 2.55 3.09E-03 4 0.95 3.50 3.95E-03 5 1.02 4.52 4.99E-03 6 1.03 5.55 6.13E-03 7 1.17 6.72 7.31E-03 8 1.21 7.93 8.49E-03 9 1.31 9.24 9.60E-03 10 1.22 10.46 1.06E-02 11 1.41 11.87 1.17E-02 12 1.30 13.17 1.27E-02 13 1.45 14.62 1.38E-02 14 1.38 16.00 1.48E-02 15 1.35 17.35 1.58E-02 16 1.30 18.65 1.67E-02 17 1.41 20.06 1.76E-02 18 1.36 21.42 1.86E-02 19 1.40 22.82 1.96E-02 20 1.34 24.16 2.06E-02 21 1.43 25.59 2.16E-02 22 1.42 27.01 2.26E-02 23 1.41 28.42 2.36E-02 24 1.46 29.88 2.46E-02 Projection(b) 30 2.47E-02 42 3.35E-02 50 3.94E-02 54 4.23E-02 72 5.56E-02 80 6.14E-02 Note(s): (a) The fluence values provided are evaluated at the vessel clad/base metal interface. (b) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-13 WCAP-18843-NP August 2023 Revision 0 Table 2-6 Projection(a) of Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by Reactor Pressure Vessel Materials in the Beltline and Extended Beltline Beltline and Extended Beltline Material Neutron Fluence (n/cm2)(b) 30 EFPY 42 EFPY 50 EFPY 54 EFPY 72 EFPY 80 EFPY Vessel Shells Lower Shell 1.59E+19 2.17E+19 2.56E+19 2.75E+19 3.62E+19 4.00E+19 Intermediate Shell 1.62E+19 2.19E+19 2.58E+19 2.77E+19 3.64E+19 4.02E+19 Nozzle Shell 2.81E+17 3.78E+17 4.42E+17 4.74E+17 6.19E+17 6.84E+17 Circumferential Welds Bottom Torus to Lower Shell 3.64E+14 4.97E+14 5.85E+14 6.29E+14 8.27E+14 9.15E+14 Lower Shell to Intermediate Shell 1.60E+19 2.17E+19 2.56E+19 2.75E+19 3.62E+19 4.01E+19 Intermediate Shell to Nozzle Shell 3.20E+17 4.30E+17 5.03E+17 5.40E+17 7.05E+17 7.78E+17 Longitudinal Welds Lower Shell 60° (101-142A) 1.28E+19 1.75E+19 2.06E+19 2.21E+19 2.91E+19 3.22E+19 180° (101-142B) 8.73E+18 1.18E+19 1.39E+19 1.49E+19 1.96E+19 2.17E+19 300° (101-142C) 9.71E+18 1.33E+19 1.56E+19 1.68E+19 2.21E+19 2.45E+19 Intermediate Shell 0° (101-124A) 8.85E+18 1.19E+19 1.40E+19 1.50E+19 1.97E+19 2.18E+19 120° (101-124B) 9.68E+18 1.30E+19 1.53E+19 1.64E+19 2.15E+19 2.37E+19 240° (101-124C) 1.30E+19 1.76E+19 2.07E+19 2.23E+19 2.92E+19 3.23E+19 Nozzle Shell 42° (101-122A) 2.78E+17 3.73E+17 4.37E+17 4.69E+17 6.12E+17 6.76E+17 162° (101-122B) 2.12E+17 2.87E+17 3.37E+17 3.62E+17 4.75E+17 5.25E+17 282° (101-122C) 1.80E+17 2.43E+17 2.85E+17 3.06E+17 4.02E+17 4.44E+17 Nozzle Forging Attachment Welds Inlet Nozzle 2.40E+16 3.27E+16 3.85E+16 4.14E+16 5.45E+16 6.02E+16 Outlet Nozzle 1.14E+16 1.55E+16 1.83E+16 1.97E+16 2.59E+16 2.87E+16 Core Support Lugs 2.81E+17 3.78E+17 4.42E+17 4.74E+17 6.19E+17 6.84E+17 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2. (b) Fluence values are taken from the vessel clad/ base metal interface.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-14 WCAP-18843-NP August 2023 Revision 0 Table 2-7 Projection(a) of Iron Atom Displacements (dpa) Experienced by Reactor Pressure Vessel Materials in the Beltline and Extended Beltline Beltline and Extended Beltline Material Iron Atom Displacements (dpa) 30 EFPY 42 EFPY 50 EFPY 54 EFPY 72 EFPY 80 EFPY Reactor Vessel Shells Plates Lower Shell Plates 2.44E-02 3.32E-02 3.91E-02 4.21E-02 5.54E-02 6.12E-02 Intermediate Shell Plates 2.47E-02 3.35E-02 3.94E-02 4.23E-02 5.56E-02 6.14E-02 Nozzle Shell Plates 4.76E-04 6.40E-04 7.49E-04 8.03E-04 1.05E-03 1.16E-03 Circumferential Welds Bottom Torus to Lower Shell 2.44E-06 3.30E-06 3.87E-06 4.15E-06 5.44E-06 6.01E-06 Lower Shell to Intermediate Shell 2.44E-02 3.32E-02 3.91E-02 4.21E-02 5.54E-02 6.12E-02 Intermediate Shell to Nozzle Shell 5.39E-04 7.23E-04 8.47E-04 9.08E-04 1.19E-03 1.31E-03 Longitudinal Welds Lower Shell Welds 60° (101-142A) 1.99E-02 2.72E-02 3.20E-02 3.44E-02 4.52E-02 5.01E-02 180° (101-142B) 1.35E-02 1.83E-02 2.16E-02 2.32E-02 3.04E-02 3.36E-02 300° (101-142C) 1.54E-02 2.11E-02 2.48E-02 2.67E-02 3.52E-02 3.89E-02 Intermediate Shell Welds 0° (101-124A) 1.37E-02 1.85E-02 2.17E-02 2.33E-02 3.05E-02 3.37E-02 120° (101-124B) 1.54E-02 2.08E-02 2.44E-02 2.62E-02 3.43E-02 3.79E-02 240° (101-124C) 2.02E-02 2.74E-02 3.22E-02 3.46E-02 4.54E-02 5.02E-02 Nozzle Shell Welds 42° (101-122A) 4.71E-04 6.32E-04 7.40E-04 7.94E-04 1.04E-03 1.14E-03 162° (101-122B) 3.70E-04 5.00E-04 5.87E-04 6.31E-04 8.27E-04 9.14E-04 282° (101-122C) 3.11E-04 4.21E-04 4.94E-04 5.31E-04 6.95E-04 7.68E-04 Nozzle Forging Attachment Welds Inlet Nozzle 5.33E-05 7.23E-05 8.50E-05 9.13E-05 1.20E-04 1.32E-04 Outlet Nozzle 2.58E-05 3.50E-05 4.11E-05 4.42E-05 5.80E-05 6.41E-05 Core Support Lugs 4.76E-04 6.40E-04 7.49E-04 8.03E-04 1.05E-03 1.16E-03 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-15 WCAP-18843-NP August 2023 Revision 0 Table 2-8 Calculated Fast Neutron Fluence Rate at the Geometric Center of the Surveillance Capsules Cycle Total Time (EFPY) Neutron Fluence Rate (n/cm2-s) Dual Single 29.0° 31.5° 31.5° 1 0.86 1.02E+11 1.12E+11 1.12E+11 2 1.64 7.79E+10 8.45E+10 8.41E+10 3 2.55 8.19E+10 8.78E+10 8.74E+10 4 3.50 7.18E+10 7.20E+10 7.14E+10 5 4.52 6.91E+10 6.99E+10 6.94E+10 6 5.55 7.86E+10 8.20E+10 8.15E+10 7 6.72 7.44E+10 7.89E+10 7.84E+10 8 7.93 6.86E+10 7.13E+10 7.08E+10 9 9.24 6.56E+10 6.99E+10 6.95E+10 10 10.46 6.33E+10 7.01E+10 6.99E+10 11 11.87 6.00E+10 6.56E+10 6.53E+10 12 13.17 6.11E+10 6.69E+10 6.66E+10 13 14.62 5.64E+10 5.85E+10 5.81E+10 14 16.00 5.79E+10 6.13E+10 6.10E+10 15 17.35 5.52E+10 5.82E+10 5.79E+10 16 18.65 5.10E+10 5.33E+10 5.30E+10 17 20.06 4.93E+10 5.08E+10 5.05E+10 18 21.42 5.56E+10 5.81E+10 5.77E+10 19 22.82 5.39E+10 5.64E+10 5.61E+10 20 24.16 5.31E+10 5.49E+10 5.46E+10 21 25.59 5.29E+10 5.54E+10 5.51E+10 22 27.01 5.50E+10 5.78E+10 5.75E+10 23 28.42 5.16E+10 5.32E+10 5.28E+10 24 29.88 5.37E+10 5.58E+10 5.55E+10 Projection(a) 5.46E+10 5.67E+10 5.64E+10 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-16 WCAP-18843-NP August 2023 Revision 0 Table 2-9 Calculated Fast Neutron Fluence at the Geometric Center of the Surveillance Capsules Cycle Total Time (EFPY) Neutron Fluence (n/cm2) Dual Single 29.0° 31.5° 31.5° 1 0.86 2.76E+18 3.04E+18 3.03E+18 2 1.64 4.69E+18 5.13E+18 5.11E+18 3 2.55 7.03E+18 7.64E+18 7.61E+18 4 3.50 9.18E+18 9.80E+18 9.75E+18 5 4.52 1.14E+19 1.20E+19 1.20E+19 6 5.55 1.40E+19 1.47E+19 1.46E+19 7 6.72 1.67E+19 1.76E+19 1.75E+19 8 7.93 1.93E+19 2.03E+19 2.02E+19 9 9.24 2.20E+19 2.32E+19 2.31E+19 10 10.46 2.45E+19 2.60E+19 2.58E+19 11 11.87 2.71E+19 2.89E+19 2.87E+19 12 13.17 2.97E+19 3.16E+19 3.14E+19 13 14.62 3.22E+19 3.43E+19 3.41E+19 14 16.00 3.48E+19 3.69E+19 3.68E+19 15 17.35 3.71E+19 3.94E+19 3.92E+19 16 18.65 3.92E+19 4.16E+19 4.14E+19 17 20.06 4.14E+19 4.39E+19 4.36E+19 18 21.42 4.38E+19 4.64E+19 4.61E+19 19 22.82 4.62E+19 4.89E+19 4.86E+19 20 24.16 4.84E+19 5.12E+19 5.09E+19 21 25.59 5.08E+19 5.37E+19 5.34E+19 22 27.01 5.33E+19 5.63E+19 5.60E+19 23 28.42 5.56E+19 5.87E+19 5.83E+19 24 29.88 5.80E+19 6.12E+19 6.09E+19 Projection(a) 30 5.82E+19 6.14E+19 6.11E+19 42 7.89E+19 8.29E+19 8.24E+19 50 9.27E+19 9.72E+19 9.67E+19 54 9.96E+19 1.04E+20 1.04E+20 72 1.31E+20 1.37E+20 1.36E+20 80 1.44E+20 1.51E+20 1.50E+20 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-17 WCAP-18843-NP August 2023 Revision 0 Table 2-10 Calculated Iron Atom Displacement Rate at the Geometric Center of the Surveillance Capsules Cycle Total Time (EFPY) Iron Atom Displacement Rate (dpa/s) Dual Single 29.0° 31.5° 31.5° 1 0.86 1.95E-10 2.15E-10 2.15E-10 2 1.64 1.48E-10 1.61E-10 1.61E-10 3 2.55 1.56E-10 1.68E-10 1.68E-10 4 3.50 1.36E-10 1.37E-10 1.36E-10 5 4.52 1.31E-10 1.33E-10 1.32E-10 6 5.55 1.50E-10 1.56E-10 1.56E-10 7 6.72 1.42E-10 1.51E-10 1.50E-10 8 7.93 1.30E-10 1.36E-10 1.35E-10 9 9.24 1.24E-10 1.33E-10 1.32E-10 10 10.46 1.20E-10 1.34E-10 1.33E-10 11 11.87 1.14E-10 1.25E-10 1.25E-10 12 13.17 1.16E-10 1.27E-10 1.27E-10 13 14.62 1.06E-10 1.11E-10 1.10E-10 14 16.00 1.10E-10 1.16E-10 1.16E-10 15 17.35 1.04E-10 1.10E-10 1.10E-10 16 18.65 9.63E-11 1.01E-10 1.01E-10 17 20.06 9.31E-11 9.62E-11 9.59E-11 18 21.42 1.05E-10 1.10E-10 1.10E-10 19 22.82 1.02E-10 1.07E-10 1.06E-10 20 24.16 1.00E-10 1.04E-10 1.04E-10 21 25.59 9.99E-11 1.05E-10 1.05E-10 22 27.01 1.04E-10 1.09E-10 1.09E-10 23 28.42 9.74E-11 1.00E-10 1.00E-10 24 29.88 1.01E-10 1.06E-10 1.05E-10 Projection(a) 1.03E-10 1.08E-10 1.07E-10 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-18 WCAP-18843-NP August 2023 Revision 0 Table 2-11 Calculated Iron Atom Displacement Rate at the Geometric Center of the Surveillance Capsules Cycle Total Time (EFPY) Iron Atom Displacements (dpa) Dual Single 29.0° 31.5° 31.5° 1 0.86 5.29E-03 5.85E-03 5.84E-03 2 1.64 8.95E-03 9.83E-03 9.82E-03 3 2.55 1.34E-02 1.46E-02 1.46E-02 4 3.50 1.75E-02 1.87E-02 1.87E-02 5 4.52 2.17E-02 2.30E-02 2.29E-02 6 5.55 2.66E-02 2.81E-02 2.81E-02 7 6.72 3.18E-02 3.37E-02 3.36E-02 8 7.93 3.68E-02 3.88E-02 3.87E-02 9 9.24 4.19E-02 4.43E-02 4.42E-02 10 10.46 4.65E-02 4.95E-02 4.94E-02 11 11.87 5.16E-02 5.50E-02 5.49E-02 12 13.17 5.63E-02 6.02E-02 6.01E-02 13 14.62 6.12E-02 6.53E-02 6.52E-02 14 16.00 6.60E-02 7.04E-02 7.02E-02 15 17.35 7.04E-02 7.51E-02 7.49E-02 16 18.65 7.44E-02 7.92E-02 7.91E-02 17 20.06 7.85E-02 8.35E-02 8.33E-02 18 21.42 8.30E-02 8.82E-02 8.80E-02 19 22.82 8.75E-02 9.29E-02 9.27E-02 20 24.16 9.18E-02 9.73E-02 9.71E-02 21 25.59 9.63E-02 1.02E-01 1.02E-01 22 27.01 1.01E-01 1.07E-01 1.07E-01 23 28.42 1.05E-01 1.11E-01 1.11E-01 24 29.88 1.10E-01 1.16E-01 1.16E-01 Projection(a) 30 1.10E-01 1.17E-01 1.16E-01 42 1.49E-01 1.57E-01 1.57E-01 50 1.75E-01 1.85E-01 1.84E-01 54 1.88E-01 1.98E-01 1.98E-01 72 2.48E-01 7.87E-03 2.45E-02 80 2.73E-01 2.86E-01 2.86E-01 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-19 WCAP-18843-NP August 2023 Revision 0 Table 2-12 Surveillance Capsule Lead Factors Cycle Total Time (EFPY) Dual Single 29.0° 31.5° 31.5° 1 0.86 3.47 3.81 3.80 2 1.64 3.60 3.94 3.93 3 2.55 3.60 3.91 3.90 4 3.50 3.55 3.78 3.76 5 4.52 3.49 3.68 3.66 6 5.55 3.48 3.66 3.64 7 6.72 3.49 3.68 3.66 8 7.93 3.48 3.66 3.64 9 9.24 3.50 3.69 3.67 10 10.46 3.54 3.75 3.73 11 11.87 3.54 3.77 3.75 12 13.17 3.57 3.81 3.79 13 14.62 3.57 3.80 3.78 14 16.00 3.58 3.81 3.79 15 17.35 3.59 3.82 3.80 16 18.65 3.59 3.82 3.80 17 20.06 3.59 3.81 3.79 18 21.42 3.60 3.81 3.79 19 22.82 3.60 3.81 3.79 20 24.16 3.60 3.80 3.78 21 25.59 3.60 3.80 3.78 22 27.01 3.60 3.80 3.78 23 28.42 3.60 3.80 3.77 24 29.88 3.60 3.80 3.77 Projection(a) 30 3.60 3.80 3.77 42 3.60 3.78 3.76 50 3.60 3.77 3.75 54 3.60 3.77 3.75 72 3.59 3.76 3.74 80 3.59 3.76 3.73 Note(s): (a) Projections beyond Cycle 24 are based on the average core power distribution and reactor operation conditions of Cycle 24 but include an increase of the RCS temperatures as discussed in Section 2.2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-20 WCAP-18843-NP August 2023 Revision 0 Table 2-13 In-Vessel Surveillance Capsule Exposure Values Capsule Capsule Type Capsule Location (First-quarter-Equivalent) Withdrawn (EOC) Withdrawal (EFPY) Capsule Fluence (E > 1.0 MeV) (n/cm2) Lead Factor Iron Atom Displacements (dpa) Z Single 301.5° (31.5°) 1 0.86 3.03E+18 3.80 5.84E-03 X Dual 241° (29°) 5 4.52 1.14E+19 3.49 2.17E-02 V Dual 61° (29°) 9 9.24 2.20E+19 3.50 4.19E-02 Y Dual 238.5° (31.5°) 9 9.24 2.32E+19 3.69 4.43E-02 W Single 121.5° (31.5°) 14 16.00 3.68E+19 3.79 7.02E-02 Table 2-14 Calculational Uncertainties Description Uncertainty Capsule Vessel Inner Radius PCA comparisons 3% 3% H.B. Robinson comparisons 5% 5% Analytical sensitivity studies 9% 11% Additional uncertainty for factors not explicitly evaluated 5% 5% Net calculational uncertainty 12% 13%

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-21 WCAP-18843-NP August 2023 Revision 0 Figure 2-1 Catawba Unit 2 Reactor Geometry in r,TT at the Core Midplane with Single Surveillance Capsule

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation) 400 300 200 100 0

0 100 200 X Axis 300 Filled Boundary Var: materials 28 Core (Zone 3) 39 Core Baffle 43 Byp ass Region 44 Core Barrel 45 Downcomer 46 Neutron Pad 47 Pressure Vessel Clad 48 Pressure Vessel 49 Cavity Air 50 Reflective Insulation 51 Bioshield

• • 1---0l Surveillance Cap. Holder L62 Surveillance Capsule 400

Westinghouse Non-Proprietary Class 3 2-22 WCAP-18843-NP August 2023 Revision 0 Figure 2-2 Plan View of the Reactor Geometry at the Core Midplane with Dual Surveillance Capsule

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation) 400 300 200 100 0

100 200 X Axis Filled Boundary Var: materials 28 Core (Zone 3) 39 Core Baffle 43 Bypass Region 44 Core Barrel 45 Downcomer 46 Neutron Pa d 47 Pressure Vessel Clad 48 Pressure Vessel 49 Cavity Air 50 Reflective Insulation 51 Bioshield

• • l---0 1 Surveillance Cap. Holder L.62 Surveillance Capsule 300 400

Westinghouse Non-Proprietary Class 3 2-23 WCAP-18843-NP August 2023 Revision 0 Figure 2-3 Section View of the Reactor Geometry - 0qq Azimuth

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Filled Boundary Var: materials 26 Core (Zone l ) 27 Core (Zone 2) 300 28 Core (Zone 3) 29 Core (Zone 4) 30 Core (Zone 5) - 31 Core (Zone 6) 32 Fuel Plenum 3 Top Water Gap 4 Top Nozzle 200 35 Bottom End Plugs 6 Bottom Water Gap 37 Bottom Nozzle Plate 38 Bottom Nozzle Leg 39 Core Baffle -40 LCS (Region N 100 41 LCS(Region B) 42 Formers 43 Bypass Region 44 Core Barrel 45 Downcomer 47 Pressure Vessel Clad Ill 48 Pressure Vessel "'"I 0 49 Cavity Air >c ~ 0 Reflective Insulation 51 Bioshield 52 UCP (Region N 53 Outlet Plenum (Region N 54 Inlet Plenum (Region N -100 55 Outlet Plenum (Region B) 57 LCP (Region N 58 UCP (Region B) 59 Upper Core Blanket 0 Lower Core Blanket 3 LCP (Region B) 4 Support (Region N -200 5 Support (Region B) -300 100 200 300 400 X Axis

Westinghouse Non-Proprietary Class 3 2-24 WCAP-18843-NP August 2023 Revision 0 Figure 2-4 Section View of the Reactor Geometry - 31qq Azimuth

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Fille_d Boun_dary 26 Core (Zone l ) 3 00 27 Core (Zone 2) 28 Core (Zone 3) 29 Core (Zone 4) 0 Core (Zone 5) l Core (Zone 6) 32 Fuel Plenum 3 Top Water Gap 200 34 Top Nozzle 35 Bottom End Plugs 6 Bottom Water Gap 7 Bottom Nozzle Plate 38 Bottom Nozzle Leg 39 Core Baffle -40 LCS (Region A; 100 -4 l LCS(Region B) I'll 42 Formers

• --t 43 Byp ass Region

~, 44 Core Barrel 45 Downcomer 0 46 Neutron Pad 0 47 Pressure Vessel Clad '"1 0 48 Pressure Vessel 49 Cavity Air 0 0 50 Reflective Insulation 51 Bioshield 0.. 2 UCP (Region A; 0 53 Outlet Plenum (Region 0 0 - 100 54 Inlet Plenum (Region A; 5 Outlet Plenum (Region B 7 LCP (Region A; 58 UCP (Region B) 59 Upper Core Blanket 0 Lower Core Blanket l Surveillance Cap. Holde - 200 2 Surveillance Capsule 3 LCP (Region B) 4 Support (Region A; 5 Support (Region B) 6 Hot Leg Water - 3 00 0 100 200 3 00 400 500 600 (0.86, 0.52, -0.00 ) -Axis

Westinghouse Non-Proprietary Class 3 2-25 WCAP-18843-NP August 2023 Revision 0 Figure 2-5 Reactor Geometry in r,TT at the Nozzles Centerline

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation) 400 300 200 100 0

100 200 X Axis 300 44 Core Barrel 45 Downcomer 47 Pressure Vessel Clad 48 Pressure Vessel 49 Cavity Air 50 Reflective Insulation 51 Bioshield 3 Outlet Plenum (Region N 55 Outlet Plenum (Region B) 6 Hot Leg Water 7 Cold Leg Water 400

Westinghouse Non-Proprietary Class 3 2-26 WCAP-18843-NP August 2023 Revision 0 Figure 2-6 Reactor Geometry in r,z Section View at 67qq Azimuthal Angle (Inlet Nozzle or Cold Leg)

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Fille_d Boun_dary 26 Core (Zone l) 300 27 Core (Zone 2) 28 Core (Zone 3) 29 Core (Zone 4) 30 Core (Zone 5) - 3 l Core (Zone 6) 32 Fuel Plenum 33 Top Water Gap 200 34 Top Nozzle 35 Bottom End Plugs 36 Bottom Water Gap 37 Bottom Nozzle Plate 8 Bottom Nozzle Leg 39 Core Baffle -40 LCS (Region A) 100 - 4 l LCS(Region B) Ill 42 Formers ""1 43 Bypass Region ~ 1 44 Core Barrel 45 Downcomer 0 47 Pressure Vessel Clad 0 48 Pressure Vessel '"'I 0 49 Cavity Air 50 Reflective Insulation 0 0 51 Bioshield 52 UCP (Region A) 0... 53 Outlet Plenum (Region A) 0 54 Inlet Plenum (Region A) 0. -100 55 Outlet Plenum (Region B) 0 57 Outlet Plenum (Region B) 58 UCP (Region B) 9 Upper Core Blanket 0 Lower Core Blanket 3 LCP (Region B) 4 Support (Region A) -200 5 Support Region B 7 Cold Leg Water 8 RPV Support -300 0 100 200 300 400 500 600 700 (0.39, 0.92, -0.00)-Axis

Westinghouse Non-Proprietary Class 3 2-27 WCAP-18843-NP August 2023 Revision 0 Figure 2-7 Reactor Geometry in r,z Section View at 22qq Azimuthal Angle (Outlet Nozzle or Hot Leg)

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Filled Boun_dary 26 Core (Zone 7) 300 27 Core (Zone 2) 28 Core (Zone 3) 29 Core (Zone 4) 0 Core (Zone 5) f-3 7 Core (Zone 6) 32 Fuel Plenum 3 Top Water Gap 200 4 Top Nozzle 35 Bottom End Plugs 6 Bottom Water Gap 37 Bottom Nozzle Plate 8 Bottom Nozzle Leg 39 Core Baffle 40 LCS (Region A) 100 41 LCS(Region B) Ill 42 Formers ""4 43 Bypass Region ~, 44 Core Barrel 45 Downcomer 0 47 Pressure Vessel Cla d 0 48 Pressure Vessel '"'I 0 49 Cavity Air 50 Reflective Insulation 0 0 51 Bioshield 52 UCP (Region A) 0... 53 Outlet Plenum (Region A) 0 54 Inlet Plenum (Region A) 0. -100 55 Outlet Plenum (Region B) 0 57 Outlet Plenum (Region B) 8 UCP (Region B) 59 Upper Core Blanket 0 Lower Core Blanket 3 LCP (Region B) 4 Supp ort (Region A) -200 5 Support Region B 6 Hot Leg Water -300 0 100 200 300 400 500 600 700 (0.93, 0.37, -0.00)-Axis

Westinghouse Non-Proprietary Class 3 3-1 WCAP-18843-NP August 2023 Revision 0 3 FRACTURE TOUGHNESS PROPERTIES The requirements for P-T limit curve development are specified in 10 CFR 50, Appendix G [Ref. 8]. 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 Catawba Unit 2 beltline materials traditionally included the intermediate and lower shell plates and welds; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. 9], 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 in the development of P-T limit curves. The additional materials that exceed this fluence threshold are referred to as the extended beltline materials and are evaluated to ensure that the applicable neutron embrittlement effects are considered. As seen from Table 2-6 of this report, the extended beltline materials include nozzle shell plates, nozzle shell longitudinal welds, and the nozzle to intermediate shell circumferential weld. The fluence for both the inlet and outlet nozzle to nozzle shell welds are less than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 54 EFPY. Therefore, the materials of the inlet and outlet nozzle forgings and the associated welds to the nozzle shell do not need to be considered in the extended beltline. Note that for reactor vessel welds, the terms girth and circumferential are used interchangeably; herein, these welds shall be referred to as circumferential welds. Similarly, for reactor vessel welds, the terms vertical and longitudinal are used interchangeably; herein, these welds shall be referred to as longitudinal welds. Although the reactor vessel nozzles are not a part of the extended beltline, per NRC RIS 2014-11, the nozzle materials must be evaluated for their potential effect on P-T limit curves due to the higher stresses in the nozzle corner region. These higher stresses can potentially result in 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. The effect of these higher stresses is addressed in Section 7, which determines that the Catawba Unit 2 beltline P-T limit curves generated in WCAP-15285 [Ref. 1] bound the inlet and outlet nozzle P-T limit curves. A summary of the best-estimate copper (Cu) and nickel (Ni) contents in units of weight percent (Wt. %), as well as initial RTNDT values, for the Catawba Unit 2 reactor vessel beltline and extended beltline materials are provided in Table 3-1. Table 3-2 provides the best-estimate Cu and Ni Wt. % values and initial RTNDT values for the reactor vessel nozzle forgings.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-2 WCAP-18843-NP August 2023 Revision 0 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Reactor Vessel Materials(a) Material Description Plate/Weld Seam Number Heat Number and Flux Type (Lot) Wt. % Cu Wt. % Ni RTNDT(U) (°F) I(d) (°F) Unirradiated USE (ft-lb) Beltline Materials Intermediate Shell Plates B8605-1 C-0543-1 0.082 0.618 15 0 89 B8605-2 C-0543-2 0.08 0.613 33 0 82 B8616-1 A-0617-1 0.045 0.595 12 0 92 Lower Shell Plates B8806-1 C-2288-1 0.057 0.56 6 0 83 B8806-2 C-2272-1 0.057 0.593 -10 0 102 B8806-3 C-2272-2 0.057 0.593 8 0 105 Intermediate Shell to Lower Shell Circumferential Weld 101-171 83648 Linde 0091 Lot 3536 0.04 0.12 -80 0 >130 Intermediate and Lower Shell Longitudinal Welds 101-124 A, B, C 101-142 A, B, C 83648 Linde 0091 Lot 3536 0.04 0.12 -80 0 >130 Extended Beltline Materials Closure Head Dome B8607-1 C-1936-2 0.13 0.64 -10 0 106 Closure Head Torus B8608-1 C-2085-1 0.07 0.62 -3 0 118 Closure Head Flange(b) B8601-1 122J147 0.7 10 0 152 Vessel Flange(b) B8602-1 125H320 0.71 10 0 175 Nozzle Shell B8604-1 C-0531-1 0.11 0.61 24 0 96 B8604-2 C-0531-2 0.11 0.61 26 0 89 B8604-3 C-0569-2 0.07 0.53 50 0 70 Nozzle Shell Longitudinal Welds 101-122 A, B, C 51912 Linde 0091 Lot 3490 0.156 0.059 -50 0 >112 Nozzle Shell to Intermediate Shell Weld 103-121 5P5622 Linde 0091 Lot 1122 0.153 0.077 -40 0 >102 Bottom Head Torus B8613-1 B-5572-1 0.14 0.48 -8 0 113 Bottom Head Dome B8612-1 B-5572-2 0.14 0.48 5 0 124 Surveillance Material Intermediate Shell Plate B8605-1 C-0543-1 0.082(c) 0.62(c) 15 0 89 Surveillance Weld 101-171 101-124 A, B, C 101-142 A, B, C 83648 Linde 0091 Lot 3536 0.04175(c) 0.153(c) -80 0 >130 Note(s): (a) Values in this table are taken from WCAP-17175-P [Ref. 11] (b) No Cu Wt. % values were given for these materials. (c) These values are the average of values taken from WCAP-15243 [Ref. 12] and restated in WCAP-15285 [Ref. 1]. (d) All RTNDT(U) values are based on measured data, therefore I = 0°F.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-3 WCAP-18843-NP August 2023 Revision 0 Table 3-2 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Reactor Vessel Inlet and Outlet Nozzle Materials(a)(d) Material Description Plate/Weld Seam Number Heat Number and Flux Type (Lot) Wt. % Cu(c) Wt. % Ni RTNDT(U) (°F) I(b) (°F) Unirradiated USE (ft-lb) Inlet Nozzle B8609-1 14211-1W 0.81 -20 0 119 B8609-2 14211-1W 0.78 -20 0 124 B8609-3 14256W 0.85 -20 0 109 B8609-4 14256W 0.85 37 0 97 Outlet Nozzle B8610-1 AV 8327 0.73 -10 0 141 B8610-2 AV 8332 0.78 -10 0 144 B8610-3 AV 8172 0.80 -20 0 140 B8610-4 AV 8148 0.80 -10 0 150 Note(s): (a) The inlet and outlet nozzle forgings are projected to be less than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at EOLE and do not need to be considered for the effects of embrittlement. However, the initial material properties are provided here for future reference. (b) All RTNDT(U) values are based on measured data with I = 0°F. (c) No Cu Wt. % values were given for the nozzle materials. (d) Values for this table taken from WCAP-17175-P [Ref. 11].

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-1 WCAP-18843-NP August 2023 Revision 0 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2 [Ref. 4], calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from surveillance programs at other plants which include Catawba Unit 2 reactor vessel beltline or extended beltline materials should also be considered when calculating Position 2.1 chemistry factors. The Catawba Unit 2 surveillance capsules contain plate material from Intermediate Shell Plate B8605-1 and the surveillance weld specimens were fabricated from weld wire Heat # 83648. Table 4-1 summarizes the surveillance data available for the shell plate and weld materials that will be used in the calculation of the Position 2.1 chemistry factor values. Per Appendix A, the surveillance data are deemed credible for Catawba Unit 2; therefore, a reduced margin term will be utilized in the ART calculations contained in Section 6 for the surveillance plate and weld materials.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-2 WCAP-18843-NP August 2023 Revision 0 Table 4-1 Surveillance Capsule Data Material Capsule(d) Capsule Fluence(b) (x 1019 n/cm2, E > 1.0 MeV) Measured RTNDT(a) (°F) Intermediate Shell Plate B8605-1 (Longitudinal) Z 0.303 24.68 X 1.14 48.07 V 2.20 44.76 Intermediate Shell Plate B8605-1 (Transverse) Z 0.303 29.55 X 1.14 52.17 V 2.20 61.29 Surveillance Weld (Heat # 83648 Linde 0091 Lot 3536) Z 0.303 0.00(c) X 1.14 34.94 V 2.20 58.80 Note(s): (a) Information extracted from WCAP-15243 [Ref. 12]. (b) The fluence values are taken from Table 2-13. (c) Per WCAP-15243 [Ref. 12], the actual measured value was -30.25 observed. Physically this should not occur; therefore, for conservatism this value was set to zero. (d) Capsule Y has been removed from the reactor vessel but is not included here as the material specimens were not tested. Only dosimetry was analyzed for this capsule.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-1 WCAP-18843-NP August 2023 Revision 0 5 CHEMISTRY FACTORS The chemistry factors (CFs) are calculated using Regulatory Guide 1.99, Revision 2 [Ref. 4], 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 reactor vessel materials are provided in Table 3-1 of this report. The Position 2.1 chemistry factor calculation is presented in Table 5-1 for the surveillance materials. These values were calculated using the surveillance data summarized in Section 4 of this report, which only includes surveillance program results. The calculations are performed using the method described in Regulatory Guide 1.99, Revision 2. Adjustment of the RTNDT values is required per Regulatory Guide 1.99, Revision 2 due to chemistry differences between the surveillance welds and the Catawba Unit 2 reactor vessel welds. The chemistry adjustment factor based on the differences between the Regulatory Guide 1.99, Position 1.1 CFs are determined below. Heat # 83648 CFBeltline Weld = 35.4°F CFSurv. Weld = 39.4°F Ratio = 35.4 y 39.4 = 0.90 The ratio procedure results in a ratio of less than 1.0; therefore, the chemistry adjustment can conservatively be neglected. From NRC Industry Meetings on November 12, 1997 and February 12 and 14, 1998, procedural guidelines were presented to adjust the RTNDT for temperature differences when using surveillance data from one reactor vessel applied to another reactor vessel. The following is taken from the handout [Ref. 13] given by the NRC at those industry meetings. Studies have shown that for temperatures near 550°F, a 1°F decrease in irradiation temperature will result in approximately a 1°F increase in RTNDT. Thus, for plants that use surveillance data from other reactor vessels that operate at a different temperature, or when the capsule is at a different temperature than the plant, then this difference must be considered. The temperature adjustment procedure is not applied when considering only surveillance data from the plant-specific program. The temperature adjustment is as follows: Temp. Adjusted RTNDT = RTNDT, Measured + (TCapsule - TPlant) (1) The temperature adjustment procedure is not applied when considering only surveillance data from the plant-specific program. Since only surveillance material from Catawba Unit 2 is utilized, no temperature adjustments are required. The Position 1.1 chemistry factors are summarized with the Position 2.1 chemistry factors in Table 5-2.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-2 WCAP-18843-NP August 2023 Revision 0 Table 5-1 Calculation of Chemistry Factors Using Surveillance Capsule Data(a) Material Capsule Capsule Fluence (x 1019 n/cm2, E > 1.0 MeV) FF(b) RTNDT (°F) FF* RTNDT (°F) FF2 Intermediate Shell Plate B8605-1 (Longitudinal) Z 0.303 0.673 24.68 16.61 0.453 X 1.14 1.037 48.07 49.83 1.075 V 2.20 1.214 44.76 54.33 1.473 Intermediate Shell Plate B8605-1 (Transverse) Z 0.303 0.673 29.55 19.88 0.453 X 1.14 1.037 52.17 54.08 1.075 V 2.20 1.214 61.29 74.39 1.473 SUM: 269.12 6.001 CFB8605-1 = (FF

• RTNDT) ÷ (FF2) = (269.12) ÷ (6.001) = 44.8°F Surveillance Weld (Heat # 83648 Linde 0091 Lot 3536)

Z 0.303 0.673 0 0 0.453 X 1.14 1.037 34.94 36.22 1.075 V 2.20 1.214 58.80 71.37 1.473 SUM: 107.59 3.001 CFSurv. Weld = (FF

• RTNDT) ÷ (FF2) = (107.59) ÷ (3.001) = 35.9°F Note(s):

(a) Fluence and measured RTNDT are taken from Table 4-1. (b) FF = fluence factor = f(0.28 - 0.10 log(f)).

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-3 WCAP-18843-NP August 2023 Revision 0 Table 5-2 Summary of Position 1.1 and 2.1 Chemistry Factors Material Description Plate/Weld Seam Number Heat Number and Flux Type (Lot) Chemistry Factor Position 1.1(a) (°F) Position 2.1(b) (°F) Beltline Materials Intermediate Shell Plates B8605-1 C-0543-1 52.4 44.8 B8605-2 C-0543-2 51.0 B8616-1 A-0617-1 28.5 Lower Shell Plates B8806-1 C-2288-1 35.2 B8806-2 C-2272-1 35.2 B8806-3 C-2272-2 35.2 Intermediate Shell to Lower Shell Circumferential Weld 101-171 83648 Linde 0091 Lot 3536 35.4 35.9 Intermediate Shell Longitudinal Welds 101-124 A, B, C 83648 Linde 0091 Lot 3536 35.4 35.9 Lower Shell Longitudinal Welds 101-142 A, B, C 83648 Linde 0091 Lot 3536 35.4 35.9 Extended Beltline Materials Closure Head Dome B8607-1 C-1936-2 92.0 Closure Head Torus B8608-1 C-2085-1 44.0 Closure Head Flange B8601-1 122J147 Vessel Flange B8602-1 125H320 Nozzle Shell B8604-1 C-0531-1 74.2 B8604-2 C-0531-2 74.2 B8604-3 C-0569-2 44.0 Nozzle Shell Longitudinal Welds 101-122 A, B, C 51912 Linde 0091 Lot 3490 73.7 Nozzle Shell to Intermediate Shell Weld 103-121 5P5622 Linde 0091 Lot 1122 74.1 Inlet Nozzle(c) B8609-1 14211-1W B8609-2 14211-1W B8609-3 14256W B8609-4 14256W Outlet Nozzle(c) B8610-1 AV 8327 B8610-2 AV 8332 B8610-3 AV 8172 B8610-4 AV 8148 Bottom Head Torus B8613-1 B-5572-1 94.6 Bottom Head Dome B8612-1 B-5572-2 94.6 Surveillance Materials Intermediate Shell Plate B8605-1 C-0543-1 51.0 Surveillance Weld 101-171 101-124 A, B, C 101-142 A, B, C 83648 Linde 0091 Lot 3536 39.2 Notes contained on next page.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-4 WCAP-18843-NP August 2023 Revision 0 Note(s): (a) Values for the beltline and extended beltline are calculated using Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 using the Cu and Ni weight percent values in Table 3-1. (b) Values are from Table 5-1. (c) No Cu Wt. % were given for these materials. These materials will not become limiting due to their low fluences as seen in Section 6 and therefore their respective CF does not need to be calculated.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-1 WCAP-18843-NP August 2023 Revision 0 6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [Ref. 4], the ART for each material in the beltline region is given by the following expression: ART = Initial RTNDT + RTNDT + Margin (2) 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 [Ref. 14]. 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 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: RTNDT = CF

• f(0.28 - 0.10 log (f)) (3)

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) = fsurf

• e(-0.24x) (4)

Where: x = depth into the vessel wall measured from the vessel inner surface in inches (reactor vessel cylindrical shell beltline thickness is 8.625 inches). The resultant fluence is then placed in Equation 3 to calculate the RTNDT at the specific depth. The projected reactor vessel neutron fluence was updated for this analysis and is documented in Section 2 of this report. Table 6-1 contains the surface fluence values at 54 EFPY, as well as 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 EOLE ART values for the Catawba Unit 2 reactor vessel materials. Margin is calculated as. The standard deviation for the initial RTNDT margin term (I) is 0°F when the initial RTNDT is a measured value. When a generic value is used, the I is obtained from the set of data used to establish the mean. The standard deviation for the RTNDT margin term () is 17°F for plates or forgings when surveillance data are not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data are used. For welds, is equal to 28°F when surveillance capsule data are not used or is non-credible and is 14°F (half the value) when credible surveillance capsule data are used. The value for need not exceed 0.5 times the mean value of RTNDT. Per Appendix A, the surveillance data for Intermediate Shell Plate B8605-1 and the surveillance weld data for Heat # 83648 are deemed credible.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-2 WCAP-18843-NP August 2023 Revision 0 Table 6-2 and Table 6-3 contain the 54 EFPY ART calculations at the 1/4T and 3/4T locations, respectively. The limiting ART values are summarized in Table 6-4. Table 6-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Reactor Vessel Materials at 54 EFPY Material Description Plate/Weld Seam Number Surface Fluence(a) (x 1019 n/cm2, E > 1.0 MeV) Surface FF(c) 1/4T Fluence(b) (x 1019 n/cm2, E > 1.0 MeV) 1/4T FF(c) 3/4T Fluence(b) (x 1019 n/cm2, E > 1.0 MeV) 3/4T FF(c) Beltline Materials Intermediate Shell Plates B8605-1 B8605-2 B8616-1 2.77 1.271 1.651 1.138 0.586 0.851 Lower Shell Plates B8806-1 B8806-2 B8806-3 2.75 1.270 1.639 1.136 0.582 0.849 Intermediate Shell to Lower Shell Circumferential Weld 101-171 2.75 1.270 1.639 1.136 0.582 0.849 Intermediate Shell Longitudinal Welds(d) 101-124 A, B, C 2.23 1.217 1.329 1.079 0.472 0.791 Lower Shell Longitudinal Welds(d) 101-142 A, B, C 2.21 1.215 1.317 1.077 0.468 0.788 Extended Beltline Materials Nozzle Shell B8604-1 B8604-2 B8604-3 0.0474 0.284 0.028 0.212 0.010 0.110 Nozzle Shell Longitudinal Welds(d) 101-122 A, B, C 0.0469 0.283 0.028 0.211 0.010 0.109 Nozzle Shell to Intermediate Shell Weld 103-121 0.0540 0.305 0.032 0.229 0.011 0.120 Inlet Nozzle B8609-1 0.0041 0.058 0.002 0.039 0.001 0.016 B8609-2 0.0041 0.058 0.002 0.039 0.001 0.016 B8609-3 0.0041 0.058 0.002 0.039 0.001 0.016 B8609-4 0.0041 0.058 0.002 0.039 0.001 0.016 Outlet Nozzle B8610-1 0.0020 0.032 0.001 0.021 0.0004 0.008 B8610-2 0.0020 0.032 0.001 0.021 0.0004 0.008 B8610-3 0.0020 0.032 0.001 0.021 0.0004 0.008 B8610-4 0.0020 0.032 0.001 0.021 0.0004 0.008 Note(s): (a) Fluence values are documented in Table 2-6. (b) The 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (8.625 inches), and equation f = fsurf*e-0.24(x) from Regulatory Guide 1.99, Revision 2 [Ref. 4], where x = depth into the vessel wall (inches). (c) FF = fluence factor = f(0.28 - 0.10 log(f)). (d) The fluence values shown are the highest values for each set of longitudinal welds, which is conservative.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-3 WCAP-18843-NP August 2023 Revision 0 Table 6-2 Adjusted Reference Temperature Evaluation for the Reactor Vessel Materials Through 54 EFPY at the 1/4T Location(a) Material Plate / Weld Seam Number R.G. 1.99, Rev. 2 Position CF(b) 1/4T Fluence(c) (x 1019 n/cm2, E > 1.0 MeV) 1/4T FF(c) RTNDT(U)(d) (°F) Predicted RTNDT (°F) I (°F) (e) (°F) M (°F) ART (°F) Beltline Materials Intermediate Shell Plate B8605-1 1.1 52.4 1.651 1.138 15 59.6 0 17.0 34.0 108.6 Using credible surveillance data(f) B8605-1 2.1 44.8 1.651 1.138 15 51.0 0 8.5 17.0 83.0 Intermediate Shell Plate B8605-2 1.1 51.0 1.651 1.138 33 58.0 0 17.0 34.0 125.0 Intermediate Shell Plate B8616-1 1.1 28.5 1.651 1.138 12 32.4 0 16.2 32.4 76.9 Lower Shell Plate B8806-1 1.1 35.2 1.639 1.136 6 40.0 0 17.0 34.0 80.0 Lower Shell Plate B8806-2 1.1 35.2 1.639 1.136 -10 40.0 0 17.0 34.0 64.0 Lower Shell Plate B8806-3 1.1 35.2 1.639 1.136 8 40.0 0 17.0 34.0 82.0 Intermediate Shell to Lower Shell Circumferential Weld 101-171 1.1 35.4 1.639 1.136 -80 40.2 0 20.1 40.2 0.4 Using credible surveillance data(f) 101-171 2.1 35.9 1.639 1.136 -80 40.8 0 14.0 28.0 -11.2 Intermediate Shell Longitudinal Welds 101-124 A, B, C 1.1 35.4 1.329 1.079 -80 38.2 0 19.1 38.2 -3.6 Using credible surveillance data(f) 101-124 A, B, C 2.1 35.9 1.329 1.079 -80 38.7 0 14.0 28.0 -13.3 Lower Shell Longitudinal Welds 101-142 A, B, C 1.1 35.4 1.317 1.077 -80 38.1 0 19.1 38.1 -3.8 Using credible surveillance data(f) 101-142 A, B, C 2.1 35.9 1.317 1.077 -80 38.7 0 14.0 28.0 -13.3 Extended Beltline Materials Nozzle Shell B8604-1 1.1 74.2 0.028 0.212 24 15.7 0 7.9 15.7 55.5 Nozzle Shell B8604-2 1.1 74.2 0.028 0.212 26 15.7 0 7.9 15.7 57.5 Nozzle Shell B8604-3 1.1 44.0 0.028 0.212 50 9.3 0 4.7 9.3 68.7 Nozzle Shell Longitudinal Welds 101-122 A, B, C 1.1 73.7 0.028 0.211 -50 15.5 0 7.8 15.5 -18.9 Nozzle Shell to Intermediate Shell Weld 103-121 1.1 74.1 0.032 0.229 -40 17.0 0 8.5 17.0 -6.1 Note(s) (continued on following page): (a) The Regulatory Guide 1.99, Revision 2 [Ref. 4] methodology was utilized in the calculation of the ART values. (b) CFs are taken from Table 5-2. (c) Fluence and FFs are taken from Table 6-1. (d) RTNDT(U) and I values are taken from Table 3-1.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-4 WCAP-18843-NP August 2023 Revision 0 (e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and 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, the weld metal = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT for either base metals or welds, with or without surveillance data. (f) The credibility evaluation in Appendix A of this report determined that the surveillance data for Intermediate Shell Plate B8605-1 (Heat # C-0543-1) and the Surveillance Weld (Heat # 83648) are deemed credible.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-5 WCAP-18843-NP August 2023 Revision 0 Table 6-3 Adjusted Reference Temperature Evaluation for the Reactor Vessel Materials Through 54 EFPY at the 3/4T Location(a) Material Plate / Weld Seam Number R.G. 1.99, Rev. 2 Position CF(b) 3/4T Fluence(c) (x 1019 n/cm2, E > 1.0 MeV) 3/4T FF(c) RTNDT(U)(d) (°F) Predicted RTNDT (°F) I (°F) (e) (°F) M (°F) ART (°F) Beltline Materials Intermediate Shell Plate B8605-1 1.1 52.4 0.586 0.851 15 44.6 0 17.0 34.0 93.6 Using credible surveillance data(f) B8605-1 2.1 44.8 0.586 0.851 15 38.1 0 8.5 17.0 70.1 Intermediate Shell Plate B8605-2 1.1 51.0 0.586 0.851 33 43.4 0 17.0 34.0 110.4 Intermediate Shell Plate B8616-1 1.1 28.5 0.586 0.851 12 24.2 0 12.1 24.2 60.5 Lower Shell Plate B8806-1 1.1 35.2 0.582 0.849 6 29.9 0 14.9 29.9 65.7 Lower Shell Plate B8806-2 1.1 35.2 0.582 0.849 -10 29.9 0 14.9 29.9 49.7 Lower Shell Plate B8806-3 1.1 35.2 0.582 0.849 8 29.9 0 14.9 29.9 67.7 Intermediate Shell to Lower Shell Circumferential Weld 101-171 1.1 35.4 0.582 0.849 -80 30.0 0 15.0 30.0 -19.9 Using credible surveillance data(f) 101-171 2.1 35.9 0.582 0.849 -80 30.5 0 14.0 28.0 -21.5 Intermediate Shell Longitudinal Welds 101-124 A, B, C 1.1 35.4 0.472 0.791 -80 28.0 0 14.0 28.0 -24.0 Using credible surveillance data(f) 101-124 A, B, C 2.1 35.9 0.472 0.791 -80 28.4 0 14.0 28.0 -23.6 Lower Shell Longitudinal Welds 101-142 A, B, C 1.1 35.4 0.468 0.788 -80 27.9 0 14.0 27.9 -24.2 Using credible surveillance data(f) 101-142 A, B, C 2.1 35.9 0.468 0.788 -80 28.3 0 14.0 28.0 -23.7 Extended Beltline Materials Nozzle Shell B8604-1 1.1 74.2 0.010 0.110 24 8.1 0 4.1 8.1 40.3 Nozzle Shell B8604-2 1.1 74.2 0.010 0.110 26 8.1 0 4.1 8.1 42.3 Nozzle Shell B8604-3 1.1 44.0 0.010 0.110 50 4.8 0 2.4 4.8 59.7 Nozzle Shell Longitudinal Welds 101-122 A, B, C 1.1 73.7 0.010 0.109 -50 8.0 0 4.0 8.0 -33.9 Nozzle Shell to Intermediate Shell Weld 103-121 1.1 74.1 0.011 0.120 -40 8.9 0 4.4 8.9 -22.2 Note(s) (continued on following page): (a) The Regulatory Guide 1.99, Revision 2 [Ref. 4] methodology was utilized in the calculation of the ART values. (b) CFs are taken from Table 5-2. (c) Fluence and FFs are taken from Table 6-1. (d) RTNDT(U) and I values are taken from Table 3-1.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-6 WCAP-18843-NP August 2023 Revision 0 (e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and 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, the weld metal = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT for either base metals or welds, with or without surveillance data. (f) The credibility evaluation in Appendix A of this report determined that the surveillance data for Intermediate Shell Plate B8605-1 (Heat # C-0543-1) and the Surveillance Weld (Heat # 83648) are deemed credible.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-7 WCAP-18843-NP August 2023 Revision 0 Table 6-4 Limiting ART Values at 54 EFPY Vessel Wall Location Limiting ART Values(a) (°F) Limiting Material 1/4T location 125 Intermediate Shell Plate B8605-2 3/4T location 110.4 Intermediate Shell Plate B8605-2 Note(s): (a) Values are the limiting values from Table 6-2 and Table 6-3.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-1 WCAP-18843-NP August 2023 Revision 0 7 P-T LIMIT CURVE APPLICABILITY EVALUATION The existing P-T limit curves, implemented in Technical Specifications Figures 3.4.3-1 and 3.4.3-2, were calculated in WCAP-15285 [Ref. 1] and are valid through 34 EFPY. However, additional P-T limit curves were developed in WCAP-15285 which are applicable up to 51 EFPY. In order to validate whether the 51 EFPY P-T limit curves are applicable through 54 EFPY, considering the new fluence values documented in this report, a comparison of the 54 EFPY limiting 1/4T and 3/4T ART values contained in Table 6-4 must be made with those used in WCAP-15285. This comparison is provided in Table 7-1. Table 7-1 Summary of the Limiting ART Values 1/4T Limiting ART (°F) 3/4T Limiting ART (°F) WCAP-15285 Table 6-4 WCAP-15285 Table 6-4 126 125 112 110.4 Table 7-1 shows that the limiting ART values at the 1/4T and 3/4T locations will not exceed the ART values used in the P-T limit curves of WCAP-15285. Furthermore, the limiting initial RTNDT value of 10°F for the closure head flange and the vessel flange remains appropriate, and the P-T limit curves flange notch and bolt-up temperature will be added to the 51 EFPY curves. Therefore, the 51 EFPY P-T limit curves generated in WCAP-15285 remain applicable through 54 EFPY with the updated fluence values. Inlet and Outlet Nozzles P-T Limit Curves NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. 9] 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 [Ref. 15] 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 [Ref. 16]) bound the generic nozzle P-T limit curves. This document has been approved by the NRC [Ref. 19] as an acceptable means to address the concerns of RIS 2014-11. The results and conclusions of PWROG-15109-NP-A are applicable as long as the plant-specific Catawba Unit 2 fluence of the nozzle corners remains less than the screening criterion of 4.28 x 1017 n/cm2 (E > 1.0 MeV), as described in PWROG-15109-NP-A. Section 2 of this report demonstrates Catawba Unit 2 adherence to this screening criterion; thus PWROG-15109-NP-A is applicable, and no further consideration of the nozzles is required with respect to P-T limits. In conclusion, PWROG-15109-NP-A demonstrates that the nozzles will not be limiting with respect to the P-T limit curves at Catawba Unit 2. Therefore, the concerns of RIS 2014-11 are adequately addressed.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-2 WCAP-18843-NP August 2023 Revision 0 Updated P-T Limit Curves The existing Catawba Unit 2 P-T limit curves in Technical Specifications Figures 3.4.3-1 and 3.4.3-2 are valid to 34 EFPY. However, another set of curves was created in WCAP-15285 [Ref. 1] that are valid to 51 EFPY. These curves were determined to be applicable to 54 EFPY as documented in this report. These curves were updated to account for the closure head/vessel flange requirements of 10 CFR Part 50, Appendix G since the original curves in WCAP-15285 did not account for this requirement. 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 percent 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 3-1. The limiting unirradiated RTNDT of 10°F is associated with the vessel flange and the closure head flange, so the minimum allowable temperature of this region is 130°F at pressures greater than 621 psig without margins for instrument uncertainties. The curves were updated to account for vacuum refill by decreasing the lowest data point from 0 psig to -14.7 psig. This will ensure that the Catawba Unit 2 P-T limit curves are valid at all times, even if pressure is taken below 0 psig during vacuum refill. The updated heatup and cooldown limit curves with their respective data tables are shown in Figure 7-1 and Figure 7-2, and Table 7-2 and Table 7-3, respectively, and are valid to 54 EFPY.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-3 WCAP-18843-NP August 2023 Revision 0 Figure 7-1 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60°, 80°, and 100°F/hr) Applicable for the First 54 EFPY (Without Margins for Instrumentation Errors) -50 200 450 700 950 1200 1450 1700 1950 2200 2450 0 50 100 150 200 250 300 350 400 450 500 550 Pressure (PSIG) Temperature (Deg. F) Unacceptable Operation Acceptable Operation Criticality Limit based on inservice hydrostatic test temperature (186ºF) for the service period up to 54 EFPY Heatup Rate 60ºF/Hr Heatup Rate 100ºF/Hr Critical Limit 60ºF/Hr Critical Limit 100ºF/Hr Leak Test Limit Boltup Temperature Lower limit for RCS pressure is 0 psia (-14.7 psig) Heatup Rate 80ºF/Hr Critical Limit 80ºF/Hr

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

I I

Westinghouse Non-Proprietary Class 3 7-4 WCAP-18843-NP August 2023 Revision 0 Figure 7-2 Catawba Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0°, 20°, 40°, 60° and 100°F/hr) Applicable for the First 54 EFPY (Without Margins for Instrumentation Errors) -50 200 450 700 950 1200 1450 1700 1950 2200 2450 0 50 100 150 200 250 300 350 400 450 500 550 Pressure (PSIG) Temperature (Deg. F) Unacceptable Operation Acceptable Operation Cooldown Rates: 0ºF/Hr -20ºF/Hr -40ºF/Hr -60ºF/Hr -100ºF/Hr Lower Limit for RCS pressure is 0 psia (-14.7 psig) -100ºF/Hr -60ºF/Hr -40ºF/Hr

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-5 WCAP-18843-NP August 2023 Revision 0 Table 7-2 54 EFPY Heatup Curve Data Points using the 1996 Appendix G (with Flange Requirements and without Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 80°F/hr Heatup 80°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) T (°F) P (psig) T (°F) P (psig) 55 -14.7 186 -14.7 55 -14.7 186 -14.7 55 -14.7 186 -14.7 55 621 186 719 55 621 186 719 55 621 186 719 60 621 186 760 60 621 186 757 60 621 186 757 65 621 186 753 65 621 186 746 65 621 186 743 70 621 186 752 70 621 186 739 70 621 186 733 75 621 186 754 75 621 186 736 75 621 186 726 80 621 186 761 80 621 186 737 80 621 186 722 85 621 186 771 85 621 186 741 85 621 186 721 90 621 186 784 90 621 186 748 90 621 186 723 95 621 186 802 95 621 186 758 95 621 186 728 100 621 186 822 100 621 186 771 100 621 186 736 105 621 186 846 105 621 186 788 105 621 186 746 110 621 186 874 110 621 186 807 110 621 186 760 115 621 186 906 115 621 186 831 115 621 186 776 120 621 186 941 120 621 186 857 120 621 186 796 125 621 190 982 125 621 190 888 125 621 190 819 130 621 195 1027 130 621 195 923 130 621 195 845 130 846 200 1077 130 788 200 962 130 746 200 875 135 874 205 1133 135 807 205 1006 135 760 205 909 140 906 210 1195 140 831 210 1055 140 776 210 948 145 941 215 1263 145 857 215 1109 145 796 215 991 150 982 220 1340 150 888 220 1170 150 819 220 1040 155 1027 225 1424 155 923 225 1237 155 845 225 1094 160 1077 230 1517 160 962 230 1312 160 875 230 1154 165 1133 235 1620 165 1006 235 1395 165 909 235 1221 170 1195 240 1734 170 1055 240 1487 170 948 240 1295 175 1263 245 1860 175 1109 245 1588 175 991 245 1376 180 1340 250 1998 180 1170 250 1700 180 1040 250 1467 185 1424 255 2151 185 1237 255 1824 185 1094 255 1567 190 1517 260 2321 190 1312 260 1961 190 1154 260 1678 195 1620 195 1395 265 2112 195 1221 265 1801 200 1734 200 1487 270 2278 200 1295 270 1936 205 1860 205 1588 275 2462 205 1376 275 2085 210 1998 210 1700 210 1467 280 2250 215 2151 215 1824 215 1567 285 2431 220 2321 220 1961 220 1678 225 2112 225 1801 230 2278 230 1936 235 2462 235 2085 240 2250 245 2431

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-6 WCAP-18843-NP August 2023 Revision 0 Table 7-3 54 EFPY Cooldown Curve Data Points using the 1996 Appendix G (with Flange Requirements and without 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) 55 -14.7 55 -14.7 55 -14.7 55 -14.7 55 -14.7 55 621 55 621 55 613 55 569 55 482 60 621 60 621 60 621 60 582 60 497 65 621 65 621 65 621 65 595 65 513 70 621 70 621 70 621 70 610 70 532 75 621 75 621 75 621 75 621 75 552 80 621 80 621 80 621 80 621 80 575 85 621 85 621 85 621 85 621 85 600 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 130 1018 130 1006 130 996 130 992 130 1001 135 1062 135 1053 135 1049 135 1050 140 1109 140 1106 140 1107 145 1162 150 1221 155 1285 160 1356 165 1435 170 1522 175 1618 180 1725 185 1842 190 1972 195 2116 200 2274 205 2449

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-1 WCAP-18843-NP August 2023 Revision 0 8 PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDLIENS 8.1 PRESSURIZED THERMAL SHOCK A limiting condition on RPV integrity known as 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 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 [Ref. 10]) 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 NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seals 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 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 [Ref. 4]. These accepted methods were used with the surface fluence values of Section 2 to calculate the following RTPTS values for the Catawba Unit 2 RPV materials. The RTPTS calculations are presented in Table 8-1 for Catawba Unit 2. PTS Conclusion All of the reactor vessel materials for Catawba Unit 2 are below the RTPTS screening criteria values of 270°F for plates, forgings, and longitudinal welds, and 300°F for circumferentially oriented welds (per 10 CFR 50.61) at 54 EFPY.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-2 WCAP-18843-NP August 2023 Revision 0 Table 8-1 RTPTS Calculations for Reactor Vessel Materials at 54 EFPY(a) Material Plate / Weld Seam Number R.G. 1.99 Rev. 2 Position CF(b) Surface Fluence(c) (x 1019 n/cm2, E > 1.0 MeV) Surf. FF(c) RTNDT(U)(d) (°F) Predicted RTNDT (°F) U (°F) (e) (°F) M (°F) RTPTS (°F) Beltline Materials Intermediate Shell Plate B8605-1 1.1 52.4 2.77 1.271 15 66.6 0 17.0 34.0 115.6 Using credible surveillance data(f) B8605-1 2.1 44.8 2.77 1.271 15 57.0 0 8.5 17.0 89.0 Intermediate Shell Plate B8605-2 1.1 51.0 2.77 1.271 33 64.8 0 17.0 34.0 131.8 Intermediate Shell Plate B8616-1 1.1 28.5 2.77 1.271 12 36.2 0 17.0 34.0 82.2 Lower Shell Plate B8806-1 1.1 35.2 2.75 1.270 6 44.7 0 17.0 34.0 84.7 Lower Shell Plate B8806-2 1.1 35.2 2.75 1.270 -10 44.7 0 17.0 34.0 68.7 Lower Shell Plate B8806-3 1.1 35.2 2.75 1.270 8 44.7 0 17.0 34.0 86.7 Intermediate Shell to Lower Shell Circumferential Weld 101-171 1.1 35.4 2.75 1.270 -80 44.9 0 22.5 44.9 9.9 Using credible surveillance data(f) 101-171 2.1 35.9 2.75 1.270 -80 45.6 0 14.0 28.0 -6.4 Intermediate Shell Longitudinal Welds 101-124 A, B, C 1.1 35.4 2.23 1.217 -80 43.1 0 21.5 43.1 6.2 Using credible surveillance data(f) 101-124 A, B, C 2.1 35.9 2.23 1.217 -80 43.7 0 14.0 28.0 -8.3 Lower Shell Longitudinal Welds 101-142 A, B, C 1.1 35.4 2.21 1.215 -80 43.0 0 21.5 43.0 6.0 Using credible surveillance data(f) 101-142 A, B, C 2.1 35.9 2.21 1.215 -80 43.6 0 14.0 28.0 -8.4 Extended Beltline Materials Nozzle Shell B8604-1 1.1 74.2 0.047 0.284 24 21.1 0 10.5 21.1 66.2 Nozzle Shell B8604-2 1.1 74.2 0.047 0.284 26 21.1 0 10.5 21.1 68.2 Nozzle Shell B8604-3 1.1 44 0.047 0.284 50 12.5 0 6.3 12.5 75.0 Nozzle Shell Longitudinal Welds 101-122 A, B, C 1.1 73.7 0.047 0.283 -50 20.8 0 10.4 20.8 -8.3 Nozzle Shell to Intermediate Shell Weld 103-121 1.1 74.1 0.054 0.305 -40 22.6 0 11.3 22.6

5.2 Note(s)

(a) The 10 CFR 50.61 [Ref. 10] methology was utilized in the calculation of the RTPTS values. (b) CFs are taken from Table 5-2. (c) Fluence and fluence factor values are taken from Table 6-1. (d) RTNDT(U) and I values are taken from Table 3-1. (e) Per 10 CFR 50.61, the base metal = 17°F when surveillance data are non-credible or not used to determine the CF, and = 8.5°F when credible surveillance data are used to determine the CF. Also, per 10 CFR 50.61, the weld metal = 28°F when surveillance data are non-credible or not used to determine the CF, and = 14°F when credible surveillance data are used to determine the CF. However, need not exceed 0.5*RTNDT. (f) The credibility evaluation for the Catawba Unit 2 surveillance data in Appendix A of this calculation determined that the Catawba Unit 2 surveillance data for the Intermediate Shell B8605-1 (Heat # C-0543-1) and the Surveillance Weld (Heat # 83648) are deemed credible.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-3 WCAP-18843-NP August 2023 Revision 0 8.2 EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION The emergency response guideline (ERG) limits, HF04BG [Ref. 17], were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event. 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 - now known as Pressurized Water Reactor Owners Group [PWROG]) to be applicable to all Westinghouse plants. The highest value of RTNDT, for which the generic category of ERG limits was developed, is 250°F for a longitudinal flaw and 300°F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250°F for a longitudinal flaw or 300°F 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 8.1 of this report. The material with the highest RTNDT defines the limiting material. Table 8-2 identifies ERG category limits and the limiting material RTNDT values at EOLE. Table 8-2 Evaluation of ERG Limit Category ERG Pressure-Temperature Limits [Ref. 17] Applicable RTNDT Value(a) ERG P-T Limit Category RTNDT < 200°F Category I 200°F < RTNDT < 250°F Category II 250°F < RTNDT < 300°F Category IIIb Limiting RTNDT Value(b) Limiting Reactor Vessel Material RTNDT Value @ EOLE Intermediate Shell Plate B8605-2 131.8 Note(s): (a) Longitudially oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F. (b) Values taken from Table 8-1. Per the ERG limit guidance document [Ref. 17], 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.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-4 WCAP-18843-NP August 2023 Revision 0 Conclusion of ERG P-T Limit Categorization Per Table 8-2, the limiting material assumes a longitudinally oriented flaw and has an RTNDT value that does not exceed 200°F at EOLE. Therefore, Catawba Unit 2 is limited to ERG Category I through EOLE and will not need plant-specific ERG P-T limits.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-1 WCAP-18843-NP August 2023 Revision 0 9 UPPER SHELF ENERGY EVALUATION Per Regulatory Guide 1.99, Revision 2, the Charpy upper-shelf energy is assumed to decrease as a function of fluence and copper content as indicated in Figure 2 of the Guide when surveillance data is not used. Linear interpolation is permitted. In addition, if surveillance data is to be used, the decrease in upper-shelf energy may be obtained by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure 9-1 of this report) 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) USE of the reactor vessel beltline materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the beltline materials and/or the results of the capsules tested to date using Figure 2 in Regulatory Guide 1.99, Revision 2. The reactor vessel beltline region thickness is 8.625 inches. 1/4T vessel fluence values for the beltline and extended beltline materials are taken from Table 6-1 for 54 EFPY. Table 9-1 provides the predicted 54 EFPY USE values.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-2 WCAP-18843-NP August 2023 Revision 0 Table 9-1 Predicted USE Values at 54 EFPY for the Reactor Vessel Beltline and Extended Beltline Materials Material Plate / Weld Seam Number Wt. % Cu(a) 1/4T Fluence (x 1019 n/cm2, E > 1.0 MeV)(b) Unirradiated USE (ft-lb)(a) Projected USE Decrease (%) Projected EOLE USE (ft-lb) Position 1.2(c) Intermediate Shell Plates B8605-1 0.082 1.651 89 22.0% 69 B8605-2 0.08 1.651 82 22.0% 64 B8616-1 0.045 1.651 92 22.0% 72 Lower Shell Plates B8806-1 0.057 1.639 83 22.0% 65 B8806-2 0.057 1.639 102 22.0% 80 B8806-3 0.057 1.639 105 22.0% 82 Intermediate Shell to Lower Shell Circumferential Weld 101-171 0.04 1.639 >130 22.0% >101 Intermediate Longitudinal Welds 101-124 A, B, C 0.04 1.329 >130 21.0% >103 Lower Longitudinal Welds 101-142 A, B, C 0.04 1.317 >130 21.0% >103 Nozzle Shell B8604-1 0.11 0.028 96 14.0% 83 B8604-2 0.11 0.028 89 14.0% 77 B8604-3 0.07 0.028 70 11.0% 62 Nozzle Shell Longitudinal Welds 101-122 A, B, C 0.156 0.028 >112 20.0% >90 Nozzle Shell to Intermediate Shell Weld 103-121 0.153 0.032 >102 20.0% >82 Position 2.2(c) Intermediate Shell B8605-1 0.082 1.651 89 7.0% 83 Intermediate Shell to Lower Shell Circumferential Weld 101-171 0.040 1.639 >130 11% >116 Intermediate Longitudinal Welds 101-124 A, B, C 0.040 1.329 >130 11% >116 Lower Longitudinal Welds 101-142 A, B, C 0.040 1.317 >130 11% >116 Note(s): (a) Copper weight percent values and unirradiated USE values were taken from Table 3-1 of this calculation note. If the base metal or weld Cu weight percentages are below the minimum value presented in Figure 2 of [Ref. 4] (0.1 for base metal and 0.05 for welds), then the Cu weight percentages were conservatively rounded up to the minimum value for projected USE decrease determination. (b) Values taken from Table 6-1 of this calculation note. 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 RG 1.99. (c) Position 1.2 percentage USE decrease values were calculated by plotting the 1/4T fluence values on RG 1.99, Figure 2 and using the material-specific Cu Wt. % values. The percent-loss lines were extended into the low fluence area of RG 1.99, Figure 2, i.e., below 1018 n/cm2, in order to determine the USE % decrease, as needed. Position 2.2 percentage USE decrease values were determined by drawing an upper-bound line parallel to the existing RG 1.99, Figure 2 lines through the applicable surveillance data points. These results should be used in preference to the existing graph lines for determining the decrease in USE, because the surveillance data is credible.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-3 WCAP-18843-NP August 2023 Revision 0 Figure 9-1 Regulatory Guide 1.99, Revision 2, Catawba Unit 2 Predicted Decrease in USE at 54 EFPY as a Function of Copper and Fluence 1 10 100 1E+17 1E+18 1E+19 1E+20 Percentage Drop in USE Neutron Fluence, n/cm2 (E > 1 MeV) Measured Plate Data for Intermediate Shell Plate B8605-1 (Heat # C-0543-1) Measured Data for Weld # 83648 Linde 0091 Lot 3536 % 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 Intermediate Shell Plate B8605-1 Line Limiting Plate Percent USE Decrease 7 % Capsule V Weld Heat # 83648 Line Limiting Weld Percent USE Decrease 10 % from Capsule X Intermediate Shell Longitudinal Weld 1/4T Fluence = 1.329E+19 n/cm2 Intermediate Shell 1/4T Fluence = 1.651E+19 n/cm2 Nozzle Shell to Intermediate Shell Weld 1/4T Fluence = 3.20E+17 n/cm2 Nozzle Shell 1/4T Fluence & Nozzle Shell Longitudinal Welds 1/4T Fluence = 2.80E+17 n/cm2 Lower Shell 1/4T Fluence & Intermediate Shell to Lower Shell Circumferential Weld 1/4T Fluence = 1.639E+19 n/cm2 Lower Shell Longitudinal Weld 1/4T Fluence = 1.317E+19 n/cm2

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

~ ~ I ~ -~ I ~ I I/ -~ \\ ~ \\ I/ -~ ~ , \\ \\ ~ \\ ~ ~ \\, \\ \\ \\ ~ ' \\ \\ \\ \\ \\ ,, 1 ~ \\ \\ \\ \\ \\ - ~ ~ \\ \\ \\ \\ \\ t11 \\ \\ \\ \\ ' \\ I \\ ' \\ \\ \\ \\ \\ -\\-\\ \\ \\ \\ \\ \\ \\ I \\ \\ \\ \\ \\ \\ \\ \\ I \\, \\ \\ \\ \\ \\ \\ ' \\ 0 ~ \\\\ \\ \\ ~ \\\\ \\ \\ \\ I \\ ~~ \\ ~-- \\ \\1\\ \\\\ \\ \\ \\ ~-- \\ - ~ \\ \\ \\ \\ \\ \\ \\ ~-- \\ \\ \\ \\ \\ \\ \\ \\ - ~ ~-- \\ \\ \\ \\ \\, \\ \\ - ~ ~-- l \\ ' \\ \\ \\ - ~ ~-- / \\ \\ \\ \\ \\ - ~ \\ \\ \\ \\ \\ ~ -~ \\ \\ \\ \\ I ...... \\-\\\\1 I \\ \\ ~ I \\ \\ \\ \\.\\ ' \\ \\ \\ \\ \\ \\ \\\\\\ \\ \\ I

Westinghouse Non-Proprietary Class 3 10-1 WCAP-18843-NP August 2023 Revision 0 10 REFERENCES

1. Westinghouse Report WCAP-15285, Revision 0, Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation using Code Case N-640, September 1999.

2. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Revision 0, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [Agencywide Documents Access and Management System [ADAMS] Accession Number ML010890301]

3. Westinghouse Report WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018. [ADAMS Accession Number ML18204A010]

4. 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]

5. Radiation Safety Information Computational Center (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, Oak Ridge National Laboratory (ORNL), July 1999.

6. ORNL Report ORNL/TM-13205, Pool Critical Assembly Pressure Vessel Facility Benchmark, (NUREG/CR-6454), July 1997.

7. ORNL Report ORNL/TM-13204, H.B.

Robinson-2 Pressure Vessel Benchmark, (NUREG/CR-6453), October 1997 (published February 1998).

8. Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, November 29, 2019.

9. 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 No. ML14149A165]

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. Westinghouse Report WCAP-17175-P, Revision 0, Catawba Units 1 and 2 Reactor Vessel Integrity Programs Plans, January 2011.

12. Westinghouse Report WCAP-15243, Revision 0, Analysis of Capsule V and the Capsule Y Dosimeters from the Duke Energy Catawba Unit 2 Reactor Vessel Surveillance Program, August 1999.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 10-2 WCAP-18843-NP August 2023 Revision 0

13. 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]

14. ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division 1, Subarticle NB-2300, Fracture Toughness Requirements for Material.

15. Pressurized Water Reactor Owners Group (PWROG) PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]

16. 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]

17. 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 2014.

18. Westinghouse Report WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, May 2022. (Report Accession No. ML22153A139, Verification Letter Accession No. ML22200A029).

19. NRC Safety Evaluation Final Safety Evaluation by the Office of Nuclear Reactor Regulation for Pressurized Water Reactor Owners Group Topical Report PWROG-15109-NP, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation EPID L-2018-TOP-0009, October 31, 2019. [ADAMS Accession Numbers ML19301D063 & ML19301D160]

20. Combustion Engineering Drawing, E-8871-171-005, Rev. 1, General Arrangement Plan, Westinghouse Electric Corp. 173 I.D. P.W.R.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-1 WCAP-18843-NP August 2023 Revision 0 APPENDIX A CREDIBILITY RE-EVALUATION OF THE CATAWBA UNIT 2 SURVEILLANCE PROGRAM Regulatory Guide 1.99, Revision 2 [Ref. A-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. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe 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 Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question. To date, three capsules have been tested from the Catawba Unit 2 reactor vessel. Table A-1 reviews the five criteria of Regulatory Guide 1.99, Revision 2. The following subsections evaluate each of these five criteria for Catawba Unit 2 in order to determine the credibility of the surveillance data for use in neutron radiation embrittlement calculations. Table A-1 Identification of Credibility Criteria Affected by Fluence Change Criterion No. Description 1 Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement. 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. 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 [Ref. A-4]. 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. 5 The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-2 WCAP-18843-NP August 2023 Revision 0 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 50, Fracture Toughness Requirements, [Ref. A-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. The Catawba Unit 2 reactor vessel consists of the following beltline region materials: x Intermediate Shell Plates B8605-1, B8605-2, and B8616-1 (Heat # C-0543-1, C-0543-2, and A-0617-1) x Lower Shell Plates B-8806-1, B8806-2, and B8806-3 (Heat # C-2288-1, C-2272-1, and C-2272-2) x Intermediate Shell Plate Longitudinal Weld Seams 101-124 A, B, C (fabricated with Type B4 Weld Wire, Heat # 83648, Flux Type Linde 0091, Lot # 3536) x Lower Shell Plate Longitudinal Weld Seams 101-142 A, B, C (fabricated with Type B4 Weld Wire, Heat # 83648, Flux Type Linde 0091, Lot # 3536) x Intermediate to Lower Shell Plate Circumferential Weld Seam 101-171 (fabricated with Type B4 Weld Wire, Heat # 83648, Flux Type Linde 0091, Lot # 3536) At the time the Catawba Unit 2 surveillance program material was developed, and as documented in WCAP-15243 [Ref. A-5], Intermediate Shell Plate B-8605-1 was judged to be the most limiting and was therefore utilized in the surveillance program. The Catawba Unit 2 surveillance program weld was fabricated with type B4 weld wire heat number 83648, Flux Type Linde 0091, Lot Number 3536. The Catawba Unit 2 Intermediate and Lower Shell Longitudinal Welds and the Intermediate Shell to Lower Shell Circumferential Weld Seams were also fabricated with Weld Wire Type B4 Weld Wire Heat # 83648, Flux Type Linde 0091, Lot Number 3536. Hence, the surveillance weld is representative of the beltline welds. Based on the discussion, Criterion 1 is met for the Catawba 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 analysis report, WCAP-15243 [Ref. A-5]. Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the surveillance materials unambiguously.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-3 WCAP-18843-NP August 2023 Revision 0 Thus, the Catawba Unit 2 surveillance program meets Criterion 2. 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 [Ref. A-4]. 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 plates. Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 [Ref. A-3]. At this meeting, the NRC presented five cases. Of the five cases, Case 1 (Surveillance Data Available from Plant but No Other Source) most closely represents the situation for the Catawba Unit 2 surveillance plate and weld materials. Evaluation of the Catawba Unit 2 Data Only (Case 1) Following the NRC Case 1 guidelines, the Catawba Unit 2 surveillance plates and weld metal (Heat # 83648) will be evaluated using the Catawba Unit 2 data. Table A-2 provides the calculation of the interim CF for Catawba Unit 2. Note that when evaluating the credibility of the surveillance weld data, the measured RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory 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 Catawba Unit 2 data are being considered; therefore, no temperature adjustment is required.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-4 WCAP-18843-NP August 2023 Revision 0 Table A-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Catawba Unit 2 Surveillance Capsule Data Only Material Capsule Capsule Fluence(a) (x 1019 n/cm2, E > 1.0 MeV) FF(b) RTNDT(a) (°F) FF*RTNDT (°F) FF2 Intermediate Shell Plate B8605-1 (Longitudinal) Z 0.303 0.673 24.68 16.61 0.453 X 1.14 1.037 48.07 49.83 1.075 V 2.20 1.214 44.76 54.33 1.473 Intermediate Shell Plate B8605-1 (Transverse) Z 0.303 0.673 29.55 19.88 0.453 X 1.14 1.037 52.17 54.08 1.075 V 2.20 1.214 61.29 74.39 1.473 SUM: 269.12 6.001 CFB8605-1 = (FF

• RTNDT) ÷ (FF2) = (269.12) ÷ (6.001) = 44.8°F Surveillance Weld (Heat # 83648 Linde 0091 Lot 3536)

Z 0.303 0.673 0 0 0.453 X 1.14 1.037 34.94 36.22 1.075 V 2.20 1.214 58.80 71.37 1.473 SUM: 107.59 3.001 CFSurv. Weld = (FF

• RTNDT) ÷ (FF2) = (107.59) ÷ (3.001) = 35.9°F Note(s):

(a) Fluence and measured RTNDT taken from Table 4-1. (b) FF = fluence factor = f(0.28 - 0.10 log(f)).

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-5 WCAP-18843-NP August 2023 Revision 0 The scatter of 'RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2 Position 2.1 is presented in Table A-3. Table A-3 Surveillance Capsule Data Scatter About the Best-Fit Line Using Only Catawba Unit 2 Surveillance Data Material Capsule CF(a) (Slopebest-fit) (°F) Capsule Fluence(b) (x 1019 n/cm2, E > 1.0 MeV) FF(c) Measured RTNDT(b) (°F) Predicted RTNDT(d) (°F) Scatter RTNDT(e) (°F) < 17°F (Base Metal) < 28°F (Weld) Intermediate Shell Plate B8605-1 (Longitudinal) Z 44.8 0.303 0.673 24.68 30.14 5.46 Yes X 44.8 1.14 1.037 48.07 46.44 1.63 Yes V 44.8 2.20 1.214 44.76 54.38 9.62 Yes Intermediate Shell Plate B8605-1 (Transverse) Z 44.8 0.303 0.673 29.55 30.14 0.59 Yes X 44.8 1.14 1.037 52.17 46.44 5.73 Yes V 44.8 2.20 1.214 61.29 54.38 6.91 Yes Surveillance Weld (Heat # 83648 Linde 0091 Lot 3536) Z 35.9 0.303 0.673 0 24.16 24.16 Yes X 35.9 1.14 1.037 34.94 37.21 2.27 Yes V 35.9 2.20 1.214 58.80 43.58 15.22 Yes Note(s): (a) CFs calculated in Table A-2. (b) Fluence and measured RTNDT values are taken from Table 5-1. (c) FF = fluence factor = f(0.28 - 0.10 log (f)). (d) Predicted RTNDT = CF x FF. (e) Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT]. From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table A-3 indicates that 100% of surveillance data points fall inside the +/- 1V of 17qF scatter band for surveillance base metals. Therefore, all the surveillance plate data are deemed credible per the third criterion. Table A-3 indicates that all surveillance data points fall inside the +/- 1V of 28qF scatter band for surveillance weld materials. 100% of the data are bounded; therefore, the surveillance weld data are deemed credible per the third criterion. 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. The capsule specimens are located in the reactor between the core barrel and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pads. 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 such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-6 WCAP-18843-NP August 2023 Revision 0 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 Catawba Unit 2 surveillance program does not include correlation monitor material. Therefore, this criterion is not applicable to Catawba Unit 2. CONCLUSION: Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2 [Ref. A-1], Section B, the Catawba Unit 2 surveillance data are credible.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-7 WCAP-18843-NP August 2023 Revision 0 REFERENCES A-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988. [ADAMS Accession Number ML003740284] A-2 Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019. A-3 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] A-4 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society for Testing and Materials, 1982. A-5 Westinghouse Report WCAP-15243, Revision 0, Analysis of Capsule V and the Capsule Y Dosimeters from the Duke Energy Catawba Unit 2 Reactor Vessel Surveillance Program, August 1999.

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation)

WCAP-18843-NP Revision 0 Proprietary Class 2 (R)

• • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Ziegler Tyler Aug-02-2023 11:53:53 Author Approval Miralles Ferrete Arturo Aug-07-2023 06:50:32 Verifier Approval Long Margaret L Aug-07-2023 07:22:44 Verifier Approval McNutt Don Aug-07-2023 09:49:36 Verifier Approval Geer Jared L Aug-07-2023 12:47:11 Manager Approval Patterson Lynn Aug-07-2023 13:11:44 Manager Approval Klingensmith Jesse J Aug-08-2023 10:10:19 Hold to Release Approval Ziegler Tyler Aug-08-2023 10:15:13 Files approved on Aug-08-2023

• • • This record was final approved on 8/8/2023, 10:15:13 AM. (This statement was added by the PRIME system upon its validation) to RA-23-0245 Westinghouse Report WCAP-15285, Revision 0, Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation using Code Case N-640.

Westinghouse Non-Proprietary Gass 3 Catawba Unit 2 Heatup and Cooldo-wn Curves for Norn1al Operation Using Code Case N-640

• W e s t i n g h o u s e E I e *c t r i c C o m p a n y L L C

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15285 Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 T. J. Laubham October 1999 Work Performed Under Shop Order DJSP-139 Prepared by the Westinghouse Electric Company LLC for the Duke Energy Company o:\\4468(cust).doc:lb-100699 Approved: (! !:JI~ C. H. Boyd, Manager Engineering and Materials Technology Approved:d~ D. M. Trombola, Manager Mechanical Systems Integration Westinghouse Electric Company LLC P.O. Box355 Pittsburgh, PA 15230-0355 ©1999 Westinghouse Electric Company LLC All Rights Reserved

PREFACE This report has been technically reviewed and verified by: E. Terek __ {_. _&-z--A,_--

ii TABLE OF CONTENTS LIST OF TABLES.................................................................................................................................. iii LIST OF FIGURES................................................................................................................................. V EXECUTIVE

SUMMARY

.................................................................................................................... vii 1 INTRODUCTION...................................................................................................................... I 2 PURPOSE.................................................................................................................................. 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS................... 3 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE......................................... 11 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES....................... 25 6 REFERENCES......................................................................................................................... 45

Table 1 Table 2 Table 3 Table 4 Table 5 Table 6 Table 7 Table 8 Table 9 Table 10 Table 11 Table 12 Table 13 Table 14 Table 15 Table 16 Table 17 Table 18 iii LIST OF TABLES Summary of the Peak Pressure Vessel Surface Neutron Fluence Values used for the Calculation of the ART Values....................................................................................... 12 Measured Integrated Neutron Exposure of the Catawba Unit 2 Surveillance Capsules Tested To Date............................................................................................... 12 Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data........ 13 Reactor Vessel Beltline Material Unirradiated Toughness Properties............................... 14 Calculation of Chemistry Factors using Catawba Unit 2 Surveillance Capsule Data....... 15 Summary of the Catawba Unit 2 Reactor Vessel Beltline Material Chemistry Factors..... 16 Summary of the Calculated Fluence Factors used for the Generation of the 22, 34 and 51 EFPY Heatup and Cooldown Curves.................................................................. 17 Calculation of the ART Values for the 1/4 T Location @ 22 EFPY.................................. 18 Calculation of the ART Values for the 3/4T Location@ 22 EFPY.................................. 19 Calculation of the ART Values for the l/4T Location @ 34 EFPY.................................. 20 Calculation of the ART Values for the 3/4T Location@34 EFPY.................................. 21 Calculation of the ART Values for the l/4T Location@51 EFPY.................................. 22 Calculation of the ART Values for the 3/4T Location @ 51 EFPY.................................. 23 Summary of the Limiting ART Values Used in the Generation of the Catawba Unit 2 Heatup/Cooldown Curves................................................................. :............................ 24 22 EFPY Heatup Curve Data Points Using 1996 App. G (Without Uncertainties for Instrumentation Errors)............................................................. 39 22 EFPY Cooldown Curve Data Points Using 1996 App. G (Without Uncertainties for Instrumentation Errors)............................................................. 40 34 EFPY Heatup Curve Data Points Using 1996 App. G (Without Uncertainties for Instrumentation Errors)............................................................. 41 34 EFPY Cooldown Curve Data Points Using 1996 App. G (Without Uncertainties for Instrumentation Errors)............................................................. 42

Table 19 Table 20 LIST OF TABLES - (Continued) 51 EFPY Heatup Curve Data Points Using 1996 App. G iv (Without Uncertainties for Instrumentation Errors)............................................................. 43 51 EFPY Cooldown Curve Data Points Using 1996 App. G (Without Uncertainties for Instrumentation Errors)............................................................. 44

Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Figure 6 Figure 7 Figure 8 Figure 9 Figure 10 Figure 11 V LIST OF FIGURES Geometry of the Upper Head/Flange Region of a Typical Westinghouse Four Loop Plant Reactor Vessel............................................................................................... 8 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Closure Head to Flange Region Weld............................................................................... 9 Determination of Boltup Requirement, Using K1c........................................................... 10 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F /hr) Applicable for the First 22 EFPY (Without Margins for Instrumentation Errors)................................................................ 27 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of80°F/hr) Applicable for the First 22 EFPY (Without Margins for Instrumentation Errors)................................................................ 28 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for the First 22 EFPY (Without Margins for Instrumentation Errors)................................................................ 2 9 Catawba Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 22 EFPY (Without Margins for Instrumentation Errors)................................................................ 30 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors)................................................................ 31 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 80°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors)...................................-............................. 3 2 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors)................................................................ 3 3 Catawba Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors)................................................................ 34

Figure 12 Figure 13 Figure 14 Figure 15 LIST OF FIGURES - (Continued) Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for the First 51 EFPY vi (Without Margins for Instrumentation Errors)................................................................ 3 5 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 80°F/hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)................................................................ 36 Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of l00°F/hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)................................................................ 3 7 Catawba Unit 2 Reactor Coolant System CooldoYm Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)................................................................ 3 8

EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature limit curves for nonnal operation of the Catawba Unit 2 reactor vessel. These curves were generated based on the latest available reactor vessel infonnation (Capsule V analysis, WCAP-15243111). vii The Catawba Unit 2 heatup and cooldown pressure-temperature limit curves have been updated based on the use of the ASME Code Case N-640131, "Alternative Reference Fracture Toughness For Development of P-T Limit Curves for Section XI, Division l ", which allows the use of the Kie methodology and a relaxation of the reactor vessel flange temperature requirement.

1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RT NDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RT NDT of the limiting material in the core region of the reactor vessel is detennined by using the 1 unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ~T NDT, and adding a margin. The unirradiated RT NDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDIT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F. RT NDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT NDT at any time period in the reactor's life, ~T NDT due to the radiation exposure associated with that time period must be added to the unirradiated RT NDT (IRT NDT). The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide l. 99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials. "141 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRT NDT + ~T NDT + margins for uncertainties) at the l/4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured from the clad/base metal interface. The most limiting ART values are used in the generation ofheatup and cooldown pressure-temperature limit curves. WCAP-15285

2 2 PURPOSE The Duke Energy Company contracted Westinghouse to analyze capsule V from the Catawba Unit 2 reactor vessel. As part of this analysis Westinghouse generated new heatup and cooldown curves for 22, 34 and 51 EFPY using K1c in place of KIR for the calculation of the stress intensity factors. The heatup and cooldown curves were generated without margins for instrumentation errors and included a hydrostatic leak test limit curve from 2485 to 2000 psig and pressure-temperature limits for the vessel flange regions per the requirements of 10 CFR Part 50, Appendix G151. The purpose of this report is to present the calculations and the development of the Duke Energy Company Catawba Unit 2 heatup and cooldown curves for 22, 34 and 51 EFPYutilizing the K1c methodology131. In addition, this report provides justification for relaxing the reactor vessel flange requirements of Appendix G to 10CFR Part 50 based on the use of K1c methodology rather than the Krr methodology. The use ofK1c and relaxation of the reactor vessel flange requirements will add substantial pressure margin to the heatup and cooldown curves. This increase in allowable pressure is presented in Section 5 of this report. WCAP-15285

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach 3 The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K1c is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division l ", of the ASME Appendix G to Section XI13 & 61. The Kie curve is given by the following equation:

where, K1c ;;;;;JJ.2+20.734 *e10*02 <T-RTNDT)l (1) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT This K1c 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, SA-508-3 steel.

3.2 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:

where, K1m K1t K1c C

C = = = = = stress intensity factor caused by membrane (pressure) stress stress intensity factor caused by the thermal gradients function of temperature relative to the RT NDT of the material

2. 0 for Level A and Level B service limits (2) 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical For membrane tension, the corresponding K1 for the postulated defect is:

K Im = Mm X (pR / I) (3) WCAP-15285

where, Mm for an inside surface flaw is given by: Mm = 1.85 for.Ji < 2, Mm = 0.926.Ji for 2:::;.Ji::; 3.464, Mm = 3.21 for.Ji > 3.464 Similarly, Mm for an outside surface flaw is given by: Mm = 1.77 for.Ji < 2, Mm = 0.893.Ji for 2 ~.Ji ~ 3.464, Mm = 3.09 for.Ji > 3.464 and p = internal pressure, Ri ""vessel inner radius, and t = vessel wall thickness. For bending stress, the corresponding K1 for the postulated defect is: Kib = Mb

• Maximum Stress, where Mb is two-thirds of Mm The maximum K1 produced by radial thermal gradient for the postulated inside surface defect of G-2120 is K11 = 0.953x10*3 x CR x t2\\ where CR is the cooldown rate in °F/hr., or for a postulated outside surface defect, K11 = 0.753xl0*3 x HU x t2*5, where HU is the heatup rate in °F/hr.

4 The through-wall temperature difference associated with the maximum thermal K1 can be detennined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal K1. (a) The maximum thermal K1 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 K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship: Ki,= (1.0359Co + 0.6322C1 + 0.4753C2 + 0.3855C3)

• J"ia (4) or similarly, Krr during heatup for a 1/4-thickness outside surface defect using the relationship:

/Gr= (1.043Co+0.630C1 +0.481C2+0.401C3)*/;o (5) WCAP-15285

where the coefficients C0, Ci, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form: (6) and xis a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth. Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040[81, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above. At any time during the heatup or cooldown transient, K1c is determined by the metal temperature at the tip of a postulated flaw at the l/4T and 3/4T location, the appropriate value for RT NDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K1t, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. 5 For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to 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 wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. 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 l/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.1.T (temperature) developed during cooldown results in a higher value of K1c at the l/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1c exceeds K11, 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 l/4T location and, therefore, allowable pressures may unknowingly be violated 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. WCAP-15285

Three separate calculations are required to detennine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a l/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 Krc for the l/4T crack during heatup is lower than the Krc for the l/4T crack 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 Krc values do not offset each other, and the pressure-temperature 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 second portion of the heatup analysis concerns the calculation of the pressure-temperature 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 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 pressure-temperature curves for both 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 lesser 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. 3.3 Closure Head/Vessel Flange Requirements IO CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions. _This rule states that the metal temperature of the closure flange regions must*exceed the material unirradiated RT NDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3 l 07 psi), which is 621 psig for Catawba Unit 2 reactor vessel. This requirement was originally based on concerns about the fracture margin in the closure flange region. During the boltup process. stresses in this region typically reach over 70 percent of the steady-state stress without being at steady-state temperature. The margin of l 20°F and the pressure limitation of 20 percent ofhydrotest pressure were developed using Kia fracture toughness, in the mid l 970's. Improved knowledge of fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of K1c in development of pressure-temperature curves, as contained in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division l ". The following discussion uses a similar approach (i.e. using K1c) here to develop equivalent reactor vessel flange temperature requirements. WCAP-15285 6

The geometry of the closure head flange region for a typical Westinghouse four loop plant reactor vessel is shown in Figure 1. The stresses in this region are highest near the outside of the head. Therefore, an outside reference flaw of 25 percent of the wall thickness parallel to the dome to flange weld (i.e. in the direction of the welding) was postulated in this region. To be consistent with ASME Section Xl, Appendix G, a safety factor of two was applied and a fracture calculation performed. 7 Figure 2 shows the crack driving force or stress intensity factor for the postulated flaw in this region, along with a second curve which incorporates the safety factor of two. Note that the stress intensity factor with a safety factor of one for this region does not exceed 55 ksiin., even for postulated flaws up to 50 percent of the wall thickness. For reference flaw, with the safety factor of two, the applied stress intensity factor is 85.15 ksiin. at 25 percent of the wall thickness. The determination of the boltup, or reactor vessel flange requirement, is shown in Figure 3, where the fracture toughness is plotted as a function of the temperature. In this figure, the intersection between the stress intensity factor curve and the Kia toughness curve occurs at a value slightly higher than T - RT NDT = I 00°F, which is in the range of the existing 120°F requirement. The reference calculation used for the original requirement (which is no longer available) must have had a slightly higher value ofK, about 98 ksiin. Note that the use ofK1c curve to determine this requirement results in a revised requirement ofT-RTNDT = 45°F, as seen in Figure 3. Therefore, the appropriate flange requirement for use with the Krc curve is as follows: The pressure in the vessel should not exceed 20 percent of the pre-service hydro-test pressure until the temperature exceeds T-RT NOT= 45°F. This requirement has been implemented with the curves presented in this report. The limiting unirradiated RT NOT of l 0°F occurs in the closure head and vessel flanges of the Catawba Unit 2 reactor vessel, so the minimum allowable temperature of this region is 55°F at pressures greater than 621 psig with no margins for instrument uncertainties. WCAP-15285

205.00 0 I A---a..i 191.875 DIA -E-" TOP HEAD DOME TORUS TO FLANGE WELD 8 NOTE: ALL DIMENSIONS ARE IN INCHES 167.00 DI A ----til..t 167.00 DIA __ 27.625 VESSEL FLANGE TO J UPPER SHELL WELD-~....,._*----~ UPPER HEAD REGION NOTEr DIMENSIONS 00 NOT INCLUDE CLADDING Figure I Geometry of the Upper Head/Flange Region of a Typical Westinghouse Four Loop Plant Reactor Vessel WCAP-15285

9 160~--------------------------------------, 140 _ 120 .5 ~ Ba ltup ISF -21 Ki I@ T-R lndt - '2D] -9B.3 fo riginal rqui,.m fttl 1100~--------------------,,,......=-----~ S BO ~ ~

l!

60 s = Ill Ill Q) 40 20 Kie@ 114T -85.15 \\ \\ i--1 B.:--o-,,-ltu-p-,,:IS,...F -_:--II 0+-----1----+---+---+-----1------+---+----f------1---+----+-- -+------1-------, 0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 0.55 0.60 0.65 Raw ~pth (alt) for Outside Surface Raw Figure 2 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Closure Head to Flange Region Weld WCAP-15285 0.70

10 250 T"'"""--------------------------------------, 200 1 ~ i: "~ i\\- 150 l ! I: ~.. ~ 100 ~ i! ~ I!... 50 ~ K,. 85.15 0 0 50 100 150 200 250 T-RTndt (Deg. F) Figure 3 Determination ofBoltup Requirement, Using K1c WCAP-15285

11 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ARn for each material in the beltline region is given by the following expression: ART== Initial RT NDT + dRT NDT + Margin (7) Initial RT NDT 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[71. If measured values of initial RT NDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class. ~T NDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows: ~T NDT :;;;: CF

• f<0.28
• 0.10 log f)

To calculate ~T NDT at any depth (e.g., at l/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth. f". _ f (-0.24><) J.(depth ><) - swface e (8) (9) where x inches (vessel beltline thickness is 8.625 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ~T NDT at the specific depth. The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections as a part ofWCAP-15243 and are also presented in a condensed version in Table l of this report. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"[3l_ Table I contains the calculated vessel surface fluences values along with the Regulatory Guide 1.99, Revision 2, l/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the Catawba Unit 2 reactor vessel. Additionally, the surveillance capsule fluence values are presented in Table 2. WCAP-15285

TABLE I Summary of the Peak Pressure Vessel Neutron Fluence Values used for the Calculation of ART Values EFPY Surface 1/4T 3/4T 22 1.3) X )019 7.8} X 1018 2.77 X 1018 34 2.01 X 1019 1.20 X 1019 4.26 X 1018 51 2.99 X 1019 1.78 X 1019 6.33 X 1018 Notes: (a) l/4T and 3/4T== F(slllfaceJ *e(-0i4"xl, where xis the depth into the vessel wall (i.e. 8.625*0.25 or 0.75) TABLE2 Measured Integrated Neutron Exposure of the Catawba Unit 2 Surveillance Capsules Tested To Date Plant Capsule Fluence Catawba Unit 2 z 3.23 x 1018 n/cm2, (E > 1.0 MeV) Catawba Unit 2 X 1.23 x I 019 n/cm2, (E > 1.0 MeV) Catawba Unit 2 V 2.38 x 1019 n/cm2, (E > 1.0 MeV) Catawba Unit 2 y 2.49 x 1019 n/cm2, (E > 1.0 MeV) 12 Margin is calculated as, M = 2..J cr i + cr !. The standard deviation for the initial RT NDT margin term, is a; 0°F when the initial RT NDT is a measured value, and l 7°F when a generic value is available. The standard deviation for the L\\RT NDT margin term, Oti, is l 7°F for plates or forgings, and 8.5°F for plates or forgings when surveillance data is used. For welds, Oti is equal to 28°F when surveillance capsule data is not used, and is 14°F (half the value) when credible surveillance capsule data is used. CT.ti need not exceed 0.5 times the mean value of ~T NDT. Contained in Table 3 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials[ll_ These measured shift values were obtained using CVGRAPH, Version 4.1 191, which is a hyperbolic tangent curve-fitting program. WCAP-15285

TABLE3 Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data Material Capsule Measured 30 ft-lb Transition Temperature Shift(b) Intermediate Shell Plate z 24.68 B8605-1 X 48.07 (Longitudinal) V 44.76 Inter. Shell Plate z 29.55 B8605-1 X 52.17 (Transverse) V 61.29 Surveillance Program z -30.25(a) Weld Metal X 34.94 V 58.80 Notes: (a) This value will be assumed to be 0°F in this evaluation for conservatism (i.e. higher CF value). (b) Values documented in WCAP-15243111. Table 4 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTNDr of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in 13 Table 4 were used to generate the calculated chemistry factor (CF) values based on Tables 1 and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 6. Table 5 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Positio'ii 2.1, which are also summarized in Table 6.

TABLE4 Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description Cu(%) Ni(%) Initial RT NDT Closure Head Flange B8601-1 0.70 10 Vessel Flange B8602-1 0.71 10 Intennediate Shell Plate B8605-1 0.08 0.62 15 lntennediate Shell Plate B8605-2 0.08 0.61 33 lntennediate Shell Plate B8616-1 0.05 0.60 12 Lower Shell Plate B8806-1 0.06 0.56 6 Lower Shell Plate B8806-2 0.06 0.59 -10 Lower Shell Plate B8806-3 0.06 0.59 8 Intermediate Shell Plate Longitudinal 0.04 0.14 -80 Weld Seams 101-124A, B, C (Heat# 83648) Lower Shell Plate Longitudinal 0.04 0.14 -80 Weld Seams 101-142A, B, C (Heat# 83648) Intermediate to Lower Shell Plate Circumferential 0.04 0.14 -80 Weld Seam 101-171 (Heat# 83648) Surveillance Weld Catawba Unit 2 (Heat # 83648) (b) 0,04cc) o.1d*) Notes: (a) Based on measured data. (b) The surveillance weld was made with the same weld wire and flux as the intermediate and lower shell longitudinal welds and the intennediate to lower shell girth weld (Weld wire heat number 83648 and Linde 0091 flux, lot number 3536). (c) Copper is the average of the two surveillance data points, 0.036 and 0.051. Nickel is the average of the two surveillance data points, 0.14 and 0.18 WCAP-15285 14 (a)

15 TABLE 5 Calculation of Chemistry Factors using Catawba Unit 2 Surveillance Capsule Data Material Capsule Capsule r*) FF(b) ~TNDr (c) FF*~TNDr FF2 Inter. Shell Plate z 0.323 0.689 24.68 17.00 0.475 B8605-l X 1.23 1.058 48.07 50.86 I.I l 9 (Longitudinal) V 2.38 1.234 44.76 55.23 1.523 Inter. Shell Plate z 0.323 0.689 29.55 20.36 0.475 B8605-1 X 1.29 1.058 52.17 55.20 1.119 (Transverse) V 2.38 1.234 61.29 75.63 1.523 SUM: 274.28 6.234 CF :e !:(FF

• RTNur) + !:( FF2) = (274.28) + (6.234) :e 44.0°F Catawba Unit 2 Surv.

z 0.323 0.689 o.oo<d. c) 0.00 0.475 Weld Material X 1.29 1.058 33.19 (34.94id) 35.12 1.119 V 2.38 1.234 55.86 (58.8f dl 68.93 1.523 SUM: 104.05 3.117 CF-I(FF*RTNDT) + L(FF2)-(104.05)+ (3.ll7)=33.4°F Notes: (a) f= Calculated fluence from capsule V dosimetry analysis resu11s<9l, (x 1019 n/cm2, E > 1.0 MeV). (b) FF= fluence factor= f<0.2S-o.t*JogO. (c) ART NOT values are the measured 30 ft-lb shift values for Catawba 2 taken from App.Bore of Reference 9. (d) The surveillance weld metal ART NDT values have been adjusted by a ratio factor of0.95 (See calc. on previous page). (e) Actual value of ARTNur is -30.25. This physically should not occur, therefore for conservatism (i.e. higher chemistry factor) a value of zero will be used. WCAP-15285

16 TABLE 6 Summary of the Catawba Unit 2 Reactor Vessel Beltline Material Chemistry Factors Material Reg. Guide I.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1.1 CF's Position 2.1 Cf's Intermediate Shell Plate B8605-l 51.0°F 44.0°F Intermediate Shell Plate B8605-2 51.0°F Intermediate Shell Plate B86 I 6-I 31.0°F Lower Shell Plate B8806-1 37.0°F Lower Shell Plate B8806-2 37.0°F Lower Shell Plate B8806-3 37.0°F Intermediate Shell Plate Longitudinal 37.3°F 33.4°F Weld Seams 101-124A, B, C (Heat# 83648) Lower Shell Plate Longitudinal 37.3°F 33.4°F Weld Seams I0l-142A, B, C (Heat # 83648) Intermediate to Lower Shell Plate 37.3°F 33.4°F Circumferential Weld Seam 101-171 (Heat # 83648) Surveillance Weld Catawba Unit 2 39.2°F (Heat # 83648) No surveillance material, thus Position 2.1 does not apply. WCAP-15285

17 Contained in Table 7 is a summary of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Catawba Unit 2 reactor vessel beltline materials. TABLE7 Summary of the Calculated Fluence Factors Used for the Generation of the 22, 34 and 51 EFPY Heatup and Cooldown Curves EFPY l/4TFF 3/4TFF 22 0.931 0.650 34 1.051 0.763 51 1.158 0.872 Note: FF = fluence<0-28

• o. JLog(fluena:)), where the fluence is taken from Table 1.

Based on the surveillance program credibility evaluation presented in Appendix D to WCAP-15243, the Catawba Unit 2 surveillance program data is credible. In addition, the surveillance program weld metal is representative of all of the beltline region girth weld seam. Hence, the adjusted reference temperature (ART) must be calculated for 22, 34 and 51 EFPY for each beltline material at the l/4T and 3/4T locations. In addition, ART values must be calculated per Regulatory Guide 1.99, Revision 2, Position 1.1 and 2.1 Contained in Tables 8 through 13 are the calculations of the 22, 34 and 51 EFPY ART values used for generation of the heatup and cooldown curves. WCAP-15285

TABLE 8 Calculation of the ART Values for the l/4T Location@ 22 EFPY Material Reg. Guide 1.99 Rev. 2 Method Intermediate Shell Plate Position 1.1 B8605-l Position 2.1 lntennediate Shell Plate Position 1.1 B8605-2 lntennediate Shell Plate Position 1.1 B8616-l Lower Shell Plate B8806-l Position I. I Lower Shell Plate B8806-2 Position 1.1 Lower Shell Plate B8806-3 Position 1.1 Inter. Shell Longitudinal Position 1.1 ~-------- Weld Seams 101-124A.,B,C Position 2.1 Lower Shell Longitudinal Position 1.1 ~-------- Weld Seams 101-124A,B,C Position 2.1 Circumferential Weld Seam Position 1.1 1---------- 101-171 Position 2.1 NOTES: (a) (b) Initial RT Nur values are measured values. ~TNOr: CF* FF CF (OF) 51.0 ~------ 44.0 51.0 31.0 37.0 37.0 37.0 37.3 33.4 37.3 33.4 37.3 ~------ 33.4 FF IRTND/*l aRTNDT(b) (OF) (OF) 0.931 15 47.5

---

------- i--------

0.931 15 41.0 0.931 33 47.5 0.931 12 28.9 0.931 6 34.4 0.931 -10 34.4 0.931 8 34.4 0.931 -80 34.7 0.931 -80 31.1 0.931 -80 34.7

---

~-------------

0.931 -80 31.1 0.931 -80 34.7

---

~-------------

0.931 -80 31.l M (OF) 34 17 34 28.9 34 34 34 34.7 28.0 34.7 28.0 34.7 28.0 (c) ART: I+ ~TNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".) WCAP-15285 18 ART0l (OF) 97 ~----- 73 115 70 74 58 76 -11 -21 -11 i------- -21 -11 -21

TABLE 9 Calculation of the ART Values for the 3/4 T Location @ 22 EFPY Material Reg. Guide 1.99 Rev. 2 Method Intermediate Shell Plate Position 1.1 ~-------- B8605-l Position 2.1 Intennediate Shell Plate Position 1.1 B8605-2 Intermediate Shell Plate Position 1.1 B8616-1 Lower Shell Plate B8806-1 Position 1.1 Lower Shell Plate B8806-2 Position 1.1 Lower Shell Plate B8806-3 Position 1.1 Inter. Shell Longitudinal Position 1.1 ~-------- Weld Seams 101-124A,B,C Position 2.1 Lower Shell Longitudinal Position 1.1 ~-------- Weld Seams IOI-124A,B,C Position 2.1 Circumferential Weld Seam Position 1.1 ~-------- 101-171 Position 2. 1 NOIBS: (a) (b) Initial RTNDT values are measured values. ~TNDT =CF. FF CF (OF) 51.0 44.0 51.0 31.0 37.0 37.0 37.0 37.3 33.4 37.3 33.4 37.3 33.4 FF IRTNDT(a) dRTNDT(b) (OF) (OF) 0.650 15 33.2

---

-------1-------

0.650 15 28.6 0.650 33 33.2 0.650 12 20.2 0.650 6 24.l 0.650 -10 24.1 0.650 8 24.l 0.650 -80 24.2 ~----- -------------- 0.650 -80 21.7 0.650 -80 24.2 ~----- -------------- 0.650 -80 21.7 0.650 -80 24.2 ~----- -------------- 0.650 -80 21.7 M (OF) 33.2 ~---- 17.0 33.2 20.2 24.1 24.1 24.1 24.2 1----- 21.7 24.2 1----- 21.7 24.2 ~---- 21.7 (c) ART= I+ ~T NOT + M (This value was rounded per ASlM E29, using the "Rounding Method".) WCAP-15285 19 AR-r<*> (OF) 81 61 99 52 54 38 56 -32 -37 -32 -37 -32 -37

TABLE 10 Calculation of the ART Values for the l/4T Location@ 34 EFPY Material Reg. Guide 1.99 Rev. 2 Method Intermediate Shell Plate Position 1.1 B8605*1 Position 2.1 Intermediate Shell Plate Position 1.1 B8605*2 Intennediate Shell Plate Position 1.1 B8616*1 Lower Shell Plate B8806* l Position 1.1 Lower Shell Plate B8806-2 Position 1.1 Lower Shell Plate B8806-3 Position 1.1 Inter. Shell Longitudinal Position 1.1 ~-------- Weld Seams 101-124A.B,C Position 2.1 Lower Shell Longitudinal Position 1.1 1---------- Weld Seains l01*124A.B,C Position 2.1 Circumferential Weld Seam Position 1.1 ~-------- 101-171 Position 2.1 NOlES: (a) (b) Initial RT NOT values are measured values. LiRT NOT = CF

• FF CF (Of) 51.0

~----- 44.0 51.0 31.0 37.0 37.0 37.0 37.3 33.4 37.3 33.4 37,3 33.4 FF IRTNDT(*) ARTNDT(b) (OF) (OF) 1.051 15 53,6 1------- i.-------~------ 1.051 15 46.2 1.05] 33 53.6 I.OSI 12 32.6 1.051 6 38.9 1.051 .IQ 38.9 1.051 8 38.9 1.051 -80 39.2 1.051 -80 35.1 1.051 -80 39.2

---

~------1---------

1.051 -80 35.1 1.051 -80 39.2 1.051 -80 35.1 M (OF) 34 17 34 32.6 34 34 34 39.2 28.0 39.2 1------ 28.0 39.2 28.0 (c) ART= I+ LiRTNor + M (This value was rounded per ASlM E29, using the Rounding Method".) WCAP-15285 20 ARrc> (OF) 103 1------- 78 121 77 69 63 81 -2 1------- -17 -2 1------- -17 -2 -17

TABLE 11 Calculation of the ART Values for the 3/4 T Location @ 34 EFPY Material Reg. Guide 1.99 Rev. 2 Method Intermediate Shell Plate Position 1.1 B8605-l Position 2.1 Intermediate Shell Plate Position 1.1 B8605-2 Intermediate Shell Plate Position 1.1 B8616-l Lower Shell Plate B8806-1 Position 1.1 Lower Shell Plate B8806-2 Position 1.1 Lower Shell Plate B8806-3 Position 1.1 Inter. Shell Longitudinal Position 1.1 Weld Seams 101-124A,B,C Position 2.1 Lower Shell Longitudinal Position 1.1 Weld Seams 101-124A,B,C Position 2.1 Circumferential Weld Seam Position 1.1 101-171 Position 2.1 NOTES: (a) (b) Initial RT Nur values are measured values. ~TNirr =CF* FF CF (OF) 51.0 44.0 51.0 31.0 37.0 37.0 37.0 37.3 33.4 37.3 33.4 37.3 33.4 FF IRTNDT(a) ~TNDT(b) (OF) (OF) 0.763 15 38.9 1------ ,..... ______ 1------- 0.763 15 33.6 0.763 33 38.9 0.763 12 23.7 0.763 6 28.2 0.763 -10 28.2 0.763 8 28.2 0.763 -80 28.5 0.763 -80 25.5 0.763 -80 28.5

---

-------~------

0.763 -80 25.5 0.763 -80 28.5 0.763 -80 25.5 M (OF) 34 17 34 23.7 28.2 28.2 28.2 28.5 25.5 28.5 25.5 28.5 25.5 (c) ART"'" I+ ~TNur + M (This value was rounded per ASTM E29, using the "Rounding Method".) WCAP-15285 21 ARr0> {°F) 88 66 106 59 62 46 64 -23 1------ -29 -23 -29 -23 -29

TABLE 12 Calculation of the ART Values for the l/4T Location@ 51 EFPY Material Reg. Guide 1.99 Rev. 2 Method Intermediate Shell Plate Position 1.1 B8605-l Position 2.1 Intermediate Shell Plate Position 1.1 B8605-2 Intermediate Shell Plate Position 1.1 B8616-l Lower Shell Plate B8806-1 Position 1.1 Lower Shell Plate B8806-2 Position 1.1 Lower Shell Plate B8806-3 Position 1.1 Inter. Shell Longitudinal Position 1.1 ~-------- Weld Seams 101-124A,B,C Position 2.1 Lower Shell Longitudinal Position 1.1 ~-------- Weld Seams 101-124A,B,C Position 2.1 Circumferential Weld Seam Position 1.1 101-171 Position 2.1 NOTES: (a) (b) Initial RT NOT values are measured values. AflT NOT = CF

• FF CF (OF) 51.0 44.0 51.0 31.0 37.0 37.0 37.0 37.3 33.4 37.3 33.4 37.3 33.4 FF IRTNDT(*)

~TNDT(b) (OF) (OF) 1.158 15 59.1 1.158 15 51.0 l.158 33 59.1 1.158 12 35.9 1.158 6 42.8 1.158 -10 42.8 1.158 8 42.8 1.158 -80 43.2 1.158 -80 38.7 1.158 -80 43.2

---

~-------------

1.158 -80 38.7 1.158 -80 43.2 1------- ~------------- 1.158 -80 38.7 M (Of) 34 17 34 34 34 34 34 43.2 28.0 43.2 i------ 28.0 43.2 i------ 28.0 (c) ART= I+ AflT NDT + M (This value was rowided per ASTM E29, using the "Rounding Method".) WCAP-15285 22 ARJ-<*> (OF) 108 ~----- 83 126 82 83 67 85 6 -13 6 -13 6 -13

TABLE 13 Calculation of the ART Values for the 3/4T Location @ 51 EFPY Material Reg. Guide 1.99 Rev. 2 Method Intermediate Shell Plate Position 1.1 B8605-1 Position 2.1 lntennediate Shell Plate Position 1.1 B8605-2 Intermediate Shell Plate Position 1. 1 B8616-1 Lower Shell Plate B8806-1 Position 1.1 Lower Shell Plate B8806-2 Position 1.1 Lower Shell Plate B8806-3 Position 1.1 Inter. Shell Longitudinal Position 1.1 Weld Seams 101-124A,B,C Position 2.1 Lower Shell Longitudinal Position l.l Weld Seams 101-124A,B,C Position 2.1 Circumferential Weld Seam Position 1.1 101-171 Position 2.1 NOTES: (a) (b) Initial RT NDT values are measured values. ~TNur =CF* FF CF (Of) 51.0 ~------ 44.0 51.0 31.0 37.0 37.0 37.0 37.3 1------- 33.4 37.3 ~------ 33.4 37.3 ~------ 33.4 FF IRTND/al ARTNDr (b) (OF) (Of) 0.872 15 44.5 0.872 15 38.4 0.872 33 44.5 0.872 12 27.0 0.872 6 32.3 0.872 -10 32.3 0.872 8 32.3 0.872 -80 32.5 0.872 -80 29.1 0.872 -80 32.S 1------ 1--------~------ 0.872 -80 29.1 0.872 -80 32.5 ,__ _____ ~------,_ ______ 0.872 -80 29.1 M (OF) 34 17 34 27.0 32.3 32.3 32.3 32.S 28.0 32.S 28.0 32.5 1----- 28.0 (c) ART= I+ AfffNDT + M (This value was rowtded per ASTM E29, using the "Rowtding Method".) WCAP-15285 23 ARrc> (Of) 94 70 112 66 71 55 73 -15 ~----- -22 -15 i------- -22 -15 -22

24 The intennediate shell plate B8605-2 is the limiting beltline material for all heatup and cooldown curves to be generated. Since the limiting material is not a circumferential weld, then the curves need not consider the circ. flaw methodology. Contained in Table 10 is a summary of the limiting ARTs to be used in the generation of the Catawba Unit 2 reactor vessel heatup and cooldown curves. WCAP-15285 TABLE14 Summary of the Limiting ART Values Used in the Generation of the Catawba Unit 2 Heatup/Cooldown Curves EFPY l/4T Limiting ART 3/4T Limiting ART 22 115 99 34 121 106 51 126 112

5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES 25 Pressure-temperature limit curves for nonnal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods[2l discussed in Sections 3.0 and 4.0 of this report. The pressure difference between the wide-range pressure transmitter and the limiting beltline region has not been accounted for in the pressure-temperature limit curves generated for nonnal operation. Figures 4 through 6, 8 through 10 and 12 through 14 present the heatup curve without margins for possible instrumentation errors using heatup rates of 60, 80 and l00°F/hr applicable for the first 22, 34 and 51 EFPY, respectively. Figures 7, 11 and 15 present the cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100°F /hr applicable for 22, 34 and 51 EFPY, respectively. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 4 through 15. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 4 through 6, 8 through 10 and 12 through 14. 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 Code Case N-640(3! as follows:

where, K1m is the stress intensity factor covered by membrane (pressure) stress, K1c = 33.2 + 20.734 e10*02 <T-RTNurll, T is the minimum permissible metal temperature, and RT NDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 5. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the in service hydrostatic leak tests for the Catawba Unit 2 reactor vessel at 22, 34 and 51 EFPY is l 75°F, 181 °F and l 86°F, respectively. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. WCAP-15285

Figures 4 through 15 define all of the above limits for ensuring prevention of nonductile failure for the Catawba Unit 2 reactor vessel. 26 The data points used for the heatup and cooldo'Ml pressure-temperature limit curves shown in Figures 4 through 15 are presented in Tables 15 through 20. When K1c and a relaxed reactor vessel flange temperature requirement is used in the calculation ofheatup and cooldown limit curves, there is a significant increase in allowable pressure. This increase in allowable pressure associated with the K1c methodology and relaxed reactor vessel flange requirement has created a "knee in the heatup curves which show the transition of the heatup curve from steady-state limiting to heatup rate "X°F /hr.". This can also be seen in Tables 15 through 20. WCAP-15285

27 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 22 EFPY: l/4T, 115°F 3/4T, 99°F b.() 00 Q., - 2500 2250 2000 \\ ,.,o,8887ll1 I I; i I i I I I I i I ! t i I ' LBAII: TEST I I ' I : I i I I 1 : i i I ' i i j \\ ~ j \\ ~ t : \\ I 0/: 7 I I I I I I /I I I I I

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28 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 22 EFPY: l/4T, I 15°F 3/4T, 99°F 2500 / : i R

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POB THI 1-+-+---41:-+: -+14-i--tl-+1~1-+_.__-+-.;..-.i......... --+-1--ts1a,1c1 PBBIOD UP to 22.0 BPPY1-1-.........+--"---~--'-'-- 0 i ' : i I I I I I I I I I 50 100 150 200 250 300 350 400 450 500 Moderator Temperature ( D e g. F ) Catawba Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 80°F/hr) Applicable for the First 22 EFPY (Without Margins for Instrumentation Errors)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 22 EFPY: l/4T, l l5°F 314T, 99°F 2500 I ' I/ ' I I i I I I R 11117un**: j I I i I i ' j i : I tW 2250 r:n I ' I I ' ' i I I i j I 'I ' I I I ~" I I I , I I I i I I rl LEAK TEST LIIIIT., f ' I i i f I I ' I I J* I I I I i I i I I I I ~ 2000 I I I I i I i I i I i j I i i I i I I i ' I I I f I I I I I i I I I I i f ' i,, I I i I I I I I I I I I Q,) 1750 I i I --- UNACCEPTABLE I I -- OPERAT I ON i/ f I I /1 I I I I r:n 1500 r:n Q.) 1250 i I i I I y i I I I/ I I I i i I iBEATUP RATE /I : I I I I I UP TO 1 00 P/Br. i i I ' . J I I I I I I I/ I I I I I I I I I I I I ACCEPTABLE I I i I I ' /I I / I I OPERAT I ON 0.. I I I V, V I I I I i I , A 1000 I I I i I v, I I i ' I i I ! I I ( ' I ""O i i i I i I '

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MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 22 EFPY: l/4T, l 15°F 3/4T, 99°F 2 500 I ~,,0,11111721 1 i I I I I I I I I I i I I ! i i i i I 7 ' I I I I 2250 i I I I I i i I tl..(l i I i ! , l r I ! i 1, ! I I

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31 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 34 EFPY: l/4T, 121 °F 3/4T, 106°F 2500 11110111lll10 I I b..O 2250 I I I I I I I LEAK TEST LIIIITI/ 00 Q., 2000 I I I I I I I J Q.) 1 7 5 0 I UNACCEPTABLE IJ I J..-4 OPERATION c<II I J 00 1500 I I I I I I/ / ' I I I I I I I I I II I 00 BE.lTUP RATE I/ ' ' I Q.) 1 2 5 0 J..-4 UP TO 80 F/Br. I/ ' J ' I / I ' I I I I I 0... I I.. I I/ 1, 1 0 0 0 /I I", I / I ""O i / I I I I I i I ACCEPTABLE OPERATION I I I I I I I I I CRIT. LIMIT I I Q.) I ,/ I ' I r-...: POR 60 P /Br. I IA I I I ~ 750 -~ I I I I I I ' I I I ro I

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32 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 34 EFPY: l/4T, 121 °F 3/4T, 106°F 2500 b.O 2250 00 Q.. 2000 Cl) 1 7 5 0 --- J.-c 00 1500 00 Cl) 1 2 5 0 0... 1 0 0 0 "'C Cl) 750 cd - 500 <:.) - 250 cd u 0 0 Figure 9 WCAP-15285 11734111144 I J I i I I I I I LEAK TEST LIMITV I I I I I UNACCEPTABLE OPERATION I I I I i I I I I I I BEATUP RATE I_ UP TO 80 F/Br. I I I I I I/ I I I I I I . I

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MATERJAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE 88605-2 LIMITING ART VALUES AT 34 EFPY: l/4T, 121 °F 3/4T, 106°F 33 2500 I j I I I I I,-,_:_r-t--t -

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MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 34 EFPY: l/4T, 121°F 3/4T, 106°F 2500 I i ! i i I

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36 MATERIAL PROPERlY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B8605-2 LIMITING ART VALUES AT 51 EFPY: l/4T, 126°F 3/4T, l 12°F 2500 bLl 2 2 5 0 00 i::i... 2000 I 1 ;,, I l i ~- 18??8810811--+-----l---+--i l--i---+--t----+~ l ---+-..j..-;HJ/~:-+....l'-+J'+-H-----i-+' -'4+-l--J,--+~l-+~l-++1-H 1 ,---',;_Lr 1-L.,-....... 1 1-.. 1--.,....,,~-1--1-+1--i-+1.-+-+-L, ---J--.j-+-+--+-1 --ll-+-1, +-H:L1H-: -+--+-+1-+-' -+-+--+-+-+-H ---'1H--i,,+ !, -+-+-, ~+t-l I 1 1 1 1 i ) I i ~ I I I LEAK TEST LIMIT~v:J_LJ' '~'_j_+-I-....L-l-l-J -l----4-l-l-l-'~--l--+-1----1---1--4--1-+-+-~'+-+-+--l--+-H-~'-+-H /, I I I I I Cl.) s-. H-h.b.:l...!.-l-abddd:::a:!:::::l:::::l:::ab.Jb;+-Lj:ij:t:t1tt-11-i.... -J-_ -t.u.iljt_tj:.tj:.tj_:.tj_=.tt:,ttjtt=t=t-:t=tt:.t:.t:j 1750 -~= UNACCEPTABLEL-L.~-l-J-4---+-J--l--l-4-__._.-+--l-4-~f-+--t--+~f-+--t--+~H-+-+-+-H ~~ OPERATION I 00 00 Cl.) s-. 0... "'O Cl.) ~ i:tS - C) - 1 5 0 0 1250 1 0 0 0 750 500 250 0 Figure 13 I J I I

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39 TABLE 15 22 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) 60 Heatup 60 Critical Limit 80 Heatup 80 Critical Limit 100 Heatup l 00 Critical Limit T p T p T p T p T p T p 55 0 175 0 55 0 175 0 55 0 175 0 55 722 175 747 55 722 175 747 55 722 175 747 60 734 175 812 60 734 175 809 60 734 175 808 65 747 175 808 65 747 175 800 65 747 175 796 70 762 175 809 70 762 175 795 70 762 175 787 75 778 175 815 75 778 175 795 75 778 175 782 80 796 175 826 80 795 175 798 80 781 175 781 85 808 175 841 85 795 175 806 85 781 175 783 90 808 175 861 90 795 175 817 90 781 175 789 95 809 175 884 95 795 175 833 95 781 175 798 100 815 175 912 100 795 175 852 100 781 175 810 105 826 175 944 105 798 175 875 l05 781 175 826 110 841 175 981 llO 806 175 902 110 783 175 845 115 861 180 1023 115 817 180 933 115 789 180 868 120 884 185 1070 120 833 185 969 120 798 185 895 125 912 190 1122 125 852 190 1010 125 810 190 927 130 944 195 1181 130 875 195 1056 130 826 195 962 135 981 200 1247 135 902 200 1107 135 845 200 1003 140 1023 205 1319 140 933 205 1165 140 868 205 1048 145 1070 210 1400 145 969 210 1229 145 895 210 1099 150 1122 215 1490 150 1010 215 1300 150 927 215 1156 155 1181 220 1589 155 1056 220 1380 155 962 220 1220 160 1247 225 1698 160 1107 225 1468 160 1003 225 1291 165 1319 230 1819 165 1165 230 1565 165 1048 230 1369 170 1400 235 1953 170 1229 235 1673 170 1099 235 1456 175 1490 240 2101 175 1300 240 1792 175 1156 240 1553 180 1589 245 2264 180 1380 245 1924 180 1220 245 1660 185 1698 250 2444 185 1468 250 2070 185 1291 250 1778 190 1819 190 1565 255 2230 190 1369 255 1909 195 1953 195 1673 260 2408 195 1456 260 2053 200 2101 200 1792 200 1553 265 2212 205 2264 205 1924 205 1660 270 2388 210 2444 210 2070 210 1778 215 2230 215 1909 220 2408 220 2053 225 2212 230 2388 WCAP-15285

Cooldown Curves Steady State T p 55 0 55 722 60 734 65 747 70 762 75 778 80 796 85 816 90 838 95 862 100 889 105 918 llO 951 115 987 120 1027 125 l071 )30 1120 135 1173 140 1233 145 1299 150 1371 155 1452 160 1541 165 1639 170 1747 175 1867 180 2000 185 2146 190 2308 WCAP-15285 TABLE 16 22 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncenainties for Instrumentation Errors) 20F 40F 60F T p T p T p 55 0 55 0 55 0 55 681 55 640 55 600 60 694 60 654 60 615 65 708 65 670 65 632 70 724 70 687 70 651 75 742 75 707 75 672 80 762 80 728 80 696 85 783 85 752 85 722 90 807 90 778 90 751 95 834 95 807 95 783 100 863 100 839 100 818 105 895 105 875 105 858 110 931 110 915 110 901 115 971 115 958 115 950 120 1015 120 1007 120 1003 125 1063 125 1060 125 1063 )30 1117 40 lOOF T p 55 0 55 519 60 538 65 558 70 581 75 607 80 636 85 668 90 703 95 743 100 786 105 835 110 889 115 948 120 1015

41 TABLE 17 34 EFPY Heatup Curve Data Points Using l 996 App. G (without Uncertainties for Instrumentation Errors) 60 Heatup 60 Critical Limit 80 Heatup 80 Critical Limit 100 Heatup 100 Critical Limit T p T p T p T p T p T p 55 0 181 0 55 0 181 0 55 0 181 0 55 709 181 731 55 709 181 731 55 709 181 731 60 719 181 782 60 719 181 779 60 719 181 779 65 731 181 777 65 731 181 769 65 731 181 766 70 744 181 776 70 744 181 763 70 744 181 756 75 759 181 780 75 759 181 761 75 747 181 750 80 775 181 789 80 761 181 763 80 747 181 747 85 776 181 801 85 761 181 769 85 747 181 748 90 776 181 817 90 761 181 778 90 747 181 751 95 776 181 837 95 761 181 790 95 747 181 758 100 780 181 861 100 761 181 806 100 747 181 768 105 789 181 888 105 763 181 825 105 747 181 780 110 801 181 920 llO 769 181 848 110 748 181 796 115 817 181 956 115 778 181 875 115 751 181 816 120 837 185 997 120 790 185 905 120 758 185 838 125 861 190 1042 125 806 190 940 125 768 190 865 130 888 195 1093 130 825 195 980 130 780 195 895 135 920 200 1150 135 848 200 1024 135 796 200 930 140 956 205 1213 140 875 205 1074 140 816 205 969 145 997 210 1283 145 905 210 1129 145 838 210 1013 150 1042 215 1361 150 940 215 ll91 150 865 215 1062 155 1093 220 1447 155 980 220 1260 155 895 220 1117 160 1150 225 1542 160 1024 225 1336 160 930 225 1178 165 1213 230 1647 165 1074 230 1421 165 969 230 1246 170 1283 235 1763 170 1129 235 1514 170 1013 235 1322 175 1361 240 1891 175 1191 240 1618 175 1062 240 1405 180 1447 245 2033 180 1260 245 1732 180 1117 245 1498 185 1542 250 2190 185 1336 250 1859 185 1178 250 1601 190 1647 255 2363 190 1421 255 1999 190 1246 255 1714 195 1763 195 1514 260 2153 195 1322 260 1839 200 1891 200 1618 265 2323 200 1405 265 1977 205 2033 205 1732 205 1498 270 2130 210 2190 210 1859 210 1601 275 2298 215 2363 215 1999 215 1714 280 2484 220 2153 220 1839 225 2323 225 1977 230 2130 235 2298 240 2484 WCAP-15285

Cooldown Curves Steady State T p 55 0 55 709 60 719 65 731 70 744 75 759 80 775 85 792 90 812 95 833 100 857 105 883 110 912 115 944 120 979 125 1018 130 1062 135 1109 140 1162 145 1221 150 1285 155 1356 160 1435 165 1522 170 1618 175 1725 180 1842 185 1972 190 2116 195 2274 200 2449 WCAP-15285 TABLE 18 34 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) 20F 40F 60F T p T p T p 55 0 55 0 55 0 55 667 55 624 55 596 60 678 60 637 60 596 65 691 65 651 65 611 70 705 70 666 70 628 75 721 75 683 75 647 80 738 80 702 80 667 85 757 85 723 85 690 90 778 90 747 90 716 95 802 95 772 95 744 100 828 100 801 l00 776 105 857 105 832 105 811 llO 888 110 867 110 849 115 924 115 906 115 892 120 963 120 949 120 940 125 1006 125 997 125 992 130 1053 130 1049 130 1050 135 1106 135 1107 42 lOOF T p 55 0 55 514 60 514 65 533 70 553 75 576 80 601 85 629 90 660 95 695 100 734 105 777 110 824 115 877 120 936 125 1001

43 TABLE19 51 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) 0 Heatup 60 Critical Limit 80 Heatup 80 Critical Limit 100 Heatup 100 Critical Limit T p T p T p T p T p T p 55 0 186 0 55 0 186 0 55 0 186 0 55 699 186 719 55 699 186 719 55 699 186 719 60 709 186 760 60 709 186 757 60 709 186 757 65 719 186 753 65 719 186 746 65 719 186 743 70 731 186 752 70 731 186 739 70 721 186 733 75 744 186 754 75 736 186 736 75 721 186 726 80 752 186 761 80 736 186 737 80 721 186 722 85 752 186 771 85 736 186 741 85 721 186 721 90 752 186 784 90 736 186 748 90 721 186 723 95 752 186 802 95 736 186 758 95 721 186 728 100 754 186 822 100 736 186 771 100 721 186 736 105 761 186 846 105 737 186 788 105 721 186 746 110 771 186 874 110 741 186 807 110 721 186 760 115 784 186 906 115 748 186 831 115 723 186 776 120 802 186 941 120 758 186 857 120 728 186 796 125 822 190 982 125 771 190 888 125 736 190 819 130 846 195 1027 130 788 195 923 130 746 _195 845 135 874 200 1077 135 807 200 962 135 760 200 875 140 906 205 1133 140 831 205 1006 140 776 205 909 145 941 210 1195 145 857 210 1055 145 796 210 948 150 982 215 1263 150 888 215 1109 150 819 215 991 155 1027 220 1340 155 923 220 1170 155 845 220 1040 160 1077 225 1424 160 962 225 1237 160 875 225 1094 165 1133 230 1517 165 l006 230 1312 165 909 230 1154 170 1195 235 1620 170 l055 235 1395 170 948 235 1221 175 1263 240 1734 175 1109 240 1487 175 991 240 1295 180 1340 245 1860 180 1170 245 1588 180 1040 245 1376 185 1424 250 1998 185 1237 250 1700 185 1094 250 1467 190 1517 255 2152 190 1312 255 1824 190 I 154 255 1567 195 1620 260 2321 195 1395 260 1961 195 1221 260 1678 200 1734 200 1487 265 2112 200 1295 265 1801 205 1860 205 1588 270 2278 205 1376 270 1936 210 1998 210 1700 275 2462 210 1467 275 2085 215 2152 215 1824 215 1567 280 2250 220 2321 220 1961 220 1678 285 2431 225 2112 225 1801 230 2278 230 1936 235 2462 235 2085 240 2250 245 2431 1 1lf,,~1ftfW8afdltt 2000 2485 WCAP-15285

Cooldown Curves Steady State T p 55 0 55 699 60 709 65 719 70 731 75 744 80 759 85 775 90 792 95 812 100 833 105 857 llO 883 115 912 120 944 125 979 130 1018 135 1062 140 1109 145 ll62 150 1221 155 1285 160 1356 165 1435 170 1522 175 1618 180 1725 185 1842 190 1972 195 2116 200 2274 205 2449 WCAP-15285 TABLE 20 51 EFPY Cooldo1/4Tl Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) 20F 40F 60F T p T p T p 55 0 55 0 55 0 55 656 55 613 55 569 60 667 60 624 60 582 65 678 65 637 65 595 70 691 70 651 70 610 75 705 15 666 75 627 80 721 80 683 80 646 85 738 85 702 85 667 90 757 90 723 90 690 95 778 95 746 95 715 100 802 100 772 100 744 105 828 105 801 105 775 110 857 llO 832 110 810 115 888 ll5 867 115 849 120 924 120 906 120 892 125 962 125 949 125 939 130 1006 130 996 130 992 135 1053 135 1049 135 1050 140 ll06 140 1107 44 lOOF T p 55 0 55 482 60 497 65 513 70 532 75 552 80 575 85 600 90 628 95 660 100 694 105 733 110 776 ll5 824 120 877 125 935 130 1001

6 REFERENCES

l.

WCAP-15243, "Analysis of Capsule V and the Capsule Y Dosimeters from the Duke Energy Catawba Unit 2 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et al., Dated September 1999.

2.

WCAP-7924-A, "Basis for Heatup and Cooldown Limit Curves," W. S. Hazelton, et al., April 1975.

3.

ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division l", February 26, 1999.

4.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988. 45

5.

Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

6.

Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure.", Dated 1989 & December 1995.

7.

1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels."

8.

WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

9.

CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.0, developed by ATI Consulting, March 1995. WCAP-15285}}