JAFP-25-0056, Pressure and Temperature Limits Report (PTLR) Revision 2

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Pressure and Temperature Limits Report (PTLR) Revision 2
ML25339A033
Person / Time
Site: FitzPatrick 
Issue date: 12/04/2025
From: Mark Hawes
Constellation Energy Generation
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
JAFP-25-0056
Download: ML25339A033 (0)
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Text

Constellation".

JAFP-25-0056 December 4, 2025 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333 James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Mark R. Hawes Regulatory Assurance

Subject:

Pressure and Temperature Limits Report (PTLR) Revision 2 Enclosed is Revision 2 to the James A. FitzPatrick (JAF) Pressure and Temperature Limits Report (PTLR). This revision incorporates analysis completed for bounding values at forty-eight (48) Effective Full Power Years (EFPY), and is submitted in accordance with JAF Technical Specification (TS) 5.6.7.

There are no new regulatory commitments contained in this letter. Questions concerning this report may be addressed to Mr. Mark Hawes, (315) 349-6659.

Very truly yours,

-ltA Mark R. Hawes Regulatory Assurance MRH

Enclosure:

Pressure and Temperature Limits Report Revision 2 cc: (w/ enclosure)

NRC Regional Administrator, NRC Region 1 NRC Project Manager NRC Resident Inspector NYSERDA NYSPSC

JAFP-25-0056 ENCLOSURE Pressure and Temperature Limits Report Revision 2 (31 Pages)

CONSTELLATION ENERGY JAMES A. FITZPATRICK NUCLEAR POWER PLANT REPORT PRESSURE AND TEMPERATURE LIMITS REPORT CPTLR) UP TO 48 EFFECTIVE FULL-POWER YEARS REVISION 2 Prepared by :

Date: JJ./JJJ_Wt-45" Reviewed by :

Approved bfu_

Date:

Date:

JAF Pressure-Temperature Limits Report Revision 2 Page 2 of 31 Table of Contents Section Page 1.0 Purpose 3

2.0 Applicability 3

3.0 Methodology 4

4.0 Operating Limits 5

5.0 Discussion 6

6.0 References 14 Figure 1 JAF Pressure Test (Curve A) --- 48 EPFY 16 Figure 2 JAF Normal Operation Core Not Critical (Curve B) --- 48 EFPY 17 Figure 3 JAF Normal Operation Core Critical (Curve C) --- 48 EFPY 18 Table 1 JAF Pressure Test (Curve A) --- 48 EFPY 19 Table 2 JAF Core Not Critical (Curve B) --- 48 EFPY 22 Table 3 JAF Core Critical (Curve C) --- 48 EFPY 25 Table 4 JAF ART Calculations for 48 EFPY (Reference 6.10) 28 Table 5 JAF ART Calculations for N16 Nozzle for 48 EFPY (Reference 6.3) 29 Appendix A FitzPatrick Reactor Vessel Materials Surveillance Program 30

JAF Pressure-Temperature Limits Report Revision 2 Page 3 of 31 1.0 PURPOSE The purpose of the James A. Fitzpatrick Nuclear Power Plant Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

  • RCS Heatup and Cooldown rates
  • RPV bottom head coolant temperature to RPV coolant T requirements during Recirculation Pump startups
  • RPV head flange boltup temperature limits This report has been prepared in accordance with the requirements of Licensing Topical Report SIR-05-044, Revision 1-A, contained within BWROG-TP-11-022-A, Revision 1 (Reference 6.1), as well as the methodology described in SI Calculation 0800846.301, Revision 1, (Reference 6.2), thereby satisfying JAF Technical Specification (TS) Section 5.6.7.

2.0 APPLICABILITY This report is applicable to the JAF RPV for up to 48 Effective Full-Power Years (EFPY).

The following JAF TS is affected by the information contained in this report:

  • Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 RCS Pressure and Temperature (P/T) Limits, The JAF Reactor Vessel Pressure and Temperature Limits for 40 to 54 EFPY have been developed per Reference 6.3. The 48 EFPY Limits are incorporated in this revision of the PTLR. Future revisions of the PTLR must be revised per the 10CFR50.59 Review process as applicable.

JAF Pressure-Temperature Limits Report Revision 2 Page 4 of 31 3.0 METHODOLOGY The limits in this report were derived as follows:

1) The methodology used is in accordance with Reference 6.1, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, incorporating the NRC Safety Evaluation in Reference 6.4.
2) The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190), Reference 6.5, using the RAMA computer code, as documented in References 6.6 and 28.
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), Reference 6.8, and supplemented by data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) in Reference 6.9. ART calculations are documented in References 6.3 and 6.10.
4) The pressure and temperature limits were calculated in accordance with References 6.1 and 6.2, as documented in Reference 6.3.
5) This revision of the pressure and temperature limits is to incorporate the following changes:
  • Revision 0: Initial issue of PTLR
  • Revision 1:

o Incorporated P-T limits for 40 EFPY and removed curves for 32 EFPY.

o Incorporated P-T limits for instrument (N16) nozzle and description of nozzle stress analyses.

o Updated information on BWRVIP ISP surveillance capsule data for JAF representative materials.

  • Revision 2:

o Incorporated P-T limits for 48 EFPY and removed curves for 40 EFPY.

o Added/Removed references as necessary.

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

JAF Pressure-Temperature Limits Report Revision 2 Page 5 of 31 Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59 (Reference 6.11),

provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon revised RPV fluence calculation methodology, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 OPERATING LIMITS The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel coolant 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, in accordance with 10 CFR 50, Appendix G (Reference 6.12).

Complete P-T curves were developed for 48 EFPY for JAF, as documented in Reference 6.3. The JAF P-T curves for 48 EFPY are provided in Figures 1 through 3, and a tabulation of the curves is included in Tables 1 through 3. For Curves A, B, and C in Tables 1, 2, and 3, respectively, three tables are included for the beltline, bottom head, and upper vessel/feedwater nozzle (non-beltline) regions.

The resulting P-T curves are based on the geometry, design and materials information for the JAF vessel with the following conditions:

Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing (Figure 1: Curve A): 20°F/hour1.

1 Interpreted as the temperature change in any 1-hour period is less than or equal to 20°F.

JAF Pressure-Temperature Limits Report Revision 2 Page 6 of 31 Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): 100°F/hour2.

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

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

RPV head installation temperature limit (Figure 1: Curve A - Hydrostatic and Class 1 Leak Testing; Figure 2: Curve B - non-nuclear heating): 70°F.

RPV flange and adjacent shell criticality temperature limit (Figure 3: Curve C -

nuclear heating): 90°F.

The minimum bolt-up temperature is selected to address the NRC condition in Section 4.0 of Reference 6.4 regarding lowest service temperature (LST) for all ferritic components of the reactor coolant pressure boundary (RCPB), including piping and other non-RPV components. The minimum temperature is set to 70°F for Curves A and B, which is equal to the closure stud LST (Reference 6.13, Table 3-3), and 90°F for Curve C, which is equal to RTNDT,max + 60°F, where RTNDT,max (the bounding nil-ductility reference temperature for the closure flange region) is 30°F. These temperatures bound the minimum temperature limits and minimum bolt-up temperatures in the current docketed P-T curves (Reference 6.14, approved by the NRC in Reference 6.15). These temperatures also bound the non-RPV ferritic components of the RCPB, such as piping, which was impact tested at a temperature that is 60°F below the anticipated minimum service temperature, per the fabrication requirements described in the JAFNPP FSAR (Reference 6.16, Section 16.5.10).

5.0 DISCUSSION The ART of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. RG 1.99 (Reference 6.8) 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.

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

JAF Pressure-Temperature Limits Report Revision 2 Page 7 of 31 The JAF reactor vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the JAF vessel plate and weld materials (Reference 6.10). The Cu and Ni values were used with Tables 1 and 2 of RG 1.99 to determine a chemistry factor (CF) per Regulatory Position 1.1 of RG 1.99 for welds and plates, respectively (Reference 6.8).

However, for materials where surveillance data exists (i.e. JAF plate heat number C3278-2 and weld heat number 13253/12008), a fitted CF has been used in the ART calculations for those heats, in accordance with Regulatory Position 2.1 in RG 1.99 (Reference 6.8).

The peak RPV ID fluence used in the P-T curve evaluation for 48 EFPY is 2.81x1018 n/cm2 for JAF (Table 3 of Reference 6.10). Fluence values were taken from the peak value for 54 EFPY provided in Reference 6.6, which were calculated using methods that comply with the guidelines of RG 1.190 (Reference 6.5). An ID fluence value of 2.30x1018 n/cm2 applies to the limiting beltline lower shell plate C3376-2 for JAF. This fluence value was adjusted for the limiting lower shell plates based upon an attenuation factor of 0.682 for a postulated flaw with depth 1/4 of the wall thickness (1/4t). As a result, the 1/4t 48 EFPY fluence for the limiting beltline lower shell plate is 1.569x1018 n/cm2 for JAF. An ID fluence value of 2.123x1018 n/cm2 applies to the limiting beltline lower shell weld 27204/12008. This fluence value was adjusted based upon an attenuation factor of 0.724 for a postulated 1/4t flaw. As a result, the 1/4t 48 EFPY fluence for the limiting beltline lower shell weld is 1.538x1018 n/cm2 for JAF. The limiting ART for the JAF beltline for 48 EFPY is 125.5°F, corresponding to the lower shell weld 27204/12008 (Reference 6.10).

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. JAF has one set of nozzles in the RPV beltline, the instrument (N16) nozzles located in the lower-intermediate shell beltline plates (Reference 6.3). The feedwater (FW) nozzle is considered in the evaluation of the non-beltline (upper vessel) region P-T limits.

The N16 nozzles and welds are fabricated from non-ferritic materials and do not require specific evaluation. However, the geometric discontinuity caused by the penetrations in the adjacent plates is considered in the development of bounding beltline P-T limits as

JAF Pressure-Temperature Limits Report Revision 2 Page 8 of 31 described in Reference 6.3. The N16 nozzle locations have a limiting RPV ID fluence of 6.66x1017 n/cm2 at 48 EFPY (Reference 6.3). For the N16 nozzles, a plant-specific damage assessment, in terms of displacements per atom (dpa), was performed to determine through-wall fluence in Reference 6.7, as permitted by RG 1.99. The ratio of the fluence at the 1/4t depth to the fluence at the ID is 0.682 for a postulated 1/4t flaw in the N16 nozzle corner. Consequently, the 1/4t fluence for 48 EFPY for the N16 nozzle locations is 4.54x1017 n/cm2 (Reference 6.3). The limiting value for ART for the N16 nozzles is 39.7°F for 48 EFPY (Reference 6.3).

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 heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4t location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4t location. This approach is conservative because irradiation effects cause the allowable toughness at 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, which is well above the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves are developed based on a coolant heatup 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 curve (Curve A), a coolant heatup and cooldown temperature rate of 20°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

JAF Pressure-Temperature Limits Report Revision 2 Page 9 of 31 pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial RTNDT, chemistry (weight-percent Cu and Ni) and adjusted reference temperature at the 1/4t location for 48 EFPY for all RPV beltline plates and welds significantly affected by fluence (i.e., fluence > 1017 n/cm2 for E > 1 MeV) are shown in Table 4. Table 5 shows the ART calculations for the N16 nozzles, based on the fluence at the nozzle locations and the material parameters of the adjacent vessel plates. The initial RTNDT values shown in Tables 4 and 5 (obtained from Reference 6.13) have been previously approved for use by the NRC (Reference 6.17).

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) representative weld and plate surveillance material data for JAF were reviewed from BWRVIP-135, Revision 4 (Reference 6.9), and in accordance with Appendix A of Reference 6.1. Although Reference 6.10 cites Revision 1 of BWRVIP-135, there were no changes in the latest revision that affected the JAF ART calculation. Use of surveillance data from the BWRVIP ISP for JAF was approved by the NRC in Reference 6.18.

For the JAF target plate material G-3415-3, (heat number C3376-2), the ISP representative plate material (C6345-1) is not the same heat number as the target plate, nor does the ISP representative heat exist in the JAF beltline. For the JAF target weld materials 2-233A, B &

C (heat number 27204/12008), the ISP representative weld material (27204) is not the same heat number as the target weld. Therefore, the CF values for the target plate and weld (heat numbers C3376-2 and 27204/12008) were calculated using table values from RG 1.99 Regulatory Position 1.1 (Reference 6.8).

However, surveillance test data is available for two heats which do exist in the JAF beltline:

plate heat number C3278-2 and weld heat number 13253/12008, with data provided in Reference 6.9.

JAF Pressure-Temperature Limits Report Revision 2 Page 10 of 31 For RPV plate G-3414-2 (heat number C3278-2), test data is available for two capsules from the JAF surveillance program and six capsules from the Supplemental Surveillance Program (SSP) (Reference 6.9). The fitted CF of 89.56°F for this material bounds the CF from the RG 1.99 tables, therefore, the higher surveillance-based CF is used for plate heat C3278-2. The scatter in the surveillance data exceeds the credibility criteria, so a standard full margin ( = 17.0 °F) was used in the ART calculations for plate heat number C3278-2 as shown in Table 4 (Reference 6.10).

The surveillance-based CF for plate heat number C3278-2 is also used in the calculation of ART for instrument nozzle N16A (

Table 5), which is based on material parameters for the adjacent shell plate and fluence at the nozzle location. Consequently, RTNDT and for the N16A nozzle differ from those calculated for plate heat number C3278-2 in Table 4.

For RPV welds 1-233A, B & C (heat number 13253/12008), test data for two capsules is available from the SSP capsules (Reference 6.9). A fitted CF of 323.68°F was reported for the 13253/12008 weld heat (Reference 6.9). Since the surveillance weld chemistry differs slightly from the vessel best-estimate weld chemistry, an adjusted chemistry factor of 326.94°F was calculated using the ratio procedure (Reference 6.10). Because the surveillance data is credible, the margin term was cut in half ( = 14.0°F) in calculations of the ART values for weld heat number 13253/12008 as shown in Table 4 (Reference 6.10). The RG 1.99 table CFs were used in the determination of the ART values for all JAF beltline materials except for plate heat C3278-2 and weld heat 13253/12008.

The only computer code used in the determination of the JAF P-T curves was the ANSYS finite element (FE) computer program for the FW nozzle (non-beltline) and instrument (N16) nozzle stresses (various code versions as identified below).

ANSYS finite element analyses (FEA) were performed to determine through-wall thermal and pressure stress distributions for the JAF FW nozzle (Reference 6.19) and for the instrument (N16) nozzles (Reference 6.2). These stress distributions were used in the

JAF Pressure-Temperature Limits Report Revision 2 Page 11 of 31 determination of the stress intensity factors for the FW nozzle (References 6.3 and 6.20) and N16 nozzles (Reference 6.21).

The plant-specific JAF FW nozzle analyses were performed to determine through-wall pressure and thermal stress distributions due to bounding thermal transients. Detailed information regarding the analyses and results can be found in References 6.20, 6.19, 6.22. The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle:

  • With respect to operating conditions, stress distributions for a thermal shock of 450°F represents the maximum thermal shock for the FW nozzle during normal and upset operating conditions. The stress results for a 450°F shock are appropriate for use in developing the non-beltline P-T curves based on the limiting FW nozzle, as a shock of 450°F is representative of the Turbine Roll transient during startup, which produces the highest tensile stresses at the FW nozzle 1/4t location. Because operation is along the saturation curve, these stresses are scaled to reflect the worst-case step change due to the available temperature difference. Therefore, these stresses represent bounding stresses in the FW nozzle associated with 100°F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel region. The boundary integral equation/influence function (BIE/IF) methodology presented in Reference 6.1 was used in Reference 6.20 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.
  • Heat transfer coefficients used in the Reference 6.19 analyses were obtained from a GE evaluation of overall heat transfer coefficients for various BWR FW nozzle thermal sleeve configurations. The configuration used in the plant-specific analyses in Reference 6.19 represents a triple thermal sleeve sparger with seal no. 1 failed.
  • With respect to pressure stress, a unit pressure of 1,000 psig was applied to the internal surfaces of the FE model. The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology presented in Reference 6.1 was used in Reference 6.20 to calculate the pressure stress intensity factor, KIp, by fitting a third order polynomial equation to the path stress

JAF Pressure-Temperature Limits Report Revision 2 Page 12 of 31 distribution for the pressure load case. The resulting KIp can be linearly scaled to determine the KIp for various RPV internal pressures.

  • A two-dimensional axisymmetric FE model of the FW nozzle was constructed in Reference 6.19 and evaluated using the ANSYS code (Reference 6.23). The pressure stress distribution used to determine KIp reflects a scaling factor of 3.0 to account for three-dimensional effects on the pressure stresses (Reference 6.3).

Details of the model are provided in Reference 6.19.

The plant-specific JAF instrument (N16) nozzle analysis was performed to determine through-wall pressure and thermal stress distributions due to bounding thermal transients.

Detailed information regarding the analysis can be found in Reference 6.2. The following summarizes the development of the thermal and pressure stress intensity factors for the N16 nozzles:

  • A one-quarter symmetric, three-dimensional FE model of the N16 nozzle was constructed in Reference 6.2 using the ANSYS code (Reference 6.24).

Temperature-dependent material properties, taken from the ASME Code,Section II, Part D, 2001 Edition with 2003 Addenda (Reference 6.25), were used in the evaluation.

  • With respect to operating conditions, the bounding thermal transient for the region corresponding to the instrument nozzles during normal and upset operating conditions, the SRV blowdown event, was analyzed (Reference 6.2). The thermal stress distribution, corresponding to the limiting time in Reference 6.2, along a linear path through the nozzle corner, is used. The BIE/IF methodology presented in Reference 6.1 was used to calculate the thermal stress intensity factor, KIt, due to the path stress by fitting a third order polynomial 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 6.2.
  • With respect to pressure stress, a unit pressure of 1,000 psig was applied to the internal surfaces of the FE model (Reference 6.2). The pressure stress distribution was taken along the same path as the thermal stress distribution. The BIE/IF methodology presented in Reference 6.1 was used to calculate the pressure stress

JAF Pressure-Temperature Limits Report Revision 2 Page 13 of 31 intensity factor, KIp, due to the path stress by fitting a third order polynomial 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.

JAF Pressure-Temperature Limits Report Revision 2 Page 14 of 31

6.0 REFERENCES

6.1 Licensing Topical Report (LTR) BWROG-TP-11-022-A, Revision 1 (SIR-05-044, Revision 1-A), Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, August 2013, ADAMS Accession No. ML13277A557.

6.2 SI Calculation No. 0800846.301, Revision 1, 2" Instrument Nozzle Stress Analysis, May 7, 2009.

6.3 SI Calculation No. 1400824.301, Revision 0, FitzPatrick Updated P-T Curve Calculation for 40, 48, and 54 EFPY, September 1, 2017.

6.4 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. ML13107A062) 6.5 U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

6.6 TransWare Report No. ENT-FLU-002-R-005, Revision 0, Non-Proprietary Version of James A. FitzPatrick Reactor Pressure Vessel Fluence Evaluation at End Of Cycle 17 and 54 EFPY, October 31, 2007.

6.7 TransWare Report No. EXL-JAF-001-R-001, Revision 1, Determination of Fast Neutron Fluence (>1.0 MeV) in the James A. FitzPatrick Reactor Pressure Vessel N16 Instrumentation Nozzle at 40 EFPY and 54 EFPY, August 16, 2017. TRANSWARE PROPRIETARY INFORMATION.

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

6.9 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. Note: Section 2, Appendix A, and Appendix B of BWRVIP-135, Revision 4 is no longer considered proprietary per EPRI Communication 2023-039.

6.10 SI Calculation No. FITZ-10Q-301, Revision 0, Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts, February 15, 2008.

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

6.12 U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix G, Fracture Toughness Requirements.

6.13 General Electric Report GENE-B1100732-01, Revision 1, Plant Fitzpatrick RPV Surveillance Materials Testing and Analysis of 120 degree capsule at 13.4 EFPY, February 1998.

6.14 Attachment 5 to Entergy Letter JAFP-08-0034, James A. FitzPatrick Nuclear Power Plant Report, Pressure and Temperature Limits Report (PTLR) up to 32 Effective Full-Power Years, Entergy Nuclear Operations, Revision 0, April 14, 2008. (ADAMS Accession No. ML081200881)

JAF Pressure-Temperature Limits Report Revision 2 Page 15 of 31 6.15 JAF License Amendment No. 292, Relocation of Pressure and Temperature Curves to the Pressure and Temperature Limits Report Consistent with TSTF-419-A (TAC No.

MD8556), dated October 3, 2008. (ADAMS Accession No. ML082630365) 6.16 James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report Update, Section 16.5, Pressure Integrity of Piping and Equipment Pressure Parts, Subsection 16.5.10.

6.17 JAF License Renewal Final SER, dated January 24, 2008, Open Item 4.2.2-1 (P-T Limits). (ADAMS Accession No. ML080250372) 6.18 JAF License Amendment No. 285, Changes to the Reactor Vessel Material Surveillance Program, dated July 26, 2006. (TAC No. MC9682, ADAMS Accession No. ML061710335) 6.19 GE Report NEDC-30799-P, James A. FitzPatrick Nuclear Power Station Feedwater Nozzle Fracture Mechanics Analysis to Show Compliance with NUREG-0619, December 1984. GE PROPRIETARY INFORMATION.

6.20 SI Calculation No. FITZ-10Q-302, Revision 0, Revised Pressure-Temperature Curves, February 26, 2008.

6.21 SI Calculation No. 0800846.302, Revision 1, Comparison of Instrument Nozzle (N16) and Beltline P-T curves for 32 EFPY, May 13, 2009.

6.22 SI Calculation No. NYPA-62Q-301, Revision 0, Benchmark Analysis, January 18, 1999.

6.23 G.J. DeSalvo and J.A. Swanson, ANSYS Engineering Analysis System Users Manual, Swanson Analysis Systems, Inc., March 1975.

6.24 ANSYS Release 8.1 (w/Service Pack 1), ANSYS, Inc., June 2004.

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

6.26 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.

6.27 U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements.

6.28 TransWare Document No. JAF-FLU-001-R-002, Revision 0, James A. FitzPatrick Nuclear Power Plant Reactor Pressure Vessel Fluence Evaluation, May 2023

JAF Pressure-Temperature Limits Report Revision 2 Page 16 of 31 Figure 1: JAF Pressure Test (Curve A) --- 48 EPFY

JAF Pressure-Temperature Limits Report Revision 2 Page 17 of 31 Figure 2: JAF Normal Operation Core Not Critical (Curve B) --- 48 EFPY

JAF Pressure-Temperature Limits Report Revision 2 Page 18 of 31 Figure 3: JAF Normal Operation Core Critical (Curve C) --- 48 EFPY

JAF Pressure-Temperature Limits Report Revision 2 Page 19 of 31 Table 1a: JAF Pressure Test (Curve A) --- Beltline (48 EFPY)

Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 575.6 103.5 625.0 114.5 674.3 123.4 723.7 131.0 773.1 137.6 822.4 143.4 871.8 148.7 921.2 153.4 970.6 157.7 1019.9 161.7 1069.3 161.7 1118.7 165.4 1168.0 168.8 1217.4 172.0 1266.8 175.0 1316.1

JAF Pressure-Temperature Limits Report Revision 2 Page 20 of 31 Table 1b: JAF Pressure Test (Curve A) --- Bottom Head (All EFPY)

Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 677.3 74.9 726.5 79.4 775.7 83.5 824.9 87.3 874.1 90.8 923.4 94.1 972.6 97.2 1021.8 100.1 1071.0 102.8 1120.2 105.4 1169.4 107.9 1218.6 110.3 1267.8 112.5 1317.0

JAF Pressure-Temperature Limits Report Revision 2 Page 21 of 31 Table 1c: JAF Pressure Test (Curve A) --- Non-Beltline (All EFPY)

Curve A - Pressure Test P-T Curve Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 312.6 120.0 312.6 120.0 1532.8

JAF Pressure-Temperature Limits Report Revision 2 Page 22 of 31 Table 2a: JAF Core Not Critical (Curve B) --- Beltline (48 EFPY)

Curve B - Core Not Critical P-T Curve Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 254.5 77.8 300.0 84.5 345.4 90.4 390.9 107.9 439.7 120.8 488.6 131.1 537.4 139.6 586.3 146.9 635.1 153.2 683.9 158.9 732.8 163.9 781.6 168.5 830.4 172.7 879.3 176.6 928.1 180.2 977.0 183.6 1025.8 186.7 1074.6 189.7 1123.5 192.5 1172.3 195.1 1221.1 197.7 1270.0 200.1 1318.8

JAF Pressure-Temperature Limits Report Revision 2 Page 23 of 31 Table 2b: JAF Core Not Critical (Curve B) --- Bottom Head (All EFPY)

Curve B - Core Not Critical P-T Curve Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 491.6 76.4 540.3 82.0 589.0 87.1 637.7 91.7 686.4 96.0 735.1 99.9 783.8 103.5 832.5 106.8 881.2 110.0 929.9 113.0 978.6 115.8 1027.3 118.4 1076.0 120.9 1124.7 123.3 1173.4 125.6 1222.1 127.8 1270.8 129.9 1319.5

JAF Pressure-Temperature Limits Report Revision 2 Page 24 of 31 Table 2c: JAF Core Not Critical (Curve B) --- Non-Beltline (All EFPY)

Curve B - Core Not Critical P-T Curve Temperature P-T Curve Pressure

°F psi 70.0 0.0 70.0 233.9 77.3 273.3 83.5 312.6 150.0 312.6 150.0 1382.7

JAF Pressure-Temperature Limits Report Revision 2 Page 25 of 31 Table 3a: JAF Core Critical (Curve C) --- Beltline (48 EFPY)

Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure

°F psi 90.0 0.0 90.0 165.3 101.1 210.4 110.2 255.5 117.9 300.7 124.5 345.8 130.4 390.9 147.9 439.7 160.8 488.6 171.1 537.4 179.6 586.3 186.9 635.1 193.2 683.9 198.9 732.8 203.9 781.6 208.5 830.4 212.7 879.3 216.6 928.1 220.2 977.0 223.6 1025.8 226.7 1074.6 229.7 1123.5 232.5 1172.3 235.1 1221.1 237.7 1270.0 240.1 1318.8

JAF Pressure-Temperature Limits Report Revision 2 Page 26 of 31 Table 3b: JAF Core Critical (Curve C) --- Bottom Head (All EFPY)

Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure

°F Psi 90.0 0.0 90.0 373.6 99.4 423.2 107.3 472.7 114.1 522.3 120.1 571.8 125.5 621.4 130.3 671.0 134.7 720.5 138.8 770.1 142.5 819.6 146.0 869.2 149.3 918.7 152.3 968.3 155.2 1017.9 158.0 1067.4 160.5 1117.0 163.0 1166.5 165.3 1216.1 167.6 1265.7 169.7 1315.2

JAF Pressure-Temperature Limits Report Revision 2 Page 27 of 31 Table 3c: JAF Core Critical (Curve C) --- Non-Beltline (All EFPY)

Curve C - Core Critical P-T Curve Temperature P-T Curve Pressure

°F psi 90.0 0.0 90.0 155.9 101.3 195.1 110.1 234.2 117.3 273.4 123.5 312.6 190.0 312.6 190.0 1382.7

JAF Pressure-Temperature Limits Report Page 28 of 31 Table 4: JAF ART Calculations for 48 EFPY (Reference 6.10)

Description Code No.

Heat No.

Flux Type & Lot No.

Initial RTNDT

(°F)

Chemistry Chemistry Factor

(°F)

Adjustments For 1/4t RTNDT Margin Terms ART Cu (wt%)

Ni (wt%)

(°F)

(°F) i (°F)

(°F)

Plates Lower Shell #1 G-3415-1R C3394-1

-10.0 0.11 0.56 73.60 37.8 17.0 0.0 61.8 Lower Shell #2 G-3415-3 C3376-2 24.0 0.13 0.60 91.00 46.7 17.0 0.0 104.7 Lower Shell #3 G-3415-2 C3103-2

-2.0 0.14 0.57 98.65 50.6 17.0 0.0 82.6 Lower-Int. Shell #1 G-3413-7 C3368-1

-10.0 0.12 0.50 81.00 46.5 17.0 0.0 70.5 Lower-Int. Shell #2 G-3414-2 C3278-2

-10.0 0.11 0.61 89.56 51.4 17.0 0.0 75.4 Lower-Int. Shell #3 G-3414-1 C3301-1

-18.0 0.18 0.57 131.15 75.2 17.0 0.0 91.2 Description Code No.

Heat No.

Flux Type & Lot No.

Initial RTNDT

(°F)

Chemistry Chemistry Factor

(°F)

Adjustments For 1/4t RTNDT Margin Terms ART Cu (wt%)

Ni (wt%)

(°F)

(°F) i (°F)

(°F)

Welds L. Int. Shell Long. Weld #1 1-233-A 13253/12008 Flux 1092 Lot 3947

-50.0 0.21 0.873 326.94 145.2 14.0 0.0 123.2 L. Int. Shell Long. Weld #2 1-233-B 13253/12008 Flux 1092 Lot 3947

-50.0 0.21 0.873 326.94 134.5 14.0 0.0 112.5 L. Int. Shell Long. Weld #3 1-233-C 13253/12008 Flux 1092 Lot 3947

-50.0 0.21 0.873 326.94 145.2 14.0 0.0 123.2 L. Int./L. Shell Girth Weld 1-240 305414 Flux 1092 Lot 3947

-50.0 0.337 0.609 209.11 110.1 28.0 0.0 116.1 Lower Shell Long. Weld #1 2-233-A 27204/12008 Flux 1092 Lot 3774

-48.0 0.219 0.996 231.06 117.5 28.0 0.0 125.5 Lower Shell Long. Weld #2 2-233-B 27204/12008 Flux 1092 Lot 3774

-48.0 0.219 0.996 231.06 99.3 28.0 0.0 107.3 Lower Shell Long. Weld #3 2-233-C 27204/12008 Flux 1092 Lot 3774

-48.0 0.219 0.996 231.06 94.9 28.0 0.0 102.9 Wall Thickness (in)

Fluence at ID Attenuation, 1/4t Fluence at 1/4t Fluence Factor, FF Location Full 1/4t (n/cm2) e-0.24x (n/cm2) f (0.28 - 0.10 log f)

Plates Lower Shell #1 6.375 1.594 2.30E+18 0.682 1.569E+18 0.513 Lower Shell #2 6.375 1.594 2.30E+18 0.682 1.569E+18 0.513 Lower Shell #3 6.375 1.594 2.30E+18 0.682 1.569E+18 0.513 Lower-Int. Shell #1 5.375 1.344 2.81E+18 0.724 2.035E+18 0.574 Lower-Int. Shell #2 5.375 1.344 2.81E+18 0.724 2.035E+18 0.574 Lower-Int. Shell #3 5.375 1.344 2.81E+18 0.724 2.035E+18 0.574 Welds L. Int. Shell Long. Weld #1 5.375 1.344 1.578E+18 0.724 1.143E+18 0.444 L. Int. Shell Long. Weld #2 5.375 1.344 1.342E+18 0.724 9.724E+17 0.411 L. Int. Shell Long. Weld #3 5.375 1.344 1.578E+18 0.724 1.143E+18 0.444 L. Int./L. Shell Girth Weld 5.375 1.344 2.30E+18 0.724 1.666E+18 0.527 Lower Shell Long. Weld #1 5.375 1.344 2.123E+18 0.724 1.538E+18 0.508 Lower Shell Long. Weld #2 5.375 1.344 1.471E+18 0.724 1.066E+18 0.430 Lower Shell Long. Weld #3 5.375 1.344 1.339E+18 0.724 9.699E+17 0.411

JAF Pressure-Temperature Limits Report Page 29 of 31 Table 5: JAF ART Calculations for N16 Nozzle for 48 EFPY (Reference 6.3) 48 EFPY Description Heat No.

Plate Location Initial RTNDT (°F)

Chemistry Chemistry Factor

(°F)

Adjustments For 1/4t RTNDT Margin ART Cu (wt %)

Ni (wt %)

(°F)

(°F) i

(°F)

(°F)

Nozzle N16B Lower-int shell #1, C3368-1

-10 0.12 0.50 81.00 22.5 10.2 0.0 35.0 Nozzle N16A Lower-int Shell #2, C3278-2

-10 0.11 0.61 89.56 24.9 11.3 0.0 39.7 Location Wall Thickness (in.)

Fluence at ID Fluence @ 1/4t Fluence Factor, FF Full 1/4t (n/cm2)

(n/cm2) f(0.28-0.10log f)

Nozzle N16B 5.375 1.344 6.66E+17 4.54E+17 0.278 Nozzle N16A 5.375 1.344 6.66E+17 4.54E+17 0.278

JAF Pressure-Temperature Limits Report Page 30 of 31 APPENDIX A JAF Reactor Vessel Material Surveillance Programs In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements (Reference 6.27), two surveillance capsules have been removed from the JAF RPV. The first surveillance capsule at JAF was removed in April 1985 after 5.98 EFPY and the second capsule was removed in November 1996 after 13.4 EFPY (Reference 6.13). Both surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel core beltline region.

The flux wires and test specimens removed from the capsules were tested according to the latest version of ASTM E185. The methods and results of testing are presented in Reference 6.13, as required by 10 CFR 50, Appendices G and H (References 6.12 and 6.27). There are two remaining JAF surveillance capsules (the capsule removed in 1996 was re-constituted and returned to the vessel in 1998) which will remain in place to serve as backup surveillance material for the BWRVIP program, or as otherwise needed.

JAF committed to replace its original RPV material surveillance program with the BWRVIP ISP (Reference 6.26) in the license amendment issued by the NRC regarding implementation of the BWRVIP ISP, dated July 26, 2006 (Reference 6.18), and has made a licensing commitment to use the ISP for JAF during the period of extended operation. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for integrated surveillance programs, and has been approved by the NRC. Under the ISP, there are no further capsules scheduled for removal from the JAF reactor vessel.

Representative surveillance materials for the JAF target beltline plate are contained in the LaSalle 1 surveillance capsules. The next LaSalle 1 surveillance capsule is scheduled to be withdrawn and tested under the ISP in approximately 2030 at 33 EFPY (Reference 6.26). Representative surveillance materials for the JAF target beltline weld

JAF Pressure-Temperature Limits Report Page 31 of 31 were contained in the SSP-D and SSP-I capsules. No further SSP capsules are scheduled for withdrawal (Reference 6.26).

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