Text

Pressure and Temperature Limits Report (PTLR) Revision 2
	ML25227A166
Person / Time
Site:	Hope Creek
Issue date:	08/15/2025
From:	Jennings J; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LR-N25-0073
	Download: ML25227A166 (1)
	v • d • e

Jason Jennings Director - Site Regulatory Compliance, PSEG Nuclear 2200 Alloway Creek Neck Road PO Box 236 Hancocks Bridge, New Jersey 08038-0221 (856) 339-1653 Jason.Jennings@PSEG.com TS 6.9.1.10 LR-N25-0073 August 15, 2025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Pressure and Temperature Limits Report (PTLR) Revision 2 PSEG Nuclear LLC (PSEG) is transmitting Revision 2 of the PTLR in accordance with Hope Creek Generating Station Technical Specification (TS) 6.9.1.10. Revision 2 of the PTLR was issued as a result of the transition to a 24 Month Fuel Cycle. Enclosure 1 provides Revision 2 of the Hope Creek PTLR No new regulatory commitments are established by this submittal. If there are any questions or if additional information is needed, please contact Mr. Brian Thomas at brian.thomas@pseg.com.

Respectfully, Jason Jennings Director - Site Regulatory Compliance

Hope Creek Pressure and Temperature Limits Report Revision 2 cc:

Administrator, Region I, NRC NRC Project Manager, Hope Creek NRC Senior Resident Inspector, Hope Creek Manager NJBNE

LR-N25-0073 Hope Creek Pressure and Temperature Limits Report Revision 2

Hope Creek Generating Station PTLR Revision 2 Page 1 of 53 PSEG Nuclear LLC Hope Creek Generating Station (HCGS)

Pressure and Temperature Limits Report (PTLR) for 44 and 56 Effective Full-Power Years (EFPY)

Revision 2 Prepared by: ~--+-g-~_-______

Date: 02/26/2025 Hainsworth

[Hope Creek Engineering Programs]

Reviewed by: )J ~~

Donnamarie Bush

[Manager, Engineering Programs]

[Sr. Director, Engineering]

Table of Contents Section Page 1.0 Purpose 4

2.0 Applicability 4

3.0 Methodology 5

4.0 Operating Limits 6

5.0 Discussion 7

6.0 References 15 Appendix A Hope Creek Reactor Vessel Materials Surveillance Program 53

List of Figures Figure 1: HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 44 EFPY [7]..............18 Figure 2: HCGS P-T Curve B (Normal Operation - Core Not Critical) for 44 EFPY [7]............19 Figure 3: HCGS P-T Curve C (Normal Operation - Core Critical) for 44 EFPY [7]...................20 Figure 4: HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 56 EFPY [7]..............21 Figure 5: HCGS P-T Curve B (Normal Operation - Core Not Critical) for 56 EFPY [7]............22 Figure 6: HCGS P-T Curve C (Normal Operation - Core Critical) for 56 EFPY [7]...................23 Figure 7: HCGS Feedwater Nozzle 3D Finite Element Model [18]..............................................24 Figure 8: HCGS Feedwater Nozzle Stress Extraction Path [18]...................................................25 Figure 9: HCGS LPCI Nozzle 3D Finite Element Model [19]......................................................26 Figure 10: HCGS LPCI Nozzle Stress Extraction Path [19].........................................................27 Figure 11: HCGS Instrument Nozzle Finite Element Model [20].................................................28 Figure 12: HCGS Instrument Nozzle Stress Extraction Path [20].................................................29 List of Tables Table 1: HCGS Pressure Test (Curve A) P-T Curves for 44 EFPY [7]........................................30 Table 2: HCGS Core Not Critical (Curve B) P-T Curves for 44 EFPY [7]..................................33 Table 3: HCGS Core Critical (Curve C) P-T Curves for 44 EFPY [7].........................................36 Table 4: HCGS Pressure Test (Curve A) P-T Curves for 56 EFPY [7]........................................39 Table 5: HCGS Core Not Critical (Curve B) P-T Curves for 56 EFPY [7]..................................42 Table 6: HCGS Core Critical (Curve C) P-T Curves for 56 EFPY [7].........................................45 Table 7: HCGS 1/4T ART Table for 44 EFPY [6]........................................................................48 Table 8: HCGS 1/4T ART Table for 56 EFPY [6]........................................................................50 Table 9: Nozzle Stress Intensity Factors [18] [19] [20].................................................................52

1.0 PURPOSE The purpose of the Hope Creek Generating Station (HCGS) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cooldown and Hydrostatic/Class 1 Leak Testing.

2. RCS Heat-up and Cooldown rates.

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of the current and previous revisions of Licensing Topical Reports SIR-05-044 contained within BWROG-TP-11-022-A, Revision 1 [1].

2.0 APPLICABILITY This report is applicable to the HCGS RPV for up to 44 and 56 Effective Full-Power Years (EFPY).

The following HCGS Technical Specifications (TS) are affected by the information contained in this report:

TS 3.4.6 RCS Pressure/Temperature (P/T) Limits TS 4.4.6 Surveillance Requirements 3.0 METHODOLOGY The limits in this report were derived as follows:

1. The methodology used is in accordance with Reference [1], Pressure - Temperature Limits Report Methodology for Boiling Water Reactors, August 2013, incorporating the NRC Safety Evaluation in Reference [2].

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [3] as documented in Reference [4].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99) [5], as documented in Reference [6].

4. The pressure and temperature limits, which were calculated in accordance with Reference

[1], are documented in Reference [7] which satisfied the requirements of References [8]

and [9].

5. This revision of the pressure and temperature limits report is to incorporate the following changes:

Revision 0: Initial issue of PTLR.

Revision 1: Updated fluence for 1.6% power uprate conditions.

Revision 2: to incorporate new irradiation fluence data [4, 6] that go out to 44 and 56 EFPY of the RPV to support transition to 24-month fuel cycles.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 [10], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the methodology of the PTLR require NRC approval. Update of a PTLR with a prior approved methodology only requires submission to the NRC (and not NRC approval).

4.0 OPERATING LIMITS The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 44 and 56 EFPY for HCGS, as documented in Reference [7]. The HCGS P-T curves for 44 EFPY are provided in Figure 1 through Figure 3, for 56 EFPY are in Figure 4 through Figure 6. A tabulation of the curves is included in Table 1 through Table 3 for 44 EFPY and Table 4 through Table 6 for 56 EFPY. The adjusted reference temperature (ART) tables for 44 and 56 EFPY for the HCGS vessel beltline materials are shown in Table 7 and Table 8 [6] respectively.

The resulting P-T curves are based on the geometry, design and materials information for the HCGS vessel. The following conditions apply to operation of the HCGS vessel:

Heat-up/Cooldown rate limit during Hydrostatic Class 1 Leak Testing (Figure 1 and Figure 4: Curve A): 25F/hour1 [1].

Normal Operating Heat-up and Cooldown rate limit (Figure 2 and Figure 5: Curve B -

core non-critical, and Figure 3 and Figure 6: Curve C - core critical): 100°F/hr2 [7].

RPV bottom head coolant temperature to RPV coolant temperature T limit during Recirculation Pump startup: 145°F [1].

1 Interpreted as the temperature change in any 1-hour period is less than or equal to 25°F.

2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100°F.

Recirculation loop coolant temperature to RPV coolant temperature T limit during Recirculation Pump startup: 50°F [1].

RPV head flange, RPV flange and adjacent shell temperature limit during vessel bolt-up 79F [7].

To address the NRC condition regarding lowest service temperature in Reference [2, Section 4.0], the minimum temperature is set to 79°F, which is equal to RTNDT,max + 60°F. This value is consistent with the minimum temperature limits and minimum bolt-up temperature in the current docketed P-T curves [11].

5.0 DISCUSSION The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [5] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the HCGS vessel plate, weld, and forging materials [6]; this evaluation included the results of two surveillance capsules. The Cu and Ni values were used with Table 1 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds. The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings.

However, for materials where surveillance data exists (i.e. HCGS plate heat no. 5K3238/1 and weld heat no. D53040), a fitted CF has been used in the calculation of ART for those heats, in accordance with Regulatory Paragraph 2.1 in RG 1.99 [5]. Use of surveillance data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) for HCGS was approved by the NRC in Reference [12].

Per Reference [6] and in accordance with Appendix A of Reference [1], the HCGS representative weld and plate surveillance materials data were reviewed from the BWRVIP ISP

[13]. Proprietary Information contained in BWRVIP-135 Revision 4, Section 2, Appendix A, and Appendix B [13] has been re-classified as non-proprietary per BWRVIP 2023-039 [14]. The representative heat of the plate material (5K3238/1) in the ISP is not the same as the target plate material (5K3025/1) for HCGS. However, the surveillance heat 5K3238/1 does exist in the HCGS vessel beltline, and there are two irradiated data sets for this plate. For plate heat 5K3238/1, since the scatter in the fitted results is less than 1-sigma (17°F) [13], the data is credible per Reference [5], and a reduced margin term ( = 17°F/2 = 8.5°F) is used for the plate material when calculating the ART. The CF from RG 1.99 table for plate heat 5K3238/1 bounds the fitted CF. Therefore, the higher RG 1.99 table CF for plate heat 5K3238/1 is used to calculate the ART. The representative heat of the weld material (D53040) in the ISP is the same as the limiting weld material in the vessel beltline region of HCGS. Per Reference [13], scatter in the surveillance data exceeds the credibility criteria for weld heat D53040, however, the fitted CF bounds the RG 1.99 CF. Therefore, the higher surveillance-based CF is used in the ART calculation for weld heat no. D53040, with a full margin term ( = 28°F). The RG 1.99 table CFs were used in the determination of the ART values for all HCGS beltline materials except for the surveillance weld heat D53040.

The limiting ART material ID fluence values of 1.33 x 1018 n/cm2 at 44 EFPY and 1.60 x 1018 n/cm2 at 56 EFPY used in the P-T curve evaluations were obtained from Reference [4]. The fluence values apply to the limiting beltline lower intermediate vertical weld (Heat No. D53040).

The fluence values for the lower intermediate vertical are based upon an attenuation of 0.693 from a generic attenuation method [6] for a postulated 1/4T flaw. Consequently, the 1/4T fluence for 44 and 56 EFPY for the limiting lower intermediate vertical weld are 9.22 x 1017 n/cm2 and 1.11 x 1018 n/cm2, respectively, for HCGS [6]. The limiting value for 1/4T ART is 110.5°F at 44 EFPY and 118.3°F at 56 EFPY [6].

The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. HCGS has two sets of nozzles in the RPV beltline: the instrument (N16) and low pressure coolant injection (LPCI, N17) nozzles are located in the intermediate shell beltline plates [15]. The feedwater (FW) nozzle is considered in the evaluation of the non-beltline (upper vessel) region P-T limits.

The instrument (N16) nozzle material is ferritic and is welded to the RPV using a partial penetration weld. The effect of the penetration on the adjacent shell is considered in the development of bounding beltline P-T limits as described in Reference [7]. The N16 nozzles have a limiting RPV ID fluence of 3.67 x 1017 n/cm2 at 44 EFPY and 4.47 x 1017 n/cm2 at 56 EFPY [4]. These fluence values apply to the adjacent intermediate shell plates in which the nozzles are located. Similar to the RPV beltline plates and welds described above, through-wall fluence for the N16 nozzles was attenuated using the generic attenuation methodology in Reference [6]. The resulting attenuation factor is 0.693 for a postulated 1/4T flaw in the N16 nozzle corner. Consequently, the 1/4T fluences for 44 and 56 EFPY for the limiting N16 nozzle location are 2.54 x 1017 n/cm2 and 3.10 x 1017 n/cm2 respectively. The limiting value for 1/4T ART of the N16 nozzles is 63.9°F at 44 EFPY and 69.5°F at 56 EFPY [6].

The LPCI (N17) nozzles have an RPV ID fluence of 5.46 x 1017 n/cm2 at 44 EFPY and 6.67 x 1017 n/cm2 at 56 EFPY, obtained from Reference [4]. Similar to the RPV beltline plates and welds described above, through-wall fluence for the LPCI nozzles was attenuated using the generic attenuation methodology in Reference [6]. The resulting attenuation factor is 0.693 for a postulated 1/4T flaw in the LPCI nozzle blend radius. Consequently, the 1/4T fluences for 44 and 56 EFPY for the limiting LPCI nozzle location are 3.79 x 1017 n/cm2 and 4.63 x 1017 n/cm2 respectively. The limiting value for 1/4T ART of N17 nozzles is 32.8°F at 44 EFPY and 39.0°F at 56 EFPY [6].

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is

usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heat-up and cooldown. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable fracture toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, and for a given pressure, the coolant saturation temperature is well above the P-T curve limiting temperature. Consequently, the material fracture toughness at a given pressure would exceed the allowable fracture toughness.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heat-up and cooldown temperature rate of 100°F/hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level A/B RPV thermal transients. For the hydrostatic pressure and leak test curves (Curve A), a coolant heat-up and cooldown temperature of 25°F/hr must be maintained.

The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heat-up /cooldown rate limits cannot be maintained.

The initial RTNDT, chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 1017 n/cm2 for E >

1 MeV) are shown in Table 7 and Table 8 for 44 and 56 EFPY respectively [6]. Use of initial RTNDT values in the determination of P-T curves for HCGS was approved by the NRC in Reference [16].

The only computer code used in the determination of the HCGS P-T curves was the ANSYS Mechanical, Release 18.1 [17] finite element computer program. ANSYS finite element analyses were used to develop the stress distributions through the FW nozzle (non-beltline), LPCI nozzle and instrument nozzle (beltline), and these stress distributions were used in the determination of the stress intensity factors for the FW nozzle [18], LPCI nozzle [19], instrument nozzle [20] and vessel shell. At the time that each of the analyses above was performed, the ANSYS program was controlled under the vendors 10 CFR 50 Appendix B [21] Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [22]

was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems.

The plant-specific HCGS FW nozzle analyses were performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analyses can be found in Reference [18]. The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle:

A three-dimensional (3D) finite element model of the FW nozzle was constructed for the determination of thermal stresses (Figure 7). Details of the model and material properties are provided in Reference [18]. Temperature-dependent material properties were based on the ASME Code,Section II, Part D, 2001 Edition through 2003 Addenda [23].

Heat transfer coefficients in Reference [18] are a function of FW temperature and flow rate. Potential leakage past the primary thermal sleeve is considered in the heat transfer calculations.

With respect to operating conditions, the bounding thermal transients during normal and upset operating conditions were analyzed [18]. The thermal stress distribution, corresponding to the limiting time presented in Reference [18], along a linear path (Figure 8) through the nozzle corner is used. The boundary integral equation/influence function (BIE/IF) methodology presented in Reference [1] was used to calculate the

thermal stress intensity factor, KIt, by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

With respect to pressure stress, a unit pressure of 1005 psig was applied to the internal surfaces of the 3D model in Reference [18]. The pressure stress distribution was taken along a linear path through the nozzle corner. The BIE/IF methodology presented in Reference [1] was used to calculate the applied pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp may be linearly scaled to determine the KIp for various RPV internal pressures.

The plant-specific HCGS LPCI nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analysis can be found in Reference [19]. The following summarizes the development of the thermal and pressure stress intensity factors for the instrument nozzle:

A one-quarter symmetric, three-dimensional finite element model of the instrument nozzle was constructed and is shown in Figure 9. Temperature-dependent material properties, taken from the ASME Code,Section II, Part D, 2001 Edition with 2003 Addenda edition [23], were used in the evaluation and are described in Reference [19].

With respect to operating conditions, the bounding thermal transient for the region corresponding to the instrument nozzles during normal and upset operating conditions was analyzed [19]. The thermal stress distribution, corresponding to the limiting time in Reference [19], along a linear path in Figure 10 through the nozzle corner is used. The BIE/IF methodology presented in Reference [1] was used to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

Boundary conditions and heat transfer coefficients used for the thermal analysis are described in Reference [19].

With respect to pressure stress, a unit pressure of 1005 psig was applied to the internal surfaces of the finite element model (FEM) [19]. The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology presented in Reference [1] is used to calculate the pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp can be linearly scaled to determine the KIp for various RPV internal pressures.

The plant-specific HCGS instrument nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analysis can be found in Reference [20]. The following summarizes the development of the thermal and pressure stress intensity factors for the instrument nozzle:

A one-quarter symmetric, three-dimensional finite element model of the instrument nozzle was constructed and is shown in Figure 11. Temperature-dependent material properties, taken from the ASME Code,Section II, Part D, 2001 Edition with 2003 Addenda edition [23], were used in the evaluation and are described in Reference [20].

With respect to operating conditions, the bounding thermal transient for the region corresponding to the instrument nozzles during normal and upset operating conditions was analyzed [20]. The thermal stress distribution, corresponding to the limiting time in Reference [20], along a linear path through the nozzle corner is used. The BIE/IF methodology presented in Reference [1] was used to calculate the thermal stress intensity factor, KIt, due to the thermal stresses by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

Boundary conditions and heat transfer coefficients used for the thermal analysis are described in Reference [20].

With respect to pressure stress, a unit pressure of 1005 psig was applied to the internal surfaces of the finite element model (FEM) [20]. The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology

presented in Reference [1] is used to calculate the pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting KIp can be linearly scaled to determine the KIp for various RPV internal pressures Table 9 summarizes the pressure stress intensity factor and maximum thermal stress intensity factor for feedwater, LPCI and instrument nozzle.

6.0 REFERENCES

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, August 2013. (ADAMS Accession No. ML13277A557)

2. U.S. NRC Letter to BWROG dated May 16, 2013, Final Safety Evaluation for Boiling Water Reactor Owners Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors (TAC NO. ME7649, ADAMS Accession No. ML13277A557).

3. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

4. TransWare Enterprises Inc, HCK-FLU-001-R-002, Revision 0, August 2024, Hope Creek Nuclear Generating Station Reactor Pressure Vessel Fluence Evaluation with Fluence Overpower Adjustments, SI File No. 2300238.201.

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

6. SI Calculation No. 2300238.301, Revision 0, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts for 44 and 56 EFPY, September 27, 2024.

7. SI Calculation No. 2300238.306, Revision 0, Hope Creek Updated P-T Curve Calculation for 44 and 56, December 16, 2024.

8. Title 10, Code of Federal Regulations, Part 50, Appendix G, Fracture Toughness Requirements, November 29, 2019.

9. ASME Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G, Fracture Toughness Criteria for Protection Against Failure, 2007 Edition with Addenda through 2008.

10. U.S. Code of Federal Regulations, Title 10, Part 50, Section 59, Changes, tests and experiments, August 28, 2007.

11. PSEG Nuclear LLC, Revision 1-NP, Pressure and Temperature Limits Report (PTLR) for 32, 44, and 56 Effective Full-Power Years (EFPY), September 11, 2017. (ADAMS Accession No. ML18159A410)

12. Hope Creek License Amendment No. 151, Revision to the Reactor Pressure Vessel Material Surveillance Program, dated July 23, 2004. (TAC No. MB7151, ADAMS Accession No. ML033230591)

13. BWRVIP-135, Revision 4: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. EPRI PROPRIETARY INFORMATION.

14. EPRI Document, BWRVIP 2023-039, Change in BWRVIP-135 Proprietary Information, May 12, 2023. EPRI PROPRIETARY INFORMATION.

15. PSEG Document No. VTD 431282, SI Calculation No. 0800118.317, Revision 0, Validation of Beltline Materials, September 12, 2008.

16. Hope Creek Generating Station, Issuance of Amendment 88 (TAC No. M93054, ADAMS Accession No. ML011760521), November 28, 1995.

17. ANSYS Mechanical APDL (UP20170403) and Workbench (March 31, 2017), Release 18.1, SAS IP, Inc.

18. SI Calculation No. 2300238.303, Revision 0, Feedwater Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development, September 11, 2024.

19. SI Calculation No. 2300238.304, Revision 0, Low Pressure Coolant Injection Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development, October 03, 2024.

20. SI Calculation No. 2300238.305, Revision 0, Instrument Nozzle Fracture Mechanics Evaluation for Pressure-Temperature Limit Curve Development, October 30, 2024.

21. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix B, Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants.

22. U.S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, License Qualification for Performing Safety Analyses, June 24, 1999.

23. ASME Boiler and Pressure Vessel Code,Section II, Part D, Material Properties, 2001 Edition with Addenda through 2003.

24. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, January 31, 2008.

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26. BWRVIP-298NP: BWR Vessel and Internals Project: Testing and Evaluation of the Hope Creek 120° Capsule, EPRI, Palo Alto, CA: 2016. 3002007844. SI File No. 1601009.201.

27. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI PROPRIETARY INFORMATION.

Figure 1: HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 44 EFPY [7]

0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 0