Text

License Amendment Request to Amend the Technical Specifications (TS) to Revise and Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report
	ML17086A364
Person / Time
Site:	Hope Creek
Issue date:	03/27/2017
From:	Carr E; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML17086A363	List: LR-N17-0032, License Amendment Request to Amend the Technical Specifications (TS) to Revise and Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report;
References
	LAR H17-02, LR-N17-0032
	Download: ML17086A364 (100)
	v • d • e

Attachment 6 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 6, this document is decontrolled.

PSEO Nuclear LLC P.O.l3ox 236, Hancoolw Bridge, New Jersey 08038*0236 PSIG Nudear!LC MAR 27 2017 10 CFR 50.90 LR-N17-0032 LAR H17-02 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:

License Amendment Request to Amend the Hope Creek Technical Specifications (TS) to Revise and Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications {TS) for Hope Creek Generating Station (Hope Creek).

The proposed amendment would revise the reactor coolant system Pressure-Temperature (P-T)

Limit Curves (Figures 3.4.6. 1-1, 3.4.6.1-2, and 3. 4.6.1-3) and relocate them to a Pressure and Temperature Limits Report {PTLR). provides an evaluation supporting the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes and are being provided for information only. Attachment 4 provides the Hope Creek Pressure and Temperature Limits Report (Non-Proprietary). Attachment 5 includes an affidavit from Electric Power Research Institute (EPRI) requesting withholding the proprietary information from public disclosure. provides the Hope Creek Pressure and Temperature Limits Report (Proprietary) to be withheld from public disclosure under 10 CFR 2.390(a)(4). contains proprietary information as defined by 10 CFR 2.390(a)(4). EPRI, as the owner of the proprietary information, has executed the Attachment 5 affidavit identifying that the proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. EPRI requests that the proprietary

LR-N17-0032 10 CFR 50.90 Page 2 information in Attachment 6 be withheld from public disclosure, in accordance with the requirements of 10 CFR 2.390(a)(4).

PSEG requests approval of this LAR by February 28, 2018 to be implemented within 60 days of the issue date.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

If you have any questions or require additional information, please contact Ms. Tanya Timberman at 856-339-1426.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on _ __,:-'-?

- ----"?;--./;**_ '--/?-

(Date)

Respectfully, c;--

Z--------- * -*

I Eric S. Carr Site Vice President Hope Creek Generating Station Attachments:

1. Evaluation of Proposed Changes

2. Mark-up of Proposed Technical Specification Pages

3. Mark-up of Proposed Technical Specification Bases Pages

4. Hope Creek Pressure and Temperature Limits Report (Non-Proprietary)

5. Affidavit for Withholding

6. Hope Creek Pressure and Temperature Limits Report (Proprietary) cc: Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Hope Creek Commitment Tracking Coordinator

LR-N17-0032 LAR H17-02 Attachment 1 Evaluation of Proposed Changes Table of Contents

1.0 DESCRIPTION

................................................................................................................1

2.0 PROPOSED CHANGE

....................................................................................................1

3.0 BACKGROUND

...............................................................................................................2

4.0 TECHNICAL ANALYSIS

..................................................................................................3

5.0 REGULATORY ANALYSIS

.............................................................................................6 5.1 No Significant Hazards Consideration .....................................................................6 5.2 Applicable Regulatory Requirements/Criteria...........................................................7 5.3 Precedent ................................................................................................................8

6.0 ENVIRONMENTAL CONSIDERATION

...........................................................................9

7.0 REFERENCES

................................................................................................................9

LR-N17-0032 LAR H17-02

1.0 DESCRIPTION

The proposed change revises the Hope Creek Technical Specifications (TS) by replacing the pressure and temperature (P-T) limit curves (Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3) with references to the Pressure and Temperature Limits Report (PTLR).

The PTLR contains updates to the pressure-temperature (P-T) limit curves for the beltline, bottom head, and non-beltline regions for the Hope Creek reactor pressure vessel (RPV). The P-T curves are developed for 32, 44, and 56 effective full power years (EFPY) of operation. The P-T curves are prepared using the methods documented in the Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044),

Pressure Temperature Limits Report Methodology for Boiling Water Reactors (Reference 1).

This BWROG LTR satisfies the requirements of 10 CFR 50 Appendix G and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, Nonmandatory Appendix G (Reference 2).

The guidance of NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," (Reference 3) was applied during P-T curve development. Also, Technical Specification Task Force (TSTF)

Traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR, (Reference 4) which has received NRC approval, was followed in development of the proposed TS changes.

2.0 PROPOSED CHANGE

This License Amendment Request revises the reactor coolant system Pressure-Temperature Limit Curves (Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3) and relocates them to a Pressure and Temperature Limits Report as follows:

1. Adds a definition in Section 1.0 for the Pressure and Temperature Limits Report. The wording for this definition is consistent with NUREG-1433, Revision 4, Standard Technical Specifications - General Electric BWR/4 Plants.

2. Revises the Index to reflect additions and deletions and pagination changes.

3. Revises TS 3.4.6.1, Reactor Coolant System Pressure/Temperature Limits to refer to the PTLR.

4. Revises TS 3.4.6.1.a, the maximum heatup rate, to refer to limits specified in the PTLR.

5. Revises TS 3.4.6.1.b, the maximum cooldown rate, to refer to limits specified in the PTLR.

6. Revises TS 3.4.6.1.c, the maximum temperature change, to refer to limits specified in the PTLR.

7. Revises TS 3.4.6.1.d, the reactor vessel flange and head flange metal temperature, to refer to limits specified in the PTLR.

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LR-N17-0032 LAR H17-02

8. Revises Surveillance Requirements (SR) 4.4.6.1.1 and 4.4.6.1.2 to refer to the pressure and temperature limits specified in the PTLR.

9. Revises SR 4.4.6.1.3 to refer to the pressure and temperature limit curves specified in the PTLR.

10. Removes the present P-T curves, Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3

11. Adds a new Specification 6.9.1.10, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to subsection 6.9, Reporting Requirements in Section 6.0, Administrative Controls. The new specification is consistent in format and content with the STS and includes:

The individual TSs that address reactor coolant system P-T limits,

References the NRC approved topical report which documents the PTLR methodology, and

Requires the PTLR and any revisions or supplements to be submitted to the NRC.

12. Revises TS Bases 3/4.4.6 to refer to the pressure and temperature limit lines and curves specified in the PTLR. Revises TS Bases to remove Figure B 3/4.4.6-1. Revises TS Bases to remove Table B 3/4.4.6-1. Revises TS Bases to relocate Table B 3/4.4.6-2 to the PTLR.

The marked-up TS pages are provided in Attachment 2. The proposed changes to the TS Bases provided in Attachment 3 are provided for information only; changes to the affected TS Bases pages will be incorporated in accordance with TS 6.15, Technical Specifications (TS)

Bases Control Program.

3.0 BACKGROUND

10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50.

Appendix G is the regulatory basis for P-T curves for light water reactors. Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Appendix G also requires that the reference temperature and Charpy upper-shelf energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel.

In Hope Creek Amendment 157 (Reference 5), the curves were calculated using the General Electric vessel fluence methodology and based on the methodology specified in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Code Cases N-588 and N-640, the 1989 ASME Code,Section XI, Appendix G, and Appendix G of 10 CFR Part 50. Adjusted reference temperatures at the nil ductility transition values were developed 2

LR-N17-0032 LAR H17-02 for the reactor pressure vessel materials in accordance with RG 1.99, Revision 2. The NRC staff approved revising the reactor pressure vessel pressure-temperature limits and extending their validity to 32 effective full-power years. These are the current P-T limits in the Hope Creek TS.

10 CFR 50 Appendix G requires reactor vessel beltline materials to be tested in accordance with the surveillance program requirements of 10 CFR 50 Appendix H.

Generic Letter 96-03 allows plants to relocate their P-T curves and associated numerical limits (such as heatup and cooldown rates) from the plant TS to a PTLR, which is a licensee-controlled document. As stated in GL 96-03, during the development of the improved Standard Technical Specifications (STS), a change was proposed to relocate the P-T limits currently contained in the plant TS to a PTLR. As one of the improvements to the STS, the NRC staff agreed with the industry that the curves may be relocated outside the plant TS to a PTLR so that the licensee could efficiently maintain these limits. One of the prerequisites for having the PTLR option is that the P-T curves and limits be derived using methodologies approved by the NRC, and that the associated licensing topical reports describing the approved methodologies be referenced in the plant TS.

4.0 TECHNICAL ANALYSIS

Generic Letter 96-03 provides regulatory guidance regarding relocation of P-T curves and associated numerical limits (such as heatup and cooldown rates) from plant TS to a PTLR (a licensee controlled document). As stated in GL 96-03, a licensee requesting such a change must satisfy the following three criteria:

1. Have NRC-approved methodologies to reference in the TS.

2. Develop a PTLR to contain the P-T limit curves, associated numerical limits, and any necessary explanation, and

3. Modify applicable sections of the TS accordingly.

Revised pressure-temperature curves were developed for hydrostatic pressure and leak tests, core not critical, and core critical conditions. A report describing the inputs, methodology and results for the revised curves is provided in Attachment 4. The revised curves have been developed for application for 32 EFPY, 44 EFPY, and 56 EFPY.

The P-T curves were prepared using the methods documented in the BWROG LTR, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, BWROG-TP-11-022-A (SIR-05-044), Revision 1. This BWROG LTR provides one of the current NRC-approved BWROG fracture mechanics methodologies for generating P-T curves/limits and allows BWR plants to adopt the PTLR option in accordance with TSTF-419-A and GL 96-03. The LTR satisfies the requirements of 10 CFR 50 Appendix G and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Nonmandatory Appendix G.

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LR-N17-0032 LAR H17-02 The licensing topical report has four sections and four appendices, the content of which is summarized below.

Section 1.0 describes the background and purpose for the LTR. Attachment 1 of GL 96-03 provides seven technical criteria that a methodology should conform to, to develop P-T limits and to be acceptable by the NRC staff. These seven criteria are explained in this section.

Section 2.0 of the BWROG LTR provides the fracture mechanics methodology and its basis for developing P-T limits.

Section 3.0 of the BWROG LTR provides a step-by-step procedure for calculating P-T limit curves. This section indicates that typically three reactor pressure vessel regions are evaluated with respect to P-T limits: (1) the beltline region; (2) the bottom head region; and (3) the non-beltline region.

The Hope Creek P-T curves consider three regions of the vessel: beltline, bottom head, and non-beltline (feedwater nozzle/upper vessel). The Hope Creek PTLR is based on the methodology and template provided in the LTR.

Section 4.0 of the BWROG LTR provides the references.

Appendix A of the LTR provides guidance for evaluating surveillance data.

Appendix B provides a template for development of an acceptable PTLR.

Appendix C provides the revision 0 requests for additional information, along with the respective responses.

Appendix D provides the revision 1 requests for additional information, along with the respective responses Neutron Fluence Calculations:

The neutron fluence calculations were updated using the NRC-approved RAMA Fluence Methodology and in accordance with NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

The fluence is based upon operation for 56 EFPY, with 12 EFPY at 3293 MWt, the original licensed thermal power, 6 EFPY at 3339 MWt, and the remaining 38 EFPY at 3840 MWt. In 2001, Hope Creek performed a measurement uncertainty recapture (MUR) that increased the rated thermal power level to 3339 MWt (Reference 6). In 2008, Hope Creek performed an extended power uprate (EPU) that increased the rated thermal power level to 3840 MWt (Reference 7).

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LR-N17-0032 LAR H17-02 The calculated fast neutron fluences at the end of plant life (56 EFPY) are provided below:

Parameter Fluence (n/cm2)

Peak Surface 1.63x1018 Peak 1/4 T 1.14x1018 Limiting Beltline Material Peak Surface 1.43x1018 Limiting Beltline Material Peak 1/4 T 1.00x1018 Hope Creek 120 Degree Surveillance Capsule Results and Adjusted Reference Temperature:

In 2015, the Hope Creek surveillance capsule located at the 120 degree reactor pressure vessel azimuth was removed from the RPV and sent out for testing in accordance with the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP),

of which Hope Creek is an active member. The capsule test results were obtained from EPRI and used to determine the Adjusted Reference Temperatures (ARTs) for the applicable representative materials. The representative heat of the plate material (5K3238/1) in the ISP is not the same as the target plate material (5K3025/1) for Hope Creek. However, the surveillance heat (5K3238/1) does exist in the Hope Creek 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, the data is credible, and a reduced margin term is used for the plate material when calculating 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 Hope Creek. Scatter in the surveillance data exceeds the credibility criteria for weld heat D53040, however, the fitted Chemistry Factor (CF) bounds the Regulatory Guide (RG) 1.99 CF. Therefore, the higher surveillance-based CF is used in the ART calculation for weld heat D53040, with a full margin term. The RG 1.99 table CFs were used in the determination of the ART values for all Hope Creek beltline materials except for plate heat 5K3238/1 and weld heat D53040.

Upper shelf energy (USE) calculations were performed and confirmed that all USE values are greater than 50 ft-lb throughout RPV life as required by 10 CFR 50 Appendix G. For the limiting USE material, the USE at 56 EFPY is 60.3 ft-lb.

Pressure-Temperature Curve Evaluation:

Three regions of the reactor pressure vessel were evaluated to develop the revised P-T curves:

the beltline region, the bottom head region, and the feedwater nozzle/upper vessel region.

These regions bound all other regions with respect to brittle fracture.

The methodology used to generate the P-T curves in this submittal is approved by the NRC. In this update, the results of the 120 degree surveillance capsule are incorporated into the ART values in accordance with RG 1.99.

The revised P-T curves and outputs from the ISP ensure that adequate RPV safety margins against non-ductile failure will continue to be maintained during normal operations, anticipated operational occurrences, and inservice leak and hydrostatic testing. Together, these measures ensure that the integrity of the reactor coolant system will be maintained for the life of the plant.

5

LR-N17-0032 LAR H17-02 Proposed revisions to applicable sections of the TS have been prepared and are provided in to this submittal. These proposed changes are consistent with the guidance provided in GL 96-03, as supplemented by TSTF-419-A.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) requests an amendment to Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (Hope Creek). The proposed Technical Specification (TS) changes modify the Hope Creek TS by replacing references to existing reactor vessel heatup and cooldown rate limits and Pressure-Temperature (P-T) limit curves with references to the Pressure and Temperature Limits Report (PTLR).

PSEG 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 license amendment adopts the NRC approved methodology described in Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044), Pressure Temperature Limits Report Methodology for Boiling Water Reactors. The Hope Creek PTLR was developed based on the methodology and template provided in the BWROG LTR.

10 CFR 50 Appendix G establishes requirements to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants.

Implementing this NRC approved methodology does not reduce the ability to protect the RCPB as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components.

Incorporation of the new methodology for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions.

The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the manner in which the plant is operated and maintained. The ability of structures, systems, and components to perform their intended safety functions is not altered or prevented by the proposed changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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LR-N17-0032 LAR H17-02

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The change in methodology for calculating P-T limits and the relocation of those limits to the PTLR do not alter or involve any design basis accident initiators. RCPB integrity will continue to be maintained in accordance with 10 CFR 50 Appendix G, and the assumed accident performance of plant structures, systems and components will not be affected. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), and the installed equipment is not being operated in a new or different manner.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed changes do not affect the function of the RCPB or its response during plant transients. Calculating the Hope Creek P-T limits using the NRC approved SI methodology ensures adequate margins of safety relating to RCPB integrity are maintained. The proposed changes do not alter the manner in which the Limiting Conditions for Operation P-T limits for the RCPB are determined. There are no changes to the setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

Based upon the above, PSEG 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.

5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(2), Limiting conditions for operation, states: (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The NRC has established requirements in 10 CFR 50, Appendix G, "Fracture Toughness Requirements," in order to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. Appendix G requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods and margins of safety of Appendix G to Section XI of the American Society of Mechanical Engineers 7

LR-N17-0032 LAR H17-02 Boiler and Pressure Vessel Code were used to generate the P-T limits. Also, Appendix G requires that applicable surveillance data from reactor pressure vessel material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Appendix H to 10 CFR Part 50 provides requirements related to facility RPV material surveillance programs. Hope Creek demonstrates its compliance with the requirements of 10 CFR 50, Appendix H through participation in the BWRVIP ISP (Reference 8) and the latest material information was utilized in preparation of the report.

Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

10 CFR 50.36, "Technical specifications," provides the regulatory requirements for the content required in the TSs which includes limiting conditions for operation (LCO's), surveillance requirements and administrative controls. Previously the plant-specific P-T limits had been incorporated into the TS and controls were placed on operation and testing by the associated specification. This proposed change revises the TS to relocate the P-T limit curves to a licensee controlled document in accordance with the guidance of Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits" and TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR.

Therefore, 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.

5.3 Precedent The NRC has approved similar license amendments to relocate the P-T limit curves to a PTLR.

Recent examples for BWR plants include:

Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, (License Amendment Nos. 277 and 221 issued by NRC letter dated March 23, 2016, ADAMS Accession No. ML16062A099)

Cooper Nuclear Station (License Amendment No. 256 issued by NRC letter dated July 25, 2016, ADAMS Accession No. ML16158A022)

Duane Arnold Energy Center (License Amendment No. 294 issued by NRC letter dated July 25, 2016, ADAMS Accession No. ML16180A086)

Nine Mile Point Nuclear Station, Unit 1 (License Amendment No. 204 issued by NRC letter dated January 21, 2010, ADAMS Accession No. ML093370002) 8

LR-N17-0032 LAR H17-02

6.0 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 significant increase in the amounts of any effluent 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.

7.0 REFERNCES

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

2. ASME Boiler and Pressure Vessel Code,Section XI, Nonmandatory Appendix G, dated July 1 2003

3. NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," dated January 31, 1996

4. TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," dated August 4, 2003

5. Hope Creek Generating Station, Issuance of Amendment Re: Change of Pressure-Temperature Limits and Extension of Validity to 32 Effective Full-Power Years (TAC No.

MC2534), November 1, 2004

6. Hope Creek Generating Station, Issuance of Amendment Re: 1.4% Increase In Licensed Power Level (TAC No. MB0644), July 30, 2001

7. Hope Creek Generating Station, Issuance of Amendment Re: Extended Power Uprate (TAC No. MD3002), May 14, 2008

8. Hope Creek Generating Station, Issuance of Amendment Re: Revision to the Reactor Pressure Vessel Material Surveillance Program (TAC No. MB7151), July 23, 2004 9

LR-N17-0032 LAR H17-02 Attachment 2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License NPF-57 are affected by this change request:

Technical Specification Page Index ii Index xi Index xviii Index xxiv 1.0, Definitions 1-5 3.4.6.1, Pressure/Temperature Limits 3/4 4-21 3.4.6.1, Pressure/Temperature Limits 3/4 4-22 3.4.6.1, Pressure/Temperature Limits 3/4 4-23 3.4.6.1, Pressure/Temperature Limits 3/4 4-23a 3.4.6.1, Pressure/Temperature Limits 3/4 4-23b 6.9.1.10, Pressure/Temperature Limits 6-20 6.9.1.10, Pressure/Temperature Limits 6-21

INDEX DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.26 OFF-GAS RADWASTE TREATMENT SYSTEM........ 1-4 1.27 OFFSITE DOSE CALCULATION MANUAL .................. 1-4 1.28 OPERABLE - OPERABILITY............................ 1-5 1.29 OPERATIONAL CONDITION - CONDITION................. 1-5 1.30 PHYSICS TESTS .................................... 1-5 1.31 PRESSURE BOUNDARY LEAKAGE ........................ 1-5 1.32 PRIMARY CONTAINMENT INTEGRITY .................... 1-5 1.33 PROCESS CONTROL PROGRAM .......................... 1-6 1.34 PURGE-PURGING .................................... 1-6 1.35 RATED THERMAL POWER .............................. 1-6 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME .......... 1-6 1.37 REPORTABLE EVENT ................................. 1-6 1.38 ROD DENSITY ...................................... 1-6 1.39 SECONDARY CONTAINMENT INTEGRITY .................. 1-7 1.40 SHUTDOWN MARGIN .................................. 1-7 1.41 SITE BOUNDARY .................................... 1-7 1.42 Not Used ......................................... 1-8 1.43 SOURCE CHECK ..................................... 1-8 1.44 SPIRAL RELOAD ........................... 1-8 1.45 SPIRAL UNLOAD .................................... 1-8 1.46 STAGGERED TEST BASIS ............................. 1-8 1.47 THERMAL POWER .................................... 1-8 1.48 TURBINE BYPASS SYSTEM RESPONSE TIME .............. 1-9 1.30-1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) .... 1-5 HOPE CREEK ii Amendment No. 121

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System................... , ... ,.,,.,,.... 3/4 4-21 Figure 3.4.6.1-1 Hydrostatic Pressure and Leak Wests Prcsourc/Tcmperaturc Limits Gurvc A 3/4 4-23 Figure 3.4.6.1-2 Non Nuclear Hcatup and Gooldown Preoourc/Tcmperaturc Limits Curve B 3/4 4-23a Figure 3.4.6.1-3 Core Critical Ilca tup and Ccoldown

&1+/--rBi-m-:l:te--*** etrr'lt 3 I4 4-2 3b Table 4.4.6.1.3-1 (Deleted) ................. , ......... 3/4 4-24 Reactor Steam Dome.,,,,,,.,.,,,....... , ..... ,,,,,,, ...,. 3/4 4-25 3/4.4.7 MAIN STEAM LINE ISOLATION VAJ"VES..........,.,,, .... ,.... 3/4 4-26 3/4.4.8 DELETED ... , .. ,.. ,,... ,.,,,,,,, ...... ,,.... ,,.,,,,,,,,.,. 3/4 4-27 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown .. ,., .........,....... ,, ......... ,.,, ..... ,.. 3/4 4-28 Cold Shutdown .. ,.,,.. ,..,.,, ..... ,.,.,., .. ,, ...,,....... 3/4 4-29 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING..., ................... , .... ,, .. ,....... 3/4 5-1 3/4.5.2 ECCS - SHUTDOWN., ........... ,.......,,.,., ......... ,.,,, 3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER., ........... ,....................... 3/4 5-8 3/4,6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary containment Integrity............. . .... , . . .. , . . . 3/4 6-1 Primary Containment Leakage .......... ,. .. , . .. . . . . . . . .. .. 3/4 6-2 Primary Containment Air Locks . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5 Primary Containment Structural Integrity.... , . . . . . ,.. . . . 3/4 6-8 Drywall and Suppression Chamber Internal Pressure * . , . . . . 3/4 6** 9 HOPE CREEK xi Amendment No. 193

INDEX BASES SECTION PAGE INSTRUMENTATION (Continued)

Remote Shutdown Monitoring Instrumentation and Controls ........................................... B 3/4 3-5 Accident Monitoring Instrumentation ...................... B 3/4 3-5 Source Range Monitors .................................... B 3/4 3-5 3/4.3.8 DELETED .................................................. B 3/4 3-7 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION .......................................... B 3/4 3-7 Figure B3/4 3-1 Reactor Vessel Water Level ............. B 3/4 3-8 3/4.3.10 MECHANICAL VACUUM PUMP TRIP INSTRUMENTATION .............. B 3/4 3-9 3/4.3.11 OSCILLATION POWER RANGE MONITOR .......................... B 3/4 3-13 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM ..................................... B 3/4 4-1 3/4.4.2 SAFETY/RELIEF VALVES ..................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ................................ B 3/4 4-3 Operational Leakage ...................................... B 3/4 4-3 3/4.4.4 CHEMISTRY ................................................ B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY ........................................ B 3/4 4-4 3/4.4.6 PRESSURE/TEMPERATURE LIMITS .............................. B 3/4 4-5 Table B3/4.4.6-1 Reactor Vessel Toughness .............. B 3/4 4-7 Figure B3/4.4.6-1 Fast Neutron Fluence (E>1Mev) at (1/4)T as a Function of Service life .............. B 3/4 4-8 Table B3/4.4.6-2 Numeric Values for Pressure/Temperature Limits ........... B 3/4 4-9 HOPE CREEK xviii Amendment No.164

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY....................................... 6-1 6.2 ORGANIZATION......................................... 6-1 6.2.1 ONSITE AND OFFSITE ORGANIZATIONS................ 6-1 6.2.2 UNIT STAFF...................................... 6-1 Figure 6.2.1-1 (Deleted)............................ 6-3 Figure 6.2.2-1 (Deleted)............................ 6-4 Table 6.2.2-1 Minimum Shift Crew Composition Single Unit Facility.................. 6-5 6.2.3 SHIFT TECHNICAL ADVISOR......................... 6-6 6.3 UNIT STAFF QUALIFICATIONS............................ 6-6 6.4 TRAINING............................................. 6-6 6.5 REVIEW AND AUDIT (THIS SECTION DELETED)............. 6-6 6.6 REPORTABLE EVENT ACTION.............................. 6-14 6.7 SAFETY LIMIT VIOLATION............................... 6-14 6.8 PROCEDURES AND PROGRAMS.............................. 6-15 6.9 REPORTING REQUIREMENTS............................... 6-17 6.9.1 ROUTINE REPORTS.................................... 6-17 STARTUP REPORT..................................... 6-17 ANNUAL REPORTS..................................... 6-17 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT. 6-18 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT......... 6-19 MONTHLY OPERATING REPORTS.......................... 6-20 CORE OPERATING LIMITS REPORT....................... 6-20 6.9.2 SPECIAL REPORTS.................................... 6-21 PRESSURE AND TEMPERATURE LIMITS ..................... 6-20 HOPE CREEK xxiv Amendment No. 161

DEFINITIONS OPERABLE - OPERABILITY 1.28 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function{s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function{s) .

OPERATIONAL CONDITION - CONDITION 1.29 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any ope inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1.30 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.31 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable faUlt in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.32 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All primary oontainment penetrations required to be closed during acoident conditions are either:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deaotivated automatie valve secured in its closed position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. Each primary containment air look is in compliance with the requirements of Specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

f. The sealing mechanism associated with each primary oontainment penetration; e.g., welds, bellows or O-rings, is OPERABLE.

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.30-1 The PTLR is the specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current vessel fluence period. The pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.10.

HOPE CREEK 1-5 Amendment No. 171

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM limits specified in the PTLR LIMITING CONDITION FORQPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (hydrostatic or leak testing), and Figure 3.4.6.1-2 (heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS), and Figure 3.4.6.1-3 (operations with a critical core other than low power PHYSICS TESTS), with:

rate within limits specified in the PTLR

a. A maximum heatup of 100°F in anyone hour period, rate within limits specified in the PTLR

b. A maximum cooldown of 100°F in anyone hour period, within limits specified in the PTLR

c. A maximum temperature change of less than or equal to 20°F in anyone hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and

d. The reactor vessel flange and head flange metal temperature shall be maintained greater than or equal to 79°F when reactor vessel head bolting studs are under tension.

within limits specified in the PTLR APPLICABILITY At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURYEIL~NCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figures 3.4.6.1-1,3.4.6.1-2, and 3.4.6.1-3 as applicable, in accordance with the Surveillance Frequency Control Program.

limits specified in the PTLR HOPE CREEK 3/44-21 Amendment No. 187

REACTOR COOLANT SYSTEM limits specified in the PTLR SURVEILLANCE REQUIREMENTS (continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-3 within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and in accordance with the Surveillance Frequency Control Program during system heatup.

4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update the curves of Figures 3.4.6.1-1,3.4.6.1-2, and 3.4.6.1-3.

specified in the PTLR.

4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to the limit specified in 3.4.6.1.d.

a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:

1. s 110°F, in accordance with the Surveillance Frequency Control Program.

2. s gO°F, in accordance with the Surveillance Frequency Control Program.

b. Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during tenSioning of the reactor vessel head bolting studs.

HOPE CREEK 3/44-22 Amendment No. 187

Figure 3.4.6.1-1 DELETED Figure 3.4.6.1-1 Hydrostatic Pressure and Leak Tests PressurelTemperature Limits - Curve A 1,200

~ . -: '.-.~

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- Upper Vessel l-I o I o 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (RF)

All system leakage and hydrostatic pressure tests performed during the service life of the pressure boundary in compliance withASME Code Section XI.

This figure is valid for 32 EFPY of operation.

HOPE CREEK 3/4 4-23 Amendment No. 157

Figure 3.4.6.1-2 DELETED Figure 3.4.6.1 2 00 NonooNuclear Heatup and Cooldown PressureITemperature Limits - Curve B 1,200 1,100 * **

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r-o o 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

All heatup and cooldowns that are perfor.med when the reactor is not critical at the nor.mal heatup and cooldown rate.

This figure is valid for 32 EFPY of operation.

HOPE CREEK 3/4 4-23a Amendment No. 157

Figure 3.4.6.1-3 DELETED Figure 3.4.6.1-3 Core Critical Heatup and Cooldown Pressurerremperature limits - Curve C 1,200 1,100 1,000

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~

100 BS'F o I o 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

All heatup and cool downs that are performed when the reactor is critical at the normal heatup and cooldown rate.

This figure is valid for 32 EFPY of operation.

HOPE CREEK 3/41 4-23b Amendment No. 157

ADMINISTRATIVE CONTROLS 6.9.1.8 Deleted CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the PSEG Nuclear LLC generated CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following Technical Specifications:

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 3/4.. 2.3 MINIMUM CRITICAL POWER RATIO 3/4.2.4 LINEAR HEAT GENERATION RATE 3/4.3.11 OSCILLATION POWER RANGE MONITOR (OPRM)

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC as applicable in the following documents:

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II)"

2. CENPD-397-P-A, "Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology" .

3. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, August 1996 The CORE OPERATING LIMITS REPORT will contain the complete identification for

each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number title, revision, date, and any supplements)

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transiont and acci.dent analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto. shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.10 See Insert HOPE CREEK 6-20 Amendment No. 161 LR-N17-0032 LAR H17-02 Insert (TS 6.9.1.10 page 6-20):

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Limiting Condition for Operation Section 3.4.6, "RCS Pressure/Temperature Limits"

2. Surveillance Requirement Section 4.4.6, "RCS Pressure/Temperature Limits"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. BWROG-TP-11-022-A (SIR-05-044), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 1, dated August 2013.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.

1

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, within the time period specified for each report.

6.9.3 DELETED 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety.

c. All REPORTABLE EVENTS submitted to the Commission.

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

e. Records of changes made to the procedures required by Specification 6.8.1.

f. Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

HOPE CREEK 6-21 Amendment No. 193

LR-N17-0032 LAR H17-02 Attachment 3 Mark-up of Proposed Technical Specification Bases Pages The following Technical Specifications Bases pages for Renewed Facility Operating License NPF-57 are affected by this change request:

Technical Specification Bases Page 3/4.4.6, Pressure/Temperature Limits B 3/4 4-5 3/4.4.6, Pressure/Temperature Limits B 3/4 4-6 3/4.4.6, Pressure/Temperature Limits B 3/4 4-7 3/4.4.6, Pressure/Temperature Limits B 3/4 4-8 3/4.4.6, Pressure/Temperature Limits B 3/4 4-9 3/4.4.6, Pressure/Temperature Limits B 3/4 4-10

REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section (3.9) of the UFSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. Specifically the average rate of change of reactor coolant temperature during normal heatup and cooldown shall not exceed lOOor during any I-hour period. specified in the PTLR The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 are derived from the fracture toughness requirements of 10 eFR 50 Appendix G and ASME Code Section XI, Appendix G and ASME Code Cases N-S8B and N-640. The curves are based on the RT~ and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in UFSAR Chapter 5, Paragraph 5.3.1.5, ItFracture Toughness." Tabulated values for the P-T curves are shown in Table B 3/4.4.6-2.

specified in the PTLR specified in the PTLR The reactor vessel materials have been tested to determine their initial RT~. The results of some of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RT~. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommendations ..of .Regulatory .Guide 1. 99, Rev *. 2, ., the PTLR "Radiation Embrittlement of Reactor Vessel Material". The pressurel temperature limit curves, Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3, includes an assumed shift in RT~ for the end of life fluence.

specified in the PTLR The fluence in Bases Figure B 3/4.4.6-1 was determined using methodology described in NRC-approved General Electric Nuclear Energy Licensing Topical Report NEDC-32983P-A. This methodology is consistent with the guidance in Regulatory Guide 1.190, Rev. 0, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." the NRC-approved The actual shift in RTN~ of the vessel material will be established RAMA Fluence periodically during operation by removing and evaluating, irradiated flux Methodology wires installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the flux wires and vessel inside radius are essentially identical, the irradiated flux wires can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3',4.6.1-3 shall be adjusted, as required, on the basis of the flux wire data and recommendations of Regulatory Guide 1.99, Rev. 2.

specified in the PTLR HOPE CREEK B 3/4 4-5 Amendment No. 157

REACTOR/COOLANT SYSTEM BASES PRESSUE/TEMPERATURE LIMITS (Continued) the PTLR The pressure-temperature limit lines shown in Figures 3.4.6.1-1 and 3.4.6.1-3, curves for inservice leak and hydrostatic testing and reactor criticality have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are provided in UFSAR Section 5.3 and Appendix SA.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in. case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 DELETED 3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

HOPE CREEK B3/4 4-6 Amendment No. 186

Delete page BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS nEAT/SLAB : PREDICTED EOL BELTLINE WELD SEAM 1.0. OR HIGHEST 6.RTHDT + UPPER SHELF MAX. EOL COMPONENT OR MAT' L TYPE HEAT/LOT ~ !:UJ.ll !ffiror..Ln MARGIN ( 0 F) iFT-LBS) !IT.NDT.L.!1.

Plate 8A-533 GR B CL.l SK3025-1 .15 0.71 +19 ' 56 66 75 Weld Vert. seams for 053040/ 0.OB1 0.611 -30 : 18 118 48 shells 4&5 1125-02205 NOTE:

These values are given only for the benefit of calculat~ng the end-of-life (EOL) RT HM

HEATiSLAB HIGHEST REFERENCE NON-BELTLINE MT'L TYPE OR OR TEMpERATURE COMPONENT WELD SEAM 1.0. HEAT/LOT S1)lDT.....r.:u Shell Ring Connected to SA 533, GR.a, Cl.l All Heats +19 Vessel Flange Sot tom Head Dome SA 533, GR.8, C1.1' All ijeats +30 Bottom Head Torus SA 533, GR.B, C1.l All Heats +30 LPCI Nozzles (1) SA SOB, Cl. 2, All ~eats -20 Top Head Torus SA 533, GR.B, Cl.l All Heats +19 Top Head Flange SA SOB, C1.2 All Heats +10 Vessel Flange SA 508, C1.2 All Heats +10 Feedwater Nozzle SA SOB, C1.2 All Heats -20 Weld Metal All RPV Wel.ds All Heats 0 Closure Studs SA 540, GR.a, 24 All Heats Meet 45 ft-lbs & 25 mils lateral expansion at +10 0 F (1) The design of the Hope Creek vessel results in these nozzles experiencing a predicted EOL fluence at 1/4T of the vessel thickness of 3.3 x 10 11 n/cm2

Therefore, these nozzles are predicted to have an EOL RTnDT of +29 F.

0

!lOPE CREEK B 3/4 4-7 Amendment No.157

Delete page 8

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o 5 10 15 20 25 30 35 40 Service LIfe, Years w LOWER - INTERMEDIATE SHELL FAST NEUTRON FLUENCE (E>l MeV)

AT 1/4 T AS A FUNCTION OF SERVICE LIFE*

Bases Figure B 314.4.6-1

At BO% capacity factor (40 years" 32 EFPY)

HOPE CREEl< B 3/4 4-8 Amendment No.157

Delete page BASES TABLE B 3/4.4.6-2 Numerio Values for Pressure/Temperature Limits Figure 3.4.6.1-1, Curve A Bottom Head Upper Vessel Beltline'

remperature Pressure ll'eIlIperature Pressure ll'empera1:W::1il l'.:essure

(°F) (pdQ') (°F) (psig) (01') (psiq) 79 0 79.0 0 79.0 0 79 929 19.0 292 19.0 691 as 1040 118.0 292 ea.o 143 90 1068 118.0 925 93.0 777 92 1097 123.0 996 98.0 814 94 1126 128.0 1074 103.0 BSS 96 1157 133.0 1161 108.0 900, 98 1190 138.0 1257 113.0 950 100 1223 118.0 1,005 123.0 1,065 128.0 1,133 133.0 1,207 Figure 3.4.6.1-2, Curve B Bottom Head Upper Vessel Beltline

'1'emperature Pressure '1'emperature Pressure ~emperature l'ressure (OF) (pdQ') (°F) (pdq) (°F) (pdQ')

79 0 79.0 0 79.0 0 79 606 79.0 50 19.0 416 88 690 79.0 75 88.0 <155 92 732 79.0 90 93.0 480 96 778 79.0 100 98.0 508 100 827 79.0 125 103.0 538 104 881 79.8 175 108.0 572 108 939 86.6 202 113.0 610 112 1002 90.6 220 118.0 651 116 1070 96.6 250 123.0 697 120 1144 98.4 260 128.0 747 124 1224 102. Ei . 285 133.0 803 103.7 292 138.0 864 148.0 292 143.0 932 148.0 140 148.0 1,008 148.0 745 153.0 1,091 148.0 150 158.0 1,183 151.6 830 163.0 1,284 155.S 910 159.7 990 163.3 1070 165.5 1150 167.5 1230

HOPE CREEK B 3/4 4-9 J\mendment No. 157

Delete page BASES TABLE B 3/4.4.6-2 (continued)

Numeric Values for Pressure/Temperature Limits Figure 3.4.6.1-3, Curve C

!remperat:ure Pressure (OF) (psig) 8e.0 0 88.0 50 88.0 75 B8.0 90 92.0 100 103.4 125 119.8 175 126.6 202 130.6 220 136.6 250 138.4 260 142.6 285 143.7 292 188.0 292 188.0 740 188.0 745 188.0 750 191.6 830 195.8 910 199.7 990 203.3 1070 205.5 1150 207.5 1230 HOPE CREEK B 3/4 4-10 Amendment No. 157

LR-N17-0032 LAR H17-02 Attachment 4 Hope Creek Generating Station Pressure and Temperature Limits Report (PTLR) for 32, 44, and 56 Effective Full-Power Years (EFPY)

(Non-Proprietary)

PSEG Nuclear LLC Hope Creek Generating Station Pressure and Temperature Limits Report (PTLR) for 32, 44, and 56 Effective Full-Power Years (EFPY)

Revision 0-NP Prepared by: / '1-SJ Donnamarie Bush

[Hope Creek Engineering Programs]

Reviewed b y;__._A L=t£ Date: ;?--;6-17--

Anthony Tramontana

[Manager, Engineering Programs]

Approved by: l Date:

Mitch Dior (Acting)

[Director, Engineering]

Hope Creek Generating Station PTLR Revision 0-NP Page 2 of 61 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 HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 32 EFPY 18 Figure 2 HCGS P-T Curve B (Normal Operation - Core Not Critical) for 32 EFPY 19 Figure 3 HCGS P-T Curve C (Normal Operation - Core Critical) for 32 EFPY 20 Figure 4 HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 44 EFPY 21 Figure 5 HCGS P-T Curve B (Normal Operation - Core Not Critical) for 44 EFPY 22 Figure 6 HCGS P-T Curve C (Normal Operation - Core Critical) for 44 EFPY 23 Figure 7 HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 56 EFPY 24 Figure 8 HCGS P-T Curve B (Normal Operation - Core Not Critical) for 56 EFPY 25 Figure 9 HCGS P-T Curve C (Normal Operation - Core Critical) for 56 EFPY 26 Figure 10 HCGS Feedwater Nozzle 2-D Finite Element Model for Thermal Stress [24] 27 Figure 11 HCGS Feedwater Nozzle 3-D Finite Element Model for Pressure Stress [26] 28 Figure 12 HCGS Instrument Nozzle Finite Element Model [19] 29 Table 1 HCGS Pressure Test (Curve A) P-T Curves for 32 EFPY 30 Table 2 HCGS Core Not Critical (Curve B) P-T Curves for 32 EFPY 33 Table 3 HCGS Core Critical (Curve C) P-T Curves for 32 EFPY 36 Table 4 HCGS Pressure Test (Curve A) P-T Curves for 44 EFPY 39 Table 5 HCGS Core Not Critical (Curve B) P-T Curves for 44 EFPY 42 Table 6 HCGS Core Critical (Curve C) P-T Curves for 44 EFPY 45 Table 7 HCGS Pressure Test (Curve A) P-T Curves for 56 EFPY 48 Table 8 HCGS Core Not Critical (Curve B) P-T Curves for 56 EFPY 51 Table 9 HCGS Core Critical (Curve C) P-T Curves for 56 EFPY 54 Table 10 HCGS ART Calculations for 32 EFPY 57 Table 11 HCGS ART Calculations for 44 EFPY 58 Table 12 HCGS ART Calculations for 56 EFPY 59 Table 13 Nozzle Stress Intensity Factors 60 Appendix A Hope Creek Reactor Vessel Materials Surveillance Program 61

Hope Creek Generating Station PTLR Revision 0-NP Page 3 of 61 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, Cool-down and Hydrostatic/Class 1 Leak Testing;

2. RCS Heat-up and Cool-down rates;

3. RPV head flange boltup temperature limits.

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

2.0 Applicability This report is applicable to the HCGS RPV for up to 32, 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

Hope Creek Generating Station PTLR Revision 0-NP Page 4 of 61 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], using the RAMA computer code, 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].

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

Revision 0: Initial issue of PTLR.

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 [8], 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 new surveillance capsule data of the RPV or other plant design assumption modifications in the UFSAR, cannot

Hope Creek Generating Station PTLR Revision 0-NP Page 5 of 61 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 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 32, 44, and 56 EFPY for HCGS, as documented in Reference [7]. The HCGS P-T curves for 32 EFPY are provided in Figures 1 through 3, for 44 EFPY are in Figures 4 through 6, and for 56 EFPY are in Figures 7 through 9. A tabulation of the curves is included in Tables 1 through 3 for 32 EFPY, Tables 4 through 6 for 44 EFPY, and Tables 7 through 9 for 56 EFPY. The adjusted reference temperature (ART) tables for the HCGS vessel beltline materials are shown in Table 10 for 32 EFPY, Table 11 for 44 EFPY, and Table 12 for 56 EFPY [6].

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

Heat-up/Cool-down rate limit during Hydrostatic Class 1 Leak Testing (Figures 1, 4, and 7: Curve A): 25F/hour1 [7].

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

Hope Creek Generating Station PTLR Revision 0-NP Page 6 of 61 Normal Operating Heat-up/Cool-down rate limit (Figures 2, 5, and 8: Curve B - non-nuclear heating, and Figures 3, 6, and 9: Curve C - nuclear heating): 100F/hour2 [7].

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

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

RPV flange and adjacent shell temperature limit 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 [9] (approved for use by the NRC in Reference [10]) and bounds the minimum temperature in the first set of P-T limits approved for initial operation [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 for the representative plate and weld materials. 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 2

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

Hope Creek Generating Station PTLR Revision 0-NP Page 7 of 61 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 Position 2.1 in RG 1.99

[5]. Use of surveillance data from the BWRVIP ISP for HCGS was approved by the NRC in Reference [12].

The peak RPV ID fluence values of 8.81 x 1017 n/cm2 at 32 EFPY, 1.26 x 1018 n/cm2 at 44 EFPY, and 1.63 x 1018 n/cm2 at 56 EFPY used in the P-T curve evaluation were obtained from Reference [6] (interpolated from the fluence values at 24.1 and 56 EFPY in Reference [4]).

Fluence values in Reference [4] were calculated in accordance with RG 1.190 [3]. These fluence values apply to the limiting beltline lower intermediate shell plates (heat nos. 5K2963/1, 5K2530/1, and 5K3238/1). A plant-specific damage assessment, in terms of displacements per atom (dpa), was performed to determine through-wall fluence for HCGS in Reference [4], as permitted by RG 1.99 [5]. The resulting attenuation factor is 0.70 for a postulated 1/4T flaw.

Consequently, the 1/4T fluence for 32, 44, and 56 EFPY for the limiting lower intermediate shell plates are 6.13 x 1017 n/cm2, 8.76 x 1017 n/cm2, and 1.14 x 1018 n/cm2, respectively, for HCGS.

The limiting value for ART is 90.3°F at 32 EFPY, 103.3°F at 44 EFPY, and 113.8°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 [13]. The feedwater (FW) nozzle is considered in the evaluation of the non-beltline (upper vessel) region P-T limits.

The limiting LPCI (N17) nozzles have an RPV ID fluence of 1.69 x 1017 n/cm2 at 56 EFPY, obtained from Reference [4] and calculated in accordance with RG 1.190 [3]. Similar to the RPV beltline plates and welds described above, through-wall fluence for the LPCI nozzles was

Hope Creek Generating Station PTLR Revision 0-NP Page 8 of 61 attenuated using the dpa methodology in Reference [4]. The resulting attenuation factor is 0.82 for a postulated 1/4T flaw in the LPCI nozzle blend radius. Consequently, the 1/4T fluence for 56 EFPY for the limiting LPCI nozzle location is 1.39 x 1017 n/cm2. The limiting value for ART is 8.7°F at 56 EFPY [6]. Comparing the limiting ART value of the LPCI nozzles with the initial RTNDT for non-beltline nozzles, 40°F, the non-beltline nozzles are more limiting. This assumption is further supported when comparing nozzle transients, where the FW nozzle, outside the beltline, experiences more severe thermal transients. Therefore, the P-T curves developed for the FW nozzle bound the LPCI nozzles. Due to this bounding assumption, LPCI nozzle evaluations should use the upper vessel P-T curve 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 1.89 x 1017 n/cm2 at 32 EFPY, 2.61 x 1017 n/cm2 at 44 EFPY, and 3.34 x 1017 n/cm2 at 56 EFPY [6]. This fluence value applies 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 dpa methodology in Reference [4]. The resulting attenuation factor is 0.84 for a postulated 1/4T flaw in the N16 nozzle corner. Consequently, the 1/4T fluence for 32, 44, and 56 EFPY for the limiting N16 nozzle location is 1.58 x 1017 n/cm2, 2.20 x 1017 n/cm2, and 2.81 x 1018 n/cm2, respectively, for HCGS. The limiting value for ART of the N16 nozzles is 52.5°F at 32 EFPY, 60.1°F at 44 EFPY, and 66.7°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 cool-down and is in the outer wall during heat-up. However, as a

Hope Creek Generating Station PTLR Revision 0-NP Page 9 of 61 conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heat-up and cool-down. 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 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 toughness at a given pressure would exceed the allowable 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 cool-down temperature rate of 100°F/hour 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 heat-up and cool-down temperature rate of 25°F/hour 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/cool-down 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 >

1MeV) are shown in Tables 10, 11, and 12 for 32, 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 [14].

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 Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [15, 16].

Hope Creek Generating Station PTLR Revision 0-NP Page 10 of 61 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) [16], 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 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 [16], 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 plate heat 5K3238/1 and weld heat D53040.

The only computer code used in the determination of the HCGS P-T curves was the ANSYS finite element computer program:

ANSYS, Release 8.1 (Service Pack 1) [17] for:

o FW nozzle (non-beltline) through-wall thermal stress distributions in Reference

[18].

o Instrument nozzle thermal and pressure stress distributions in Reference [19].

ANSYS Mechanical APDL and PrepPost, Release 11.0 (Service Pack 1) [20] for the FW nozzle (non-beltline) pressure stress distribution in Reference [21].

ANSYS finite element analyses were used to develop the stress distributions through the FW and instrument nozzles, and these stress distributions were used in the determination of the stress intensity factors for these nozzles [18, 19, 21]. At the time that each of the analyses above was performed, the ANSYS program was controlled under the vendors 10 CFR 50 Appendix B [22]

Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 [23] was performed as a part of the computer program

Hope Creek Generating Station PTLR Revision 0-NP Page 11 of 61 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 References [18] and [21]. The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle:

A two-dimensional axisymmetric finite element model of the FW nozzle was constructed for the determination of thermal stresses (Figure 10). Details of the model and material properties are provided in Reference [24]. Temperature-dependent material properties were based on the ASME Code,Section II, Part D, 2001 Edition through 2003 Addenda

[25].

Heat transfer coefficients were calculated in Reference [24] and 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 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.

A three-dimensional finite element model of the FW nozzle was constructed (Figure 11) for the determination of pressure stresses. Details of the model and material properties are provided in Reference [26]. Material properties were based on the ASME Code,Section II, Part D, 2001 Edition through 2003 Addenda [25]. The properties were taken at a

Hope Creek Generating Station PTLR Revision 0-NP Page 12 of 61 temperature of 350°F, which is the approximate average service temperature of the FW nozzle at HCGS [21]. The use of temperature-independent material properties is consistent with design basis documents. Use of temperature-dependent material properties is expected to have minimal impact on the results of the analysis.

With respect to pressure stress, a unit pressure of 1000 psig was applied to the internal surfaces of the 3-D model in Reference [21]. 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 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 [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 12. Temperature-dependent material properties, taken from the ASME Code,Section II, Part D, 2001 Edition with 2003 Addenda edition [25], were used in the evaluation and are described in Reference [27].

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 through the nozzle corner is used. The BIE/IF methodology presented in Reference [1] was used to calculate the thermal stress intensity

Hope Creek Generating Station PTLR Revision 0-NP Page 13 of 61 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 1000 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

Hope Creek Generating Station PTLR Revision 0-NP Page 14 of 61 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 Report No. EPR-HC1-001-R-002, Revision 0, Hope Creek Nuclear Generating Station Unit 1 Reactor Pressure Vessel Fluence Evaluation at End of Cycle 19 with Projections to 56 EFPY, March 30, 2016, SI File No. 1600507.201.

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

6. Structural Integrity Associates Calculation No. 1601009.301, Revision 0, Hope Creek RPV Beltline ART Evaluation, February 9, 2017.

7. Structural Integrity Associates Calculation No. 1601009.303, Revision 0, Hope Creek Updated P-T Curve Calculation for 32, 44, and 56 EFPY, February 9, 2017.

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

9. Attachment 3 to PSEG Letter No. LR-N04-0136, dated March 31, 2004, SI Report No.

SIR-00-136, Revision 1, Revised Pressure-Temperature (P-T) Curves for Hope Creek, March 23, 2004. (ADAMS Accession No. ML040990522)

Hope Creek Generating Station PTLR Revision 0-NP Page 15 of 61

10. Hope Creek License Amendment No. 157, Change of Pressure-Temperature Limits and Extension of Validity to 32 Effective Full-Power Years, dated November 1, 2004. (TAC No. MC2534, ADAMS Accession No. ML042050079)

11. PSEG Design Information Transmittal No. H-TODI-2016-0066, dated November 14, 2016. SI File No. 1601009.201.

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

14. Hope Creek Generating Station, Issuance of Amendment 88 (TAC No. M93054),

November 28, 1995.

15. BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. EPRI PROPRIETARY INFORMATION.

16. EPRI Letter 2017-017, Final BWRVIP Integrated Surveillance Program (ISP) Data Applicable to Hope Creek Generating Station, January 19, 2017. EPRI PROPRIETARY INFORMATION.

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

18. PSEG Document No. VTD 327370, Structural Integrity Associates Calculation No. HC-05Q-304, Revision 2, Feedwater Nozzle (N4A through N4F) Thermal and Stress Analysis, October 22, 2009.

19. PSEG Document No. VTD 431277, Structural Integrity Associates Calculation No.

0800118.311, Revision 1, Stress Analysis of Reactor Pressure Vessel 2" Instrument Nozzles N16A through N16D, October 29, 2009.

Hope Creek Generating Station PTLR Revision 0-NP Page 16 of 61

20. ANSYS Mechanical and PrepPost, Release 11.0 (w/ Service Pack 1), ANSYS, Inc.,

August 2007.

21. PSEG Document No. VTD 432637, Structural Integrity Associates Calculation No.

1200619.305, Revision 0, Greens Function Analysis of Feedwater Nozzle, March 19, 2015.

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

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

24. PSEG Document No. VTD 327367, Structural Integrity Associates Calculation No. HC-05Q-301, Revision 3, Feedwater Nozzle (N4A through N4F) Finite Element Model and Heat Transfer Coefficients, June 4, 2010.

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

26. PSEG Document No. VTD 431269, Structural Integrity Associates Calculation No.

0800118.303, Revision 1, Feedwater Nozzle (N4A through N4F) Finite Element Model, October 19, 2009.

27. PSEG Document No. VTD 431267, Structural Integrity Associates Calculation No.

0800118.301, Revision 0, Material Properties and Methodology of Heat Transfer Coefficient Calculation for Hope Creek Environmental Fatigue Evaluation, May 30, 2008.

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

29. GE Nuclear Energy Report No. GE-NE-523-A164-1294R1, Hope Creek Generating Station RPV Surveillance Materials Testing and Fracture Toughness Analysis, December 1997. SI File No. PSEG-10Q-201P. GE PROPRIETARY INFORMATION.

Hope Creek Generating Station PTLR Revision 0-NP Page 17 of 61

30. BWRVIP- 298NP: BWR Vessel and Internals Project: Testing and Evaluation of the Hope Creek 120° Capsule, EPRI, Palo Alto, CA: 2014. 3002007844. SI File No.

1601009.201.

31. 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.

Hope Creek Generating Station PTLR Revision 0-NP Page 1 8 of 61 Figure 1: HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 32 EFPY Curve A - Pressure Test, Composite Curves

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Hope Creek Generating Station PTLR Revision 0-NP Page 1 9 of 61 Figure 2: HCGS P-T Curve B (Normal Operation- Core Not Critical) for 32 EFPY Curve B-:- Core Not Critical, Composite Curves

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Temperature= 79°F 0

! i l 0 so 100 150 200 250 Minimum Reactor Vessel Metal Temperature ("F)

Hope Creek Generating Station PTLR Revision 0-NP Page 20 of 61 Figure 3: HCGS P-T Curve C (Normal Operation- Core Critical) for 32 EFPY Curve C- Core Critical, Composite Curves

-Beltline ---Bottom Head - Non-Beltline -overall

'M 5 800 +---*--:--+-:- ------ -+---

]

>ilJ 700 Ill

!}.

600 +-+--7--+-+--r-+--+-+-

e ilJ QJ 500 +--+-+--+-+-+-+--+--+-+--F-4--+-H ll.

Temperature = 79°F 0 50 100 150 200 250 Minimum Reactor Vessel Metal Temperature (°F)

Hope Creek Generating Station PTLR Revision 0-NP Page 21 of 61 Figure 4: HCGS P-T Curve A {Hydrostatic Pressure and Leak Tests) for 44 EFPY Curve A .. Pressure Test, Composite Curves

- Beltline ---Bottom Head - Non-Beltline -overall 900

.!:; 800 Ill 0

700 -

u A!

=

... 600

e Q)

J 500 Q) c.

0 50 100 150 200 250 Minimum Reactor Vesse.l Metal Temperature ("F)

Hope Creek Generating Station PTLR Revision 0-NP Page 22 of 61 Figure 5: HCGS P-T Curve B (Normal Operation- Core Not Critical) for 44 EFPY Curve B- Core Not Critical, Composite Curves

-Beltline ---Bottom Head - Non-Beltline -overall 1300 1200 1100 1000 900

'tin

v;

.3: 800 Qi Ill Ill Q) 0 700 t:

<a

<11 cr::

.5

..... 600

e Q)

I

"' 500 Q) c.

400 300 200 100 Minimum Bolt-Up Temperature = 79°F 0

0 50 100 150 200 250 Minimum Reactor Vessel Metal Temperature ("F)

Hope Creek Generating Station PTLR Revision 0-NP Page 23 of 61 Figure 6: HCGS P-T Curve C (Normal Operation- Core Critical) for 44 EFPY Curve c .. Core Critical, Composite Curves

-Beltline .. --Bottom Head - Non-Beltline -overall 1300 1200 1100 1000 900 lin

'iii

.e 800 ru II) n:l 700

.5

.... 600

e II)

I ClJ 500

400 300 200 100 Temperature = 79°F 0

0 50 100 150 200 250 Minimum Reactor Vessel Metal Temperature (°F)

Hope Creek Generating Station PTLR Revision 0-NP Page 24 of 61 Figure 7: HCGS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 56 EFPY Curve A - Pressure Test, Composite Curves

-Beltline --- Bottom Head - Non-Beltline '"'-overall 5 800 +-------- --L- --+-:.

700 +----c.-----

600 +--+-+--0--

'e CIJ CIJ soo t---r-i--:-4--'--,1L+--;--:--+-r-----;--+--r--1 ts:.

Minimum Bolt-Up Temperature = 79°F 0 50 100 150 200 250 Minimum Reactor Vessel Metal Temperature (°F)

Hope Creek Generating Station PTLR Revision 0-NP Page 25 .of 61 Figure 8: HCGS P-TCurve B (Normal Operation- Core Not Critical) for 56 EFPY Curve B .. Core Not Critical, Composite Curves

-Beltline - --Bottom Head - Non-Beltline -overall

.e 800 +--:-:*-0-'---+-,__ _.;:,.__; __,;_+--

(1J 0 700 +---<--*-.;__....,c..-+-

600 t--t--t--r--t-!-i--T-1--i,_+-+-r-E m

(1J 500 +--r--r--+--1-----+---+-*

lr, 0 50 100 150 200 250 Mlnimum Reactor Vesse.l Metal Temperature ("F)

Hope Creek Generating Station PTLR Revision 0-NP Page 26 of 61 Figure 9: HCGS P-T Curve C (Normal Operation- Core Critical) for 56 EFPY Curve C - Core Critical, Composite Curves

-Beltline ---Bottom Head - Non-Beltline -overall 0 50 100 150 200 250 Minimum Reactor Vessel Metal Temperature ("F)

Hope Creek Generating Station PTLR Revision 0-NP Page 27 of 61 Figure 10: HCGS Feedwater Nozzle 2D Finite Element Model for Thermal Stress [24]

"1 ll.LE,l!!DNT J\1\1'>0

rAii . 9 2oo<i 0@,1AJ;*.t9

Hope Creek Generating Station PTLR Revision 0-NP Page 28 of 61 Figure 11: HCGS Feedwater Nozzle 3-D Finite Element Model for Pressure Stress [26]

HOPE er Nozzle Finite Element Model

Hope Creek Generating Station PTLR Revision 0-NP Page 29 of 61 Figure 12: HCGS Instrument Nozzle Finite Element Model [19]

.1\N ELEMENTS JAN 20 2009 MAT NUM 09:14:47 2 inch Instrtunent Nozzle ELEMENTS JAN 20 2009 MAT NUM 09:19:35

Hope Creek Generating Station PTLR Revision 0-NP Page 30 of 61 Table 1: HCGS Pressure Test (Curve A) P-T Curves for 32 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 497.8 87.2 543.7 94.2 589.7 100.4 635.6 105.9 681.5 110.8 727.4 115.3 773.4 119.4 819.3 123.2 865.2 126.8 911.1 130.1 957.1 134.2 1003.7 138.0 1050.3 141.6 1096.9 144.9 1143.5 148.0 1190.1 150.9 1236.7 153.7 1283.3 156.3 1329.9

Hope Creek Generating Station PTLR Revision 0-NP Page 31 of 61 Table 1: HCGS Pressure Test (Curve A) P-T Curves for 32 EFPY (continued)

Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 312.6 109.0 312.6 109.0 815.1 113.1 864.9 116.9 914.8 120.4 964.7 123.7 1014.5 126.8 1064.4 129.7 1114.2 132.4 1164.1 135.0 1214.0 137.5 1263.8 139.8 1313.7

Hope Creek Generating Station PTLR Revision 0-NP Page 32 of 61 Table 1: HCGS Pressure Test (Curve A) P-T Curves for 32 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 693.7 84.3 742.0 89.1 790.3 93.4 838.6 97.4 886.9 101.1 935.2 104.6 983.5 107.8 1031.8 110.9 1080.1 113.7 1128.4 116.4 1176.6 119.0 1224.9 121.4 1273.2 123.8 1321.5

Hope Creek Generating Station PTLR Revision 0-NP Page 33 of 61 Table 2: HCGS Core Not Critical (Curve B) P-T Curves for 32 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 197.4 90.2 245.9 99.3 294.4 107.0 342.8 113.7 391.3 119.6 439.8 124.8 488.2 129.6 536.7 134.0 585.2 138.0 633.6 141.7 682.1 145.1 730.6 148.3 779.0 151.4 827.5 155.2 876.6 158.7 925.6 162.0 974.6 165.1 1023.6 168.0 1072.7 170.8 1121.7 173.4 1170.7 175.9 1219.8 178.2 1268.8 180.5 1317.8

Hope Creek Generating Station PTLR Revision 0-NP Page 34 of 61 Table 2: HCGS Core Not Critical (Curve B) P-T Curves for 32 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 133.7 87.5 178.4 94.8 223.2 101.1 267.9 106.8 312.6 139.0 312.6 139.0 692.3 141.9 740.7 144.7 789.0 147.3 837.4 149.8 885.8 152.2 934.2 154.5 982.5 156.7 1030.9 158.7 1079.3 160.7 1127.6 162.7 1176.0 164.5 1224.4 166.3 1272.8 168.0 1321.1

Hope Creek Generating Station PTLR Revision 0-NP Page 35 of 61 Table 2: HCGS Core Not Critical (Curve B) P-T Curves for 32 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 445.5 86.0 494.1 92.1 542.7 97.5 591.3 102.4 639.9 106.9 688.4 111.0 737.0 114.8 785.6 118.3 834.2 121.6 882.8 124.7 931.4 127.6 980.0 130.3 1028.6 133.0 1077.1 135.4 1125.7 137.8 1174.3 140.0 1222.9 142.2 1271.5 144.3 1320.1

Hope Creek Generating Station PTLR Revision 0-NP Page 36 of 61 Table 3: HCGS Core Critical (Curve C) P-T Curves for 32 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 90.7 101.4 139.9 116.8 189.0 128.5 238.1 138.0 287.2 146.0 336.3 152.9 385.4 159.0 434.6 164.4 483.7 169.2 532.8 173.7 581.9 177.8 631.0 181.5 680.2 185.0 729.3 188.3 778.4 191.4 827.5 195.2 876.6 198.7 925.6 202.0 974.6 205.1 1023.6 208.0 1072.7 210.8 1121.7 213.4 1170.7 215.9 1219.8 218.2 1268.8 220.5 1317.8

Hope Creek Generating Station PTLR Revision 0-NP Page 37 of 61 Table 3: HCGS Core Critical (Curve C) P-T Curves for 32 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 1.1 96.2 45.6 109.0 90.1 119.2 134.6 127.6 179.1 134.9 223.6 141.2 268.1 146.8 312.6 179.0 312.6 179.0 692.3 181.9 740.7 184.7 789.0 187.3 837.4 189.8 885.8 192.2 934.2 194.5 982.5 196.7 1030.9 198.7 1079.3 200.7 1127.6 202.7 1176.0 204.5 1224.4 206.3 1272.8 208.0 1321.1

Hope Creek Generating Station PTLR Revision 0-NP Page 38 of 61 Table 3: HCGS Core Critical (Curve C) P-T Curves for 32 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 266.4 93.7 316.3 105.0 366.2 114.2 416.0 122.0 465.9 128.8 515.8 134.7 565.6 140.0 615.5 144.8 665.4 149.2 715.2 153.2 765.1 156.9 815.0 160.4 864.8 163.7 914.7 166.7 964.6 169.6 1014.4 172.3 1064.3 174.9 1114.2 177.3 1164.1 179.6 1213.9 181.9 1263.8 184.0 1313.7

Hope Creek Generating Station PTLR Revision 0-NP Page 39 of 61 Table 4: HCGS Pressure Test (Curve A) P-T Curves for 44 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 461.4 88.5 507.7 96.4 554.1 103.3 600.4 109.4 646.7 114.7 693.1 119.6 739.4 124.0 785.7 130.2 834.3 135.7 882.9 140.7 931.5 145.2 980.0 149.3 1028.6 153.1 1077.2 156.7 1125.8 160.0 1174.4 163.1 1222.9 166.0 1271.5 168.8 1320.1

Hope Creek Generating Station PTLR Revision 0-NP Page 40 of 61 Table 4: HCGS Pressure Test (Curve A) P-T Curves for 44 EFPY (continued)

Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 312.6 109.0 312.6 109.0 815.1 113.1 864.9 116.9 914.8 120.4 964.7 123.7 1014.5 126.8 1064.4 129.7 1114.2 132.4 1164.1 135.0 1214.0 137.5 1263.8 139.8 1313.7

Hope Creek Generating Station PTLR Revision 0-NP Page 41 of 61 Table 4: HCGS Pressure Test (Curve A) P-T Curves for 44 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 693.7 84.3 742.0 89.1 790.3 93.4 838.6 97.4 886.9 101.1 935.2 104.6 983.5 107.8 1031.8 110.9 1080.1 113.7 1128.4 116.4 1176.6 119.0 1224.9 121.4 1273.2 123.8 1321.5

Hope Creek Generating Station PTLR Revision 0-NP Page 42 of 61 Table 5: HCGS Core Not Critical (Curve B) P-T Curves for 44 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 170.1 91.2 216.2 101.0 262.2 109.2 308.3 116.3 354.4 122.4 400.4 127.9 446.5 132.9 492.6 137.4 538.7 141.5 584.7 145.3 630.8 150.8 679.9 155.7 728.9 160.2 778.0 164.3 827.0 168.1 876.1 171.7 925.2 175.0 974.2 178.1 1023.3 181.0 1072.4 183.7 1121.4 186.4 1170.5 188.8 1219.6 191.2 1268.6 193.5 1317.7

Hope Creek Generating Station PTLR Revision 0-NP Page 43 of 61 Table 5: HCGS Core Not Critical (Curve B) P-T Curves for 44 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 133.7 87.5 178.4 94.8 223.2 101.1 267.9 106.8 312.6 139.0 312.6 139.0 692.3 141.9 740.7 144.7 789.0 147.3 837.4 149.8 885.8 152.2 934.2 154.5 982.5 156.7 1030.9 158.7 1079.3 160.7 1127.6 162.7 1176.0 164.5 1224.4 166.3 1272.8 168.0 1321.1

Hope Creek Generating Station PTLR Revision 0-NP Page 44 of 61 Table 5: HCGS Core Not Critical (Curve B) P-T Curves for 44 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 445.5 86.0 494.1 92.1 542.7 97.5 591.3 102.4 639.9 106.9 688.4 111.0 737.0 114.8 785.6 118.3 834.2 121.6 882.8 124.7 931.4 127.6 980.0 130.3 1028.6 133.0 1077.1 135.4 1125.7 137.8 1174.3 140.0 1222.9 142.2 1271.5 144.3 1320.1

Hope Creek Generating Station PTLR Revision 0-NP Page 45 of 61 Table 6: HCGS Core Critical (Curve C) P-T Curves for 44 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 78.5 103.0 124.5 119.1 170.5 131.3 216.5 141.1 262.6 149.3 308.6 156.3 354.6 162.5 400.7 168.0 446.7 172.9 492.7 177.4 538.7 181.5 584.8 185.3 630.8 190.8 679.9 195.7 728.9 200.2 778.0 204.3 827.0 208.1 876.1 211.7 925.2 215.0 974.2 218.1 1023.3 221.0 1072.4 223.7 1121.4 226.4 1170.5 228.8 1219.6 231.2 1268.6 233.5 1317.7

Hope Creek Generating Station PTLR Revision 0-NP Page 46 of 61 Table 6: HCGS Core Critical (Curve C) P-T Curves for 44 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 1.1 96.2 45.6 109.0 90.1 119.2 134.6 127.6 179.1 134.9 223.6 141.2 268.1 146.8 312.6 179.0 312.6 179.0 692.3 181.9 740.7 184.7 789.0 187.3 837.4 189.8 885.8 192.2 934.2 194.5 982.5 196.7 1030.9 198.7 1079.3 200.7 1127.6 202.7 1176.0 204.5 1224.4 206.3 1272.8 208.0 1321.1

Hope Creek Generating Station PTLR Revision 0-NP Page 47 of 61 Table 6: HCGS Core Critical (Curve C) P-T Curves for 44 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 266.4 93.7 316.3 105.0 366.2 114.2 416.0 122.0 465.9 128.8 515.8 134.7 565.6 140.0 615.5 144.8 665.4 149.2 715.2 153.2 765.1 156.9 815.0 160.4 864.8 163.7 914.7 166.7 964.6 169.6 1014.4 172.3 1064.3 174.9 1114.2 177.3 1164.1 179.6 1213.9 181.9 1263.8 184.0 1313.7

Hope Creek Generating Station PTLR Revision 0-NP Page 48 of 61 Table 7: HCGS Pressure Test (Curve A) P-T Curves for 56 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 434.0 89.7 480.2 98.5 526.5 105.9 572.8 112.4 619.0 118.2 665.3 123.3 711.6 130.8 758.9 137.2 806.2 143.0 853.5 148.1 900.8 152.7 948.1 157.0 995.4 160.9 1042.7 164.6 1090.0 168.0 1137.3 171.1 1184.6 174.1 1231.9 176.9 1279.2 179.6 1326.5

Hope Creek Generating Station PTLR Revision 0-NP Page 49 of 61 Table 7: HCGS Pressure Test (Curve A) P-T Curves for 56 EFPY (continued)

Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 312.6 109.0 312.6 109.0 815.1 113.1 864.9 116.9 914.8 120.4 964.7 123.7 1014.5 126.8 1064.4 129.7 1114.2 132.4 1164.1 135.0 1214.0 137.5 1263.8 139.8 1313.7

Hope Creek Generating Station PTLR Revision 0-NP Page 50 of 61 Table 7: HCGS Pressure Test (Curve A) P-T Curves for 56 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 693.7 84.3 742.0 89.1 790.3 93.4 838.6 97.4 886.9 101.1 935.2 104.6 983.5 107.8 1031.8 110.9 1080.1 113.7 1128.4 116.4 1176.6 119.0 1224.9 121.4 1273.2 123.8 1321.5

Hope Creek Generating Station PTLR Revision 0-NP Page 51 of 61 Table 8: HCGS Core Not Critical (Curve B) P-T Curves for 56 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 149.5 93.6 199.0 104.9 248.6 114.1 298.1 121.9 347.6 128.6 397.1 134.5 446.6 139.8 496.1 144.6 545.6 151.3 594.1 157.2 642.5 162.5 691.0 167.2 739.4 171.6 787.9 175.6 836.3 179.3 884.8 182.7 933.2 186.0 981.7 189.0 1030.1 191.8 1078.5 194.6 1127.0 197.1 1175.4 199.6 1223.9 201.9 1272.3 204.1 1320.8

Hope Creek Generating Station PTLR Revision 0-NP Page 52 of 61 Table 8: HCGS Core Not Critical (Curve B) P-T Curves for 56 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 133.7 87.5 178.4 94.8 223.2 101.1 267.9 106.8 312.6 139.0 312.6 139.0 692.3 141.9 740.7 144.7 789.0 147.3 837.4 149.8 885.8 152.2 934.2 154.5 982.5 156.7 1030.9 158.7 1079.3 160.7 1127.6 162.7 1176.0 164.5 1224.4 166.3 1272.8 168.0 1321.1

Hope Creek Generating Station PTLR Revision 0-NP Page 53 of 61 Table 8: HCGS Core Not Critical (Curve B) P-T Curves for 56 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 445.5 86.0 494.1 92.1 542.7 97.5 591.3 102.4 639.9 106.9 688.4 111.0 737.0 114.8 785.6 118.3 834.2 121.6 882.8 124.7 931.4 127.6 980.0 130.3 1028.6 133.0 1077.1 135.4 1125.7 137.8 1174.3 140.0 1222.9 142.2 1271.5 144.3 1320.1

Hope Creek Generating Station PTLR Revision 0-NP Page 54 of 61 Table 9: HCGS Core Critical (Curve C) P-T Curves for 56 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 69.2 106.3 116.9 123.9 164.5 136.9 212.1 147.2 259.8 155.7 307.4 163.0 355.1 169.3 402.7 175.0 450.4 180.0 498.0 184.6 545.6 191.3 594.1 197.2 642.5 202.5 691.0 207.2 739.4 211.6 787.9 215.6 836.3 219.3 884.8 222.7 933.2 226.0 981.7 229.0 1030.1 231.8 1078.5 234.6 1127.0 237.1 1175.4 239.6 1223.9 241.9 1272.3 244.1 1320.8

Hope Creek Generating Station PTLR Revision 0-NP Page 55 of 61 Table 9: HCGS Core Critical (Curve C) P-T Curves for 56 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 1.1 96.2 45.6 109.0 90.1 119.2 134.6 127.6 179.1 134.9 223.6 141.2 268.1 146.8 312.6 179.0 312.6 179.0 692.3 181.9 740.7 184.7 789.0 187.3 837.4 189.8 885.8 192.2 934.2 194.5 982.5 196.7 1030.9 198.7 1079.3 200.7 1127.6 202.7 1176.0 204.5 1224.4 206.3 1272.8 208.0 1321.1

Hope Creek Generating Station PTLR Revision 0-NP Page 56 of 61 Table 9: HCGS Core Critical (Curve C) P-T Curves for 56 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

°F psi 79.0 0.0 79.0 266.4 93.7 316.3 105.0 366.2 114.2 416.0 122.0 465.9 128.8 515.8 134.7 565.6 140.0 615.5 144.8 665.4 149.2 715.2 153.2 765.1 156.9 815.0 160.4 864.8 163.7 914.7 166.7 964.6 169.6 1014.4 172.3 1064.3 174.9 1114.2 177.3 1164.1 179.6 1213.9 181.9 1263.8 184.0 1313.7

Hope Creek Generating Station PTLR Revision 0-NP Page 57 of 61 Table 10: HCGS ART Calculations for 32 EFPY Chemistry Adjustments For 1/4T Initial Chemistry Fluence Fluence Fluence Flux (wt%)

Description Heat/Lot No. RTNDT Factor, at ID at 1/4T Factor, RTNDT Margin Terms ART Type

(°F) Cu Ni CF (°F) (n/cm2) (n/cm2) FF (°F) (°F)

(°F) i (°F)

Intermediate Shell (3) 5K3025/1 - 19 0.15 0.71 112.8 3.14E+17 2.22E+17 0.184 20.7 10.4 0.0 60.4 Intermediate Shell (3) 5K2608/1 - 19 0.09 0.58 58.0 3.14E+17 2.22E+17 0.184 10.7 5.3 0.0 40.3 Intermediate Shell (3) 5K2698/1 - 19 0.10 0.58 65.0 3.14E+17 2.22E+17 0.184 11.9 6.0 0.0 42.9 Lower Intermediate Shell (4) 5K2963/1 - -10 0.07 0.58 44.0 8.81E+17 6.13E+17 0.326 14.3 7.2 0.0 18.7 Plates Lower Intermediate Shell (4) 5K2530/1 - 19 0.08 0.56 51.0 8.81E+17 6.13E+17 0.326 16.6 8.3 0.0 52.3 Lower Intermediate Shell (4) 5K3238/1 - 7 0.09 0.64 58.0 8.81E+17 6.13E+17 0.326 18.9 9.5 0.0 44.8 Lower Shell (5) 5K3230/1 - -10 0.07 0.56 44.0 6.05E+17 4.22E+17 0.267 11.7 5.9 0.0 13.5 Lower Shell (5) 6C35/1 - -11 0.09 0.54 58.0 6.05E+17 4.22E+17 0.267 15.5 7.7 0.0 19.9 Lower Shell (5) 6C45/1 - 1 0.08 0.57 51.0 6.05E+17 4.22E+17 0.267 13.6 6.8 0.0 28.2 Vertical W13 510-01205 SMAW -40 0.09 0.54 108.7 3.10E+17 2.19E+17 0.182 19.8 9.9 0.0 -0.5 Vertical W13 D53040/1125-02205 SAW -30 0.08 0.63 110.1 3.10E+17 2.19E+17 0.182 20.0 10.0 0.0 10.1 Vertical W14 510-01205 SMAW -40 0.09 0.54 108.7 7.72E+17 5.40E+17 0.305 33.2 16.6 0.0 26.3 Vertical W14 D53040/1125-02205 SAW -30 0.08 0.63 110.1 7.72E+17 5.40E+17 0.305 33.6 16.8 0.0 37.2 Vertical W15 510-01205 SMAW -40 0.09 0.54 108.7 5.27E+17 3.67E+17 0.247 26.8 13.4 0.0 13.7 Vertical W15 D53040/1125-02205 SAW -30 0.08 0.63 110.1 5.27E+17 3.67E+17 0.247 27.2 13.6 0.0 24.4 Welds Girth W6 (Shell 3-4) 519-01205 SMAW -49 0.01 0.53 20.0 3.14E+17 2.22E+17 0.184 3.7 1.8 0.0 -41.7 Girth W6 (Shell 3-4) 504-01205 SMAW -31 0.01 0.51 20.0 3.14E+17 2.22E+17 0.184 3.7 1.8 0.0 -23.7 Girth W6 (Shell 3-4) 510-01205 SMAW -40 0.09 0.54 108.7 3.14E+17 2.22E+17 0.184 20.0 10.0 0.0 -0.1 Girth W6 (Shell 3-4) D53040/1810-02205 SAW -49 0.08 0.63 110.1 3.14E+17 2.22E+17 0.184 20.2 10.1 0.0 -8.5 Girth W6 (Shell 3-4) D55733/1810-02205 SAW -40 0.10 0.68 126.4 3.14E+17 2.22E+17 0.184 23.2 11.6 0.0 6.4 Girth W7 (Shell 4-5) 510-01205 SMAW -40 0.09 0.54 108.7 6.05E+17 4.22E+17 0.267 29.0 14.5 0.0 18.0 Girth W7 (Shell 4-5) D53040/1125-02205 SAW -30 0.08 0.63 110.1 6.05E+17 4.22E+17 0.267 29.4 14.7 0.0 28.8 LPCI (N17; A-D) 19468/1 - -20 0.12 0.80 86.0 9.75E+16 8.02E+16 0.094 8.1 4.0 0.0 -3.8 Nozzles LPCI (N17; A-D) 10024/1 - -20 0.14 0.82 105.1 9.75E+16 8.02E+16 0.094 9.9 4.9 0.0 -0.2 Instrument (N16; A, D) 5K3025/1 (adj. plate) - 19 0.15 0.71 112.8 1.89E+17 1.58E+17 0.149 16.7 8.4 0.0 52.5 Instrument (N16; B, C) 5K2698/1 (adj. plate) - 19 0.10 0.58 65.0 1.89E+17 1.58E+17 0.149 9.7 4.8 0.0 38.3 LPCI Nozzle W179 001-01205 SMAW -40 0.02 0.51 27.0 2.76E+17 2.08E+17 0.176 4.8 2.4 0.0 -30.5 Nozzle LPCI Nozzle W179 519-01205 SMAW -49 0.01 0.53 20.0 2.76E+17 2.08E+17 0.176 3.5 1.8 0.0 -42.0 Welds LPCI Nozzle W179 504-01205 SMAW -31 0.01 0.51 20.0 2.76E+17 2.08E+17 0.176 3.5 1.8 0.0 -24.0 Surveillance Plate 5K3238/1 - 7 0.09 0.64 (( 8.81E+17 6.13E+17 0.326 16.0 8.0 0.0 39.1 ISP Surveillance Weld D53040 SAW -30 0.07 0.57 (( }} 7.72E+17 5.40E+17 0.305 64.3 28.0 0.0 90.3 REDACTED EPRI PROPRIETARY INFORMATION [16] (such information is marked with double braces (( }} and a bar in the right-hand margin)

Hope Creek Generating Station PTLR Revision 0-NP Page 58 of 61 Table 11: HCGS ART Calculations for 44 EFPY Chemistry Adjustments For 1/4T Initial Chemistry Fluence Fluence Fluence Flux (wt%) Description Heat/Lot No. RTNDT Factor, at ID at 1/4T Factor, RTNDT Margin Terms ART Type (°F) Cu Ni CF (°F) (n/cm2) (n/cm2) FF (°F) (°F) (°F) i (°F) Intermediate Shell (3) 5K3025/1 - 19 0.15 0.71 112.8 4.35E+17 3.08E+17 0.223 25.2 12.6 0.0 69.3 Intermediate Shell (3) 5K2608/1 - 19 0.09 0.58 58.0 4.35E+17 3.08E+17 0.223 12.9 6.5 0.0 44.9 Intermediate Shell (3) 5K2698/1 - 19 0.10 0.58 65.0 4.35E+17 3.08E+17 0.223 14.5 7.3 0.0 48.0 Lower Intermediate Shell (4) 5K2963/1 - -10 0.07 0.58 44.0 1.26E+18 8.76E+17 0.391 17.2 8.6 0.0 24.4 Plates Lower Intermediate Shell (4) 5K2530/1 - 19 0.08 0.56 51.0 1.26E+18 8.76E+17 0.391 19.9 10.0 0.0 58.9 Lower Intermediate Shell (4) 5K3238/1 - 7 0.09 0.64 58.0 1.26E+18 8.76E+17 0.391 22.7 11.3 0.0 52.4 Lower Shell (5) 5K3230/1 - -10 0.07 0.56 44.0 8.78E+17 6.12E+17 0.326 14.3 7.2 0.0 18.7 Lower Shell (5) 6C35/1 - -11 0.09 0.54 58.0 8.78E+17 6.12E+17 0.326 18.9 9.5 0.0 26.8 Lower Shell (5) 6C45/1 - 1 0.08 0.57 51.0 8.78E+17 6.12E+17 0.326 16.6 8.3 0.0 34.2 Vertical W13 510-01205 SMAW -40 0.09 0.54 108.7 4.30E+17 3.04E+17 0.221 24.1 12.0 0.0 8.1 Vertical W13 D53040/1125-02205 SAW -30 0.08 0.63 110.1 4.30E+17 3.04E+17 0.221 24.4 12.2 0.0 18.7 Vertical W14 510-01205 SMAW -40 0.09 0.54 108.7 1.10E+18 7.70E+17 0.367 39.9 19.9 0.0 39.7 Vertical W14 D53040/1125-02205 SAW -30 0.08 0.63 110.1 1.10E+18 7.70E+17 0.367 40.4 20.2 0.0 50.8 Vertical W15 510-01205 SMAW -40 0.09 0.54 108.7 7.63E+17 5.32E+17 0.303 32.9 16.5 0.0 25.8 Vertical W15 D53040/1125-02205 SAW -30 0.08 0.63 110.1 7.63E+17 5.32E+17 0.303 33.3 16.7 0.0 36.7 Welds Girth W6 (Shell 3-4) 519-01205 SMAW -49 0.01 0.53 20.0 4.35E+17 3.08E+17 0.223 4.5 2.2 0.0 -40.1 Girth W6 (Shell 3-4) 504-01205 SMAW -31 0.01 0.51 20.0 4.35E+17 3.08E+17 0.223 4.5 2.2 0.0 -22.1 Girth W6 (Shell 3-4) 510-01205 SMAW -40 0.09 0.54 108.7 4.35E+17 3.08E+17 0.223 24.3 12.1 0.0 8.5 Girth W6 (Shell 3-4) D53040/1810-02205 SAW -49 0.08 0.63 110.1 4.35E+17 3.08E+17 0.223 24.6 12.3 0.0 0.1 Girth W6 (Shell 3-4) D55733/1810-02205 SAW -40 0.10 0.68 126.4 4.35E+17 3.08E+17 0.223 28.2 14.1 0.0 16.4 Girth W7 (Shell 4-5) 510-01205 SMAW -40 0.09 0.54 108.7 8.78E+17 6.12E+17 0.326 35.4 17.7 0.0 30.9 Girth W7 (Shell 4-5) D53040/1125-02205 SAW -30 0.08 0.63 110.1 8.78E+17 6.12E+17 0.326 35.9 17.9 0.0 41.8 LPCI (N17; A-D) 19468/1 - -20 0.12 0.80 86.0 1.33E+17 1.10E+17 0.117 10.0 5.0 0.0 0.1 Nozzles LPCI (N17; A-D) 10024/1 - -20 0.14 0.82 105.1 1.33E+17 1.10E+17 0.117 12.3 6.1 0.0 4.5 Instrument (N16; A, D) 5K3025/1 (adj. plate) - 19 0.15 0.71 112.8 2.61E+17 2.20E+17 0.182 20.6 10.3 0.0 60.1 Instrument (N16; B, C) 5K2698/1 (adj. plate) - 19 0.10 0.58 65.0 2.61E+17 2.20E+17 0.182 11.8 5.9 0.0 42.7 LPCI Nozzle W179 001-01205 SMAW -40 0.02 0.51 27.0 3.79E+17 2.85E+17 0.213 5.8 2.9 0.0 -28.5 Nozzle LPCI Nozzle W179 519-01205 SMAW -49 0.01 0.53 20.0 3.79E+17 2.85E+17 0.213 4.3 2.1 0.0 -40.5 Welds LPCI Nozzle W179 504-01205 SMAW -31 0.01 0.51 20.0 3.79E+17 2.85E+17 0.213 4.3 2.1 0.0 -22.5 Surveillance Plate 5K3238/1 - 7 0.09 0.64 (( }} 1.26E+18 8.76E+17 0.391 19.2 8.5 0.0 43.2 ISP Surveillance Weld D53040 SAW -30 0.07 0.57 (( }} 1.10E+18 7.70E+17 0.367 77.3 28.0 0.0 103.3 REDACTED EPRI PROPRIETARY INFORMATION [16] (such information is marked with double braces (( }} and a bar in the right-hand margin)

Hope Creek Generating Station PTLR Revision 0-NP Page 59 of 61 Table 12: HCGS ART Calculations for 56 EFPY Chemistry Adjustments For 1/4T Initial Chemistry Fluence Fluence Fluence Flux (wt%) Description Heat/Lot No. RTNDT Factor, at ID at 1/4T Factor, RTNDT Margin Terms ART Type (°F) Cu Ni CF (°F) (n/cm2) (n/cm2) FF (°F) (°F) (°F) i (°F) Intermediate Shell (3) 5K3025/1 - 19 0.15 0.71 112.8 5.55E+17 3.94E+17 0.257 28.9 14.5 0.0 76.9 Intermediate Shell (3) 5K2608/1 - 19 0.09 0.58 58.0 5.55E+17 3.94E+17 0.257 14.9 7.4 0.0 48.8 Intermediate Shell (3) 5K2698/1 - 19 0.10 0.58 65.0 5.55E+17 3.94E+17 0.257 16.7 8.3 0.0 52.4 Lower Intermediate Shell (4) 5K2963/1 - -10 0.07 0.58 44.0 1.63E+18 1.14E+18 0.444 19.5 9.8 0.0 29.0 Plates Lower Intermediate Shell (4) 5K2530/1 - 19 0.08 0.56 51.0 1.63E+18 1.14E+18 0.444 22.6 11.3 0.0 64.2 Lower Intermediate Shell (4) 5K3238/1 - 7 0.09 0.64 58.0 1.63E+18 1.14E+18 0.444 25.7 12.9 0.0 58.5 Lower Shell (5) 5K3230/1 - -10 0.07 0.56 44.0 1.15E+18 8.02E+17 0.374 16.5 8.2 0.0 22.9 Lower Shell (5) 6C35/1 - -11 0.09 0.54 58.0 1.15E+18 8.02E+17 0.374 21.7 10.9 0.0 32.4 Lower Shell (5) 6C45/1 - 1 0.08 0.57 51.0 1.15E+18 8.02E+17 0.374 19.1 9.5 0.0 39.2 Vertical W13 510-01205 SMAW -40 0.09 0.54 108.7 5.50E+17 3.89E+17 0.255 27.7 13.9 0.0 15.4 Vertical W13 D53040/1125-02205 SAW -30 0.08 0.63 110.1 5.50E+17 3.89E+17 0.255 28.1 14.0 0.0 26.1 Vertical W14 510-01205 SMAW -40 0.09 0.54 108.7 1.43E+18 1.00E+18 0.417 45.3 22.7 0.0 50.6 Vertical W14 D53040/1125-02205 SAW -30 0.08 0.63 110.1 1.43E+18 1.00E+18 0.417 45.9 23.0 0.0 61.8 Vertical W15 510-01205 SMAW -40 0.09 0.54 108.7 1.00E+18 6.97E+17 0.349 37.9 18.9 0.0 35.8 Vertical W15 D53040/1125-02205 SAW -30 0.08 0.63 110.1 1.00E+18 6.97E+17 0.349 38.4 19.2 0.0 46.8 Welds Girth W6 (Shell 3-4) 519-01205 SMAW -49 0.01 0.53 20.0 5.55E+17 3.94E+17 0.257 5.1 2.6 0.0 -38.7 Girth W6 (Shell 3-4) 504-01205 SMAW -31 0.01 0.51 20.0 5.55E+17 3.94E+17 0.257 5.1 2.6 0.0 -20.7 Girth W6 (Shell 3-4) 510-01205 SMAW -40 0.09 0.54 108.7 5.55E+17 3.94E+17 0.257 27.9 14.0 0.0 15.8 Girth W6 (Shell 3-4) D53040/1810-02205 SAW -49 0.08 0.63 110.1 5.55E+17 3.94E+17 0.257 28.3 14.1 0.0 7.5 Girth W6 (Shell 3-4) D55733/1810-02205 SAW -40 0.10 0.68 126.4 5.55E+17 3.94E+17 0.257 32.5 16.2 0.0 24.9 Girth W7 (Shell 4-5) 510-01205 SMAW -40 0.09 0.54 108.7 1.15E+18 8.02E+17 0.374 40.7 20.3 0.0 41.3 Girth W7 (Shell 4-5) D53040/1125-02205 SAW -30 0.08 0.63 110.1 1.15E+18 8.02E+17 0.374 41.2 20.6 0.0 52.4 LPCI (N17; A-D) 19468/1 - -20 0.12 0.80 86.0 1.69E+17 1.39E+17 0.137 11.7 5.9 0.0 3.5 Nozzles LPCI (N17; A-D) 10024/1 - -20 0.14 0.82 105.1 1.69E+17 1.39E+17 0.137 14.3 7.2 0.0 8.7 Instrument (N16; A, D) 5K3025/1 (adj. plate) - 19 0.15 0.71 112.8 3.34E+17 2.81E+17 0.211 23.8 11.9 0.0 66.7 Instrument (N16; B, C) 5K2698/1 (adj. plate) - 19 0.10 0.58 65.0 3.34E+17 2.81E+17 0.211 13.7 6.9 0.0 46.5 LPCI Nozzle W179 001-01205 SMAW -40 0.02 0.51 27.0 4.81E+17 3.62E+17 0.245 6.6 3.3 0.0 -26.8 Nozzle LPCI Nozzle W179 519-01205 SMAW -49 0.01 0.53 20.0 4.81E+17 3.62E+17 0.245 4.9 2.4 0.0 -39.2 Welds LPCI Nozzle W179 504-01205 SMAW -31 0.01 0.51 20.0 4.81E+17 3.62E+17 0.245 4.9 2.4 0.0 -21.2 Surveillance Plate 5K3238/1 - 7 0.09 0.64 (( }} 1.63E+18 1.14E+18 0.444 21.8 8.5 0.0 45.8 ISP Surveillance Weld D53040 SAW -30 0.07 0.57 (( }} 1.43E+18 1.00E+18 0.417 87.8 28.0 0.0 113.8 REDACTED EPRI PROPRIETARY INFORMATION [16] (such information is marked with double braces (( }} and a bar in the right-hand margin)

Hope Creek Generating Station PTLR Revision 0-NP Page 60 of 61 Table 13: Nozzle Stress Intensity Factors Nozzle Applied Pressure, KIp-app Thermal, KIt Feedwater 78.46 41.85 Instrument (N16) 75.94 24.75 0.5 KI in units of ksi-in

Hope Creek Generating Station PTLR Revision 0-NP Page 61 of 61 Appendix A HOPE CREEK REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [28], two surveillance capsules have been removed from the HCGS reactor vessel in 1994 after 6.01 EFPY [29] and in 2015 after 24.1 EFPY [30]. The 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 materials within the core beltline region. HCGS is currently committed to use the BWRVIP ISP, and has made a licensing commitment to use the ISP for HCGS 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 NRC. HCGS committed to use the ISP in place of its existing surveillance programs in the license amendment issued by the NRC regarding the implementation of the BWRVIP ISP, dated July 23, 2004 [12]. Under the ISP, a capsule was removed in 2015 after 24.1 EFPY [30]. HCGS continues to be a host plant under the ISP. One additional standby HCGS capsule is currently scheduled to be removed and tested under the ISP during the license renewal period in approximately 2036 at 40 EFPY [31].

LR-N17-0032 LAR H17-02 Attachment 5 Affidavit for Withholding signed by Electric Power Research Institute (EPRI)

-=*-*-. ELECTRIC POWeR RSEARCI{ INSTITUTE NElLWILMSHUnS.T Vice President and Chief Nvdeor Offh:er RefEPRI Pro]eot Numqer 669 February 16,2017 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nucle ar Regulatory Co.mmlsslon Washington, DC 205550001

Subject:

RequesHorWithholding ofthe following Proprleary Information Included in:

   't!crertse Amendmeht Request to Amend the Hope Creek Technical Specifications (TS) t9 R1;3Vise and Relocate the PressureTemperature Lim!t Curves to a Pressure and Temperature Limits Repottj' dated February 2017 TpWhom It May Concern:

This is a request uhder 10 G.F.R. §2.390(a)(4) that the U.s. Nuc!ear Regulatory Commission ("NRC") withhold from public disclosure the report Identified In the enclosed Affidavit consisting of the proprietary Information owned by Electric Power Research Institute Inc. ("EPRI") identified in the attached report. Proprietary and non propdetary versions of the Report and the Affidavit in support of this request ate enclosed. EPRl desires to disclose the Proprietary Information in confidence to assist the. NRC review of the enclosed submittal to the NRC by PSEG. The Proprietary Information is not to be divulged to anyone outside ()f the NRC or to any of its contractors, norshall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the Information enclosed. lfyou have any que$Uons aboutthe legal aspects ofthisrequest for withholding,please do not hesitate to contact me at (704) 5$52732. Questions on the contentofthe Report should be direCted to Andy MgOehee ()f EPRI at (704) 502-6440. Attachment(s) c: Sheldon Stuchell, NRC. (shelclon.stuchell@nro,gov) Together .. Shaping the Future of Electricity PALO ALTO OFFIE

 $420 HilivleW Avenv, Pe1!o Alto, CA 94304*1395 USA

650.855.20QO
Colomor Servlt 1300.313.3774
www.apfl.<:om

1 -11;8 meTRic POWER RmARCH IN$TITUTf AFFIOA\llT Re R,equst fQr WithhQlding of th Followin Proprietary lnformationJnctuded In:

      License AtnendmentRequest to Amend tbe Hope Crk Tephnicl epegiflcaUons (T$) to Revtse and Relocate the PressureTemperature Utnl't Curves to a Pressure nd T(l}mp:rature Umits.Report"
                                                         *dated. February 2017 t,Ne!IW!Imshurst1 be.ing duly sworn,depose and state a$ follows:

I am the 'Vic Pre!)ic;lentand Clh!f Nuclear Officer at Eleotrlc Power Research Jnsiitute,. hie; Whose principal office is .located at .1360 W WTHartls Blvd, Charlotte, NQ, eEPRIP)and l*have'beenPeQifloaUy delegatti!d responlb.IJity for: the abPYe*Ustd. report that contalns EPRI Proprietary' Information that is sought under this Affid.aVit t<Lba Withheld P:rop(iet9lty !hformatlor". 1 .?J.m vlnorl.ed to apply to thf;l IJ,,a, NuolearHegulatory Qomrntssitm (NRCn) for tha wlthholdln,g of'theProprletatY*hiformation otY behlfol EPRJ.. I:PRI Proprietary lnfothmUon I$ Identified inthe abqve ref\treMI:lQ rf;!port bY double braces. J:;xarnples of uoh lc!.entiftcalionJs. as t(>!lows: ((This sehtehce is aQ ex.ample,;{J}} Table .cont..lning PRL Proprietary Information are*ldentified .with .double bra.ckets before: and atterth object. lneaah case me sUpersoript:notationlftt.re,fars:to thi§. i:!ffldavttand allthe bas.es.JncJU'ded below; Which provide the r(;}_asolls.Jorthe proprietary determination. EP.RI requ$$ts that:the Proprl.etacy.lofot:tmlti onbe withbeld;f.roorthe publi'c:on thefoilowingb.ases: Withholding Based Upon f>rlvi!aged

                                   .            ..          AM *confld(inllal Trade .secrets .or        Commercial Or FinanCial lnformat!on{see e;g:. 10C.F.R.§;2.39Q(a)(4t:               *           .     ,,     * *         *
              . ..    . .. . a.  . The Propr!etaJy lnfotl)1.?tJoh ls l)WMd: 'py f;PRI nd h .been Mld ln ponfidence by f2P:Rl. All (:!!ltitl.es. &¢epting cQples oftor; Proprietary lnfonuation do $o subjaorto 'written ag*r(;}rnent IIJ1p¢s!n.g an obl ig ation lipo.n the reclpl.ent to mG1intain the>cgnfidentlali!toffhe .Proprietary lnformtron. The Proprietary

Information-Is disclosed onlyla parties.who agree, in wrltlng1 to preserve the confl9otiality-Jhre-of.

b, . EPI qonsiqrs the Flropritary Tnform:tion containedtherein to contiMe trade secrets* of E:PRI. As such.xEPRI hOlds the Information in donfidE!nce and disclosure th,ereofJ:WictlyJtroiteg to lnd.ivid'ual!l*. M enliUes whq h;.ve 1:1greed, in wdtlng,to malntalnthconflctenUalityoftMlnformalion.

c. The Information sought to bawrthel(l ts cons!<trdlo b pr<;mr[taryfQr,Jha folloWing r$on. EPRI mad? a s.upstniJal economlctnvstm(1nttodeveloptha. f1roprietaty Information <:iha1. yprohibltin public disc[osura,EPRI deriVes an ecOnomlc,ben.euti.nlhe, form of licensing royaiU$ ,nd Qther additional fees from the .ooofid.entlal nature, of the. Proprietary* lnfon:riation. Jf fhe Proprietary. lnfortnatloh were publi!:: IY C!YC;lJif!lbla

*to col1Sultants and/or othf'b(Jsinesses'pr9vidin.g services In fhe elc.trlc and/or O.\ICJat powerindustry, they would

be able to us(? the Propriet;ry Information for their own commercial benefit and profJfand without expehding the substantial economic resources re.qulred of EPRI to develop the Proprietary Information.

    .         . d.         EPRI's cla$slfication of the Proprietary Information as trade secrets isjustified by the

.Uniform Trade Secrets Act whlch Californi a adopted in 1984 and a version ofwhich has been adoptr:;dby over forty states. The California Uniform Trade SecretsAct, California Civil Code §§3426:342Mt defines a "trade secret" as follows:

                  .,Trade secret' rneans ihformatlon, lncludlng a formula, patter n , compilatioh, program device, method, technique, or proc(;)ss, that:

(1) Derives independent economic value actual or potential, from not being generally known to the public or to other persons who can o btai n economic value from its disclosure oruse; and (2) Is the sul)jectof effe cts that are rttsonab!e 11nder the circumstances to maintain itssecrecy."

e. The PropriE?tary Information contained therein are not generally kMWn oravallable to the public . EPRI developed the I nform at ion only after making . a determin(ltion tht the Proprietary Information was not available from public $ourcf)s, EPRI rnade a substantiallnvestmetit of bofh mtmey and employee h?urs in me development of the Proprietary Information. EPRiwasrequlred to devote these resources and. efforlto derive the Proprietary Information, As a result.of such effort and cost, both intrms of dollar$ s pent and dedicated employee time, the Proprietary Information Is highly valuable to EPRI.

f. A publ!c disClos ure of the Proprietary Information would be highly likely to cause substariUalharmto EPRI's competitive position and the ability of EPRl to licensethe Propr!etattlnfarmation both domestically and internationally.*The Propriet ary Information can only be'acquired and/or duplicated by others uslng'an equivalent inwstmemtoftime and effort.

I have .read the foregoing and 'the matters stated heteln are true and .correct to the best of my knowledge, informtion and belie(. l make this affidavit under penalty of p'erjury under the lawsofthe United States ofAtnetlca and underthe laws of the State of Californi§. Executed at 3420 Hillview Avenue, Palo Alto, CA 94304 being the premises and place of business of Electric PoWer Research Institute, Inc.

Clif9rntaAII*PUrpose* Acknowiedgeme.nt Civil bode 1189 A npiQIY publlo* orothroffl(}ercompJatfng thi certificate Verifies only the idehtityP,fth J.nctlvlt;lual wno. $lgnact the

  .documt'mt towhleh tnrs certlflqat Is attaohed, and nQt th$ trqtfuln aouraey. or*Valldlty of. thatd<rcVtiwnt.

{State ofCallfotnia) (County ofsanta ¢lata)}}

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CAC:MF9502, Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR (Approved, Closed)
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MONTHYEAR2017-04-26 [Table View]

LR-N17-0032, License Amendment Request to Amend the Technical Specifications (TS) to Revise and Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report

Text

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

Subject:

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