JAFP-08-0034, Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A

Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A

ML081200881
Person / Time
Site:	FitzPatrick
Issue date:	04/22/2008
From:	Peter Dietrich Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-08-0034
Download: ML081200881 (61)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. Fitzpatrick NPP Enterg P.O. Box 110 Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 Pete Dietrich Site Vice President - JAF April 22, 2008 JAFP-08-0034 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59 Application for Amendment to Technical Specifications Regardingq Relocation of -

Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limitsý Report (PTLR) Consistent with TSTF-419-A

References:

1. Structural Integrity Associates Topical Report (TR) SIR-05-044-A,."Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", April 2007.:.

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

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests an amendment to the Technical Specifications (TS) for the James A. FitzPatrick Nuclear Power Plant (JAF).

The proposed amendment modifies Technical Specifications (TS) Section 1.0 (."Definitions"),

Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3,4;9

("RCS Pressure and Temperature (P/T) Limits"), and Section 5.0 ("Administrative Controls").to delete reference to the pressure and temperature curves, and include reference to the Pressure-and Temperature Limits Report (PTLR). This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated April :

2007 (Reference 1), for preparation of the pressure ahd temperature curves, and incorporates the guidance of TSTF-419-A ("Revise PTLR Definition and References in ISTS 5.6.6, RCS_

PTLR") (Reference 2). The JAF PTLR has been developed based on the methodology and template provided in SIR-05-044-A, and.is enclosed as a reference for NRC review. provides a description and evaluation of the proposed TS changes. provides the proposed changes to the current TS on marked up pages. provides the proposed TS changes in final typed format. provides the proposed changes to the current TS Bases on marked up pages. provides the Pressure and Temperature Limits Report (PTLR).

/- 6o If

ý_ (Ff,

The Bases changes are provided for NRC information only. The final TS Bases pages will be submitted with a future update in accordance with TS 5.5.11, "Technical Specifications (TS)

Bases Control Program".

Entergy requests NRC approval of the proposed TS amendment by August 01, 2008, to support the upcoming Refuel Outage beginning September 14,2008, with the amendment being implemented within 30 days from approval.

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

There are no new commitments made in this letter.

Should you have any questions concerning this submittal, please contact Mr. Jim Costedio at 315-349-6358.

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

Executed on the 2 2 nd day of April 2008.

PD/ed Attachments: 1. Description and evaluation of the proposed TS changes

2. Proposed changes to the current TS on marked up pages

3. Proposed TS changes in final typed format

4. Proposed changes to the current TS Bases on marked up pages

5. Pressure and Temperature Limits Report (PTLR) cc: next page

cc:

Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Resident Inspector's Office U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. Adrian Muniz, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-G5 Washington, DC 20555 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 1 0 th Floor Albany, NY 12223 Mr. Paul Tonko, President NYSERDA 17 Columbia Circle Albany, NY 12203-6399

JAFP-08-0034 Attachment I Description and Evaluation Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419

JAFP-08-0034 Attachment 1 1.0 Description The proposed amendment modifies Technical Specifications (TS) Section 1.0

("Definitions"), Limiting Conditions for Operation and Surveillance Requirement Appplicability Section 3.4.9 ("RCS Pressure and Temperature (P/T) Limits"), and Section 5.0 ("Administrative Controls") to delete reference to the pressure and temperature curves, and include reference to the Pressure and Temperature Limits Report (PTLR).

This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A (Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR). The JAF PTLR has been developed based on the methodology and template provided in SIR-05-044-A, and is enclosed as reference for NRC review.

2.0 Proposed Changes The proposed change modifies:

1) TS Section 1.0 to add a definition of the "Pressure and Temperature Limits Report".

2) TS Limiting Conditions for Operation and Surveillance Requirement, Section 3.4.9

("RCSPressure and Temperature (P/T) Limits").

3) TS Section 5.6.7 is being added to include wording from TSTF-419-A concerning:

1) the individual TSs that address reactor coolant system P-T limits; 2) the NRC-approved topical report that docu ment PTLR methodologies; and 3) the requirements for providing a revised PTLR to the NRC.

The copies of the TS Bases pages are provided for NRC information only. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

3.0 Background

In a letter dated February 6, 2007 (Reference 1), "the NRC staff has found that SIR 044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE". This Safety Evaluation Report (SER) permits Boiling Water Reactor (BWR) licensees to relocate their pressure-temperature (P-T) curves from the facility TS to a Pressure and Temperature Limits Report (PTLR) utilizing the guidance in TS Task Force (TSTF) Traveler No. 419-A.

The Reference 1 NRC safety evaluation discussed the Boiling Water Reactor Owners' Group (BWROG) Licensing Topical Report (LTR) SIR-05-044, "Pressure Temperature Limits Report Methodology for Boiling Water Reactors", Revision 0, dated December 2005 (ADAMS Accession No. ML053560336) for review and acceptance for referencing in subsequent licensing actions. The BWROG provided this LTR to support the ability of Page 1 of 9

JAFP-08-0034 Attachment 1 Boiling Water Reactor (BWR) licensees to relocate their pressure-temperature (P-T) curves and the associated numerical values (such as heatup/cooldown rates) from the facility TS to a Pressure and Temperature Limits Report (PTLR), a licensee-controlled document, using the guidelines provided in Generic Letter (GL) 96-03, "Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits" (Reference 2). Proposed revisions to this LTR and responses to NRC staff requests for additional information (RAIs) were provided in a letter from the BWROG dated August 29, 2006 (ADAMS Accession No. ML062440387). The Reference 1 NRC safety evaluation approved the use of this report. The Structural Integrity Associates Report was issued as a final report (-A) in April 2007 (Reference 3).

TS Task Force (TSTF) Traveler No. 419 (Reference 4) amended the Standard TS (NUREGs-1430, -1431, -1432, -1433, and -1434) to: (1) delete references to the TS LCO specifications for the P-T limits in the TS definition for the PTLR, and (2) revise STS 5.6.6 to identify by number and title, NRC-approved topical reports that document PTLR methodologies, or the NRC safety evaluation for a plant-specific methodology by NRC letter and date. A requirement was added to the reviewers note to specify the complete citation of PTLR methodology in the plant specific PTLR, including the re 'ort number, title, revision, date, and any supplements. Only the figures, values, and parameters associated with the P-T limits are relocated to the PTLR. The TSTF also specified that the methodology, and any subsequent changes, must be reviewed and approved by the NRC. In this Case, the methodology was approved in the Reference 1 letter.

The JAF Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05-044-A is being submitted for review. The purpose of the JAF PTLR is to present operating limits related to Reactor Coolant System (RCS) pressure versus temperature limits during heatup, cooldown and hydrostatic/class 1 leak testing. The curves, which have been prepared using NRC approved methodology, will allow system pressurization at lower temperatures thus saving critical path time and provide improved work environment conditions for the inspectors during leak testing inspections. The pressure and temperature curves utilize the methodology of SIR 044-A.

Page 2 of 9

JAFP-08-0034 Attachment 1 4.0 Technical Analysis NRC GL 96-03 (Reference 2) allows plants to relocate their pressure-temperature (P-T) curves and numerical values of other P-T limits (such as heatup/cooldown rates) from the plant Technical Specifications to a PTLR, which is a licensee-controlled document.

As stated in GL 96-03, during the development of theimproved standard technical specifications (STS), a change was proposed to relocate the P-T limits currently contained in the plant Technical Specifications 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 Technical Specifications to a PTLR so that the licensee could maintain these limits efficiently. One of the prerequisites for having the PTLR option is that all of the. methods used to develop the P-T curves and limits be NRC approved, and that the associated Licensing Topical Report (LTR) for such approval is referenced in the plant.

Technical Specifications. Based on this prerequisite, the purpose of the Structural Integrity Associates Report is to provide BWRs with an NRC-approved LTR that can be referenced in plant Technical Specifications to establish BWR fracture mechanics methods for generating P-T curves/limits that allow BWR plants to adopt the PTLR option.

Historically, utilities that own BWRs have submitted license amendment requests to update their P-T curves. In addition, the current situation causes both the regulator and licensees to expend. resources that could otherwise be devoted to other activities. The objective of theStructural Integrity Associates Report is to avoid these situations by.

providing P-T curve methods that are generically approved by the NRC so that P-T curves can be documented in a PTLR.

In order to implement the PTLR, the analytical methods used to develop the P-T limits must be consistent with those previously reviewed and approved by the NRC and must be referenced in the Administrative Controls section of the plant Technical Specifications. The Structural Integrity Associates Report provides the current BWROG methodology for developing reactor coolant system (RCS) pressure test, core not critical, and core critical P-T curves for BWRs.:

As discussed in Section 2.1 of the Reference 1 NRC Safety Evaluation Report, 10 CFR Part 50, Appendix G, requires licensees to establish limits on the pressure and temperature of the Reactor Coolant Pressure Boundary (RCPB) in order to protect the

.RCPB against brittle failure (i.e., against brittle "fast-fracture"). These limits are defined by P-T limit curves for normal operations (including heatup and cooldown operations of the Reactor Coolant System (RCS), normal operation of the RCS with the reactor being in a critical condition, and transient operating conditions) and during pressure testing conditions (i.e., either inservice leak rate testing and/or hydrostatic testing conditions)..

As discussed in the NRC's Safety Evaluation Report (SER), which approves the BWROG LTR SIR-05-044-A (Reference 1), this LTR was prepared by Structural Integrity Associates and has three sections and two appendices. Section 1.0 describes the background and purpose for the LTR. Section 2.0 provides the fracture mechanics methodology and its basis for developing P-T limits. Section 3.0 provides a step-by-step Page 3 of 9

JAFP-08-0034 Attachment 1 procedure for calculating P-T limits. Appendix A provides guidance for evaluating surveillance data. Appendix B provides a template PTLR.

Section 2.0 provides the fracture mechanics methodology and its basis for developing P-T limits. The NRC staff evaluation of this section is based on the criteria contained in of GL 96-03. Attachment 1 of GL 96-03 contains seven technical criteria that the contents of proposed methodology should conform to if license amendments requesting PTLRs are to be approved by the NRC staff. The NRC staff's evaluations of the contents of the BWROG methodology against the seven criteria in Attachment 1 of GL 96-03 are provided in Section 3.1 of the SER.

Section 3.0 of the LTR provides a step-by-step procedure for calculating P-T limit curves. This section indicates that P-T limits may be developed for other RPV regions to provide additional operating flexibility. The JAF P-T curves consider three regions of the vessel; beltline, bottom head and non-beltline (including flange) regions.

The JAF Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05-044-A is being submitted for review. The pressure and temperature curves utilize the methodology of SIR-05-044-A. The PTLR will be an appendix to the TRM, which is a licensee-controlled document.

As noted in the NRC Safety Evaluation Report (SER) related to the license renewal of JAF (Reference 6), NRC Open Item 4.2.1 - Reactor Vessel Neutron Fluence, determined that the previous fluence calculation did not conform to Regulatory Guide (RG)1.190:. New fluence calculations were subsequently submitted on November 5, 2007 (Reference 7). The NRC staff reviewed the new calculational values and found the new fluence calculation acceptable and adhered to the guidance of RG 1.190. The NRC staff's detailed evaluation of the fluence calculation is documented in section 4.2.1.2 of the SER for JAF license renewal (Reference 6)

Page 4 of 9

JAFP-08-0034 Attachment 1

'5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change modifies Technical Specifications (TS)' Section 1.0

("Definitions"), Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 ("RCS Pressure and Temperature (P/T) Limits"), and 5.0

("Administrative Controls"), to delete reference to the pressure and temperature curves and include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A ("Revised PTLR Definition and References in ISTS 5.6.6, RCS PTLR"). In an NRC Safety Evaluation Report dated February 6, 2007, "the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE".

As part of this change, the JAF Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05-044-A is being supplied for review. The pressure and temperature curves utilize the methodology of SIR 044-A.

The NRC has established requirements in Appendix G to 10 CFR 50 in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Additionally, the regulation in 10 CFR Part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating lightwater reactors. Implementing this NRC approved methodology does not reduce the ability to protect the reactor coolant pressure boundary 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. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

Page 5 of 9

JAFP-08-0034 Attachment I The proposed change does not affect the assumed accident performance of the RCPB, nor any plant structure, system, or component previously evaluated. The proposed change does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The change in methodology ensures that the RCPB remains capable of performing its safety functions. No set points are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced which could introduce the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change does not affect the function of the RCPB or its response during plant transients. There are no changes proposed which alter the set points at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated April 2007 for preparation of the p*ressure and temperature curves. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

. ,Thischange adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A ("Revise PTLR Definition and References in ISTS 5.6.6, RCS

.PTLR"). In an NRC Safety Evaluation Report dated February 6, 2007, the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors".

Based upon the above, Entergy concludes that the proposed amendment 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 The NRC has established requirements in Appendix G of Part 50 to Title 10 of the Code of FederalRegulations, in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The regulation at 10 CFR Part 50, 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 of Appendix G to Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code, Section Xl, Appendix G) were used to generate the P-T limits. The regulation at 10 CFR Part 50, Appendix G, also requires that applicable surveillance data from reactor pressure vessel (RPV) 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.

Page 6 of 9

JAFP-08-0034 Attachment 1 Table 1 to 10 CFR Part 50, Appendix G provides the NRC staff's criteria for meeting the P-T limit requirements of ASME Code, Section Xl, Appendix G, as well as the minimum temperature requirements of the rule for bolting up the vessel during normal and pressure testing operations. In addition, the NRC staff regulatory guidance related to P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", and Standard Review Plan Chapter 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock".

The regulation at 10 CFR Part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating lightwater reactors. JAF Nuclear Power Plant demonstrates its compliance with the Appendix H through participation in the BWRVIP Integrated Surveillance Program (ISP)

(Reference 5).

In March 2001, the NRC staff issued RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence". Fluence calculations are acceptable if they are done with approved methodologies or with methods which are shown to conform to the guidance in RG 1.190 (References 6 and 7).

Section 182a of the Atomic Energy Act of 1954 requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

The regulation at 10 CFR 50.36(d)(2)(ii) requires that LCOs be established for the P-T limits, because the parameters fall within the scope of the Criterion 2 identified in the rule:

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The P-T limits for BWR-designed light-water reactors fall within the scope of Criterion 2 of 10 CFR 50.36(d)(2)(ii) and are therefore ordinarily required to be included within the TS LCOs for a plant-specific facility operating license. On January 31, 1996, the NRC staff issued GL 96-03 to inform licensees that they may request a license amendment to relocate the P-T limit curves from the TS LCOs into a PTLR or other licensee-controlled document that would be controlled through the Administrative Controls Section of the TS. In GL 96-03, the NRC staff informed licensees that in order to implement a PTLR, the P-T limits for light-water reactors would need to be generated in accordance with an NRC-approved methodology and that the methodology to generate the P-T limits would need to comply with the requirements of 10 CFR Part 50, Appendices G and H; be documented in an NRC-approved topical report or plant-specific submittal; and be incorporated by reference in the Administrative Controls Section of the TS.

Page 7 of 9

JAFP-08-0034 Attachment 1 This change implements the methodology provided in the Structural Integrity Associates report (Reference 3), which will continue to ensure compliance with Appendices G and H of the Code of Federal Regulations in conjunction with plant commitments to the BWRVIP ISP program, and the associated regulatory guidance, including TSTF-419-A, which provides TS changes.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

6.0 Environmental Assessment A review has determined that the proposed changes 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 changes do not involve: (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet 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 changes.

Page 8 of 9

JAFP-08-0034 Attachment 1 7.0 Precedent This change is generally consistent with the changes to the Improved TS described in TSTF-419-A, (Reference 4). Plants which have received approval for similar changes, in whole or in part, are listed below:

Calloway Plant Unit 1, (TAC MD3053) 8.0 References

1. Letter from H. K. Nieh (NRC) to R. C. Bunt (Southern Nuclear Operating Company),

"Final Safety Evaluation for the Boiling Water Reactor Owners' Group (BWROG)

Structural Integrity Associates Topical Report (TR) SIR-05-044, "Pressure Temperature Report Methodology forBoiling Water Reactors" (TAC NO. MC9694)", dated February 6,-

2007.

2. Generic Letter (GL) 96-03, "Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits", January 31, 1996.

3. Structural Integrity Associates Topical Report (TR) SIR-05-044-A, "Pressure-Temperature Report Methodology for Boiling Water Reactors", April 2007.

4. TSTF-419-A, "Revised PTLR Definition and References in ISTS 5.6.6, RCS PTLR",

dated August 4, 2003.

5. Letter from J.B. Boska (NRC) to M. Kansler (Entergy Nuclear Operations, Inc), "JAF Nuclear Power Plant (JAF) - Issuance of Amendment RE: Changes to the Reactor Vessel Specimen Material Surveillance Program(TAC NO. MC9682)", dated July 26, 2006.

6. "Safety Evaluation Report Related to the License Renewal of JAFNPP", dated January 24, 2008.

7. Letter from P. Dietrich to NRC, License Renewal Amendment 14, JAFP 07-0125, dated November 5, 2007.

Page 9 of 9

JAFP-08-0034 Attachment 2 Proposed Technical Specification Changes (Mark up)

Pages 1.1-4 3.4.9-1 3.4.9-3 3.4.9-4 3.4.9-5 3.4.9-6 3.4.9-7(deleted) 3.4.9-8(deleted) 5.6-3

Definitions 1.1 1.1 Definitions (continued)

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit RATE (LHGR) length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power that RATIO (MCPR) exists in the core for each type of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

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

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatuD and cooldown rates. for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

(continued)

JAFNPP 1.1-4 Amendment 287

RCS P/T Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR. I APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.------- NOTE -------- A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed if this Condition is entered. AND A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Requirements of the LCO acceptable for not met in MODE 1, 2, or 3. continued operation.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

JAFNPP 3.4.9-1 Amendment 2-74

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 --------------- NOTE------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits specified in Figrc.3.4.9 1 or Figc,

.4.9 2 the curves in the PTLR as applicable; and

b. RCS temperature change averaged over a one hour period is:

1. _*100°F when the RCS pressure and RCS temperature are on or to the right of curve C ef-Fig .4.0 1 or. Furc,3.4.9

• .. 2 in the PTLR as applicable, during inservice leak and hydrostatic testing;

2. _ 20°F when the RCS pressure and RCS temperature are to the left of curve C ef-Figure-A491o Figuc 3.4.9.2 in the PTLR as applicable, during inservice leak and hydrostatic testing; and

3. < 100°F during other heatup and cooldown operations.

(continued)

JAFNPP 3.4.9-3 Amendment 2-7-4

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and RCS temperature are within Once within the criticality limits specified in Figurc 3.4.9 1 or 15 minutes prior to FigU.. 3.4.9 2 as .ppli.. the PTLR. control rod withdrawal for the purpose of achieving criticality SR 3.4.9.3 --------------- NOTES------------

1., Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

2. Not required to be performed if SR 3.4.9.4 is satisfied.

Verify the difference between the bottom head coolant Once within temperature and the reactor pressure vessel (RPV) 15 minutes prior to coolant temperature is < 145°F within the limits each startup of a recirculation pump I specified in the PTLR.

SR 3.4.9.4 -------------- NOTES-------------

1. Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

2. Not required to be met if SR 3.4.9;3 is satisfied.

Verify the active recirculation pump flow exceeds 40%

of rated pump flow or the active recirculation pump Once within has been operating below 40% rated flow for a period 15 minutes prior to no longer than 30 minutes. each startup of a recirculation pump (continued)

JAFNPP 3.4.9-4 Amendment 27-4

RCS P/T Limits 3.4.9

'SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.5 ----------------- NOTES -------------- Once within Only required to be met in MODES 1, 2, 3, and 4 15 minutes prior during recirculation pump startup. to each startup of a recirculation pump Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is L<500 within the limits specified in the PTLR.

SR 3.4.9.6 ---------------- NOTES-------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify, when the reactor vessel head bolting studs are 30 Minutes under tension, reactor vessel flange and head flange temperatures are Ž90ýF within the limits specified in the PTLR.

SR 3.4.9.7 -------------- NOTES-------------

Not required to be performed until 30 minutes after RCS temperature < IG00 80°F with any reactor vessel.

head bolting stud tensioned.

Verify, when the reactor vessel head bolting studs are 30 minutes under tension, reactor vessel flange and head flange temperatures areŽ-90:F within the limits specified in the PTLR.

(continued)

JAFNPP 3.4.9-5 Amendment 2-7-4

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

.4.

SR 3.4.9.8 -- -- -- -

-NOTES-------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 1-2-- 100°F with any reactor vessel head bolting stud tensioned.

Verify, when the reactor vessel head bolting studs are 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> under tension, reactor vessel flange and head flange temperatures are ý-90OF. within the limits specified in the PTLR.

I JAFNPP- 3.4.9-6 Amendment 2-74

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatu,. 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:

D) Limiting Conditions for Operation Section 3.4.9 "ROS Pressure and Temperature (P/T) Limits" ii) Surveillance Reauirements Section 3.4.9 "RCS Pressure and Temperature (P/T) 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 document:

i) SIR-05-044-A. "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors"

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

JAFNPP 5.6-3 Amendment 282

JAFP-08-0034 Attachment 3 Proposed Technical Specification Changes (Final Typed)

Pages 1.1-4 3.4.9-1 3.4.9-3 3.4.9-4 3.4.9-5 3.4.9-6 3.4.9-7(deleted) 3.4.9-8(deleted) 5.6-3

Definitions 1.1 1.1 Definitions (continued)

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit RATE (LHGR) length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power that RATIO (MCPR) exists in the core for each type of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

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

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

(continued)

JAFNPP 1.1-4 Amendment

RCS P/T Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR. I APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.------- NOTE - ------- A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed if this Condition is entered. AND A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Requirements of the LCO acceptable for not met in MODE 1, 2, or 3. continued operation.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

JAFNPP 3.4.9-1 Amendment

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 ---------------- NOTE------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits specified in the curves in the PTLR as applicable; and

b. RCS temperature change averaged over a one hour period is:

1. < 100°F when the RCS pressure and RCS temperature are on or to the right of curve C in the PTLR as applicable, during inservice leak and hydrostatic testing;

2. _ 20°F when the RCS pressure and RCS temperature are to the left of curve C in the PTLR as applicable, during inservice leak and hydrostatic testing; and

3. < 100°F during other heatup and cooldown operations.

(continued)

JAFNPP 3.4.9-3 Amendment

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and RCS temperature are within Once within the criticality limits specified in the PTLR. 15 minutes prior to control rod I withdrawal for the purpose of achieving criticality SR 3.4.9.3 ---------------- NOTES------------

1. Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

2. Not required to be performed if SR 3.4.9.4 is satisfied.

Verify the difference between the bottom head coolant Once within temperature and the reactor pressure vessel (RPV) 15 minutes prior to coolant temperature is within the limits specified in each startup of a the PTLR. recirculation pump SR 3.4.9.4 -------------- NOTES-------------

1. Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

2. Not required to be met if SR 3.4.9.3 is satisfied.

Verify the active recirculation pump flow exceeds 40%

of rated pump flow or the active recirculation pump Once within has been operating below 40% rated flow for a period 15 minutes prior to no longer than 30 mihutes. each startup of a recirculation pump (continued)

JAFNPP 3.4.9-4 Amendment

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 4

SR 3.4.9.5 -- -- -

-- -NOTES------------- Once within Only required to be met in MODES 1, 2, 3, and 4 15 minutes prior during recirculation pump startup. to each startup of a recirculation pump Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is within the limits specified in the PTLR. I 4

SR 3.4.9.6 -- -- -

-- -NOTES-------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify, when the reactor vessel head bolting studs are 30 Minutes under tension, reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.

4 SR 3.4.9.7 -

-- -- -- -NOTES--------------

Not required to be performed until 30 minutes after RCS temperature < 80°F with any reactor vessel head bolting stud tensioned.

Verify, when the reactor vessel head bolting studs are 30 minutes under tension, reactor vessel flange and head flange temperatures are within the limits specified in the PTLR. I (continued)

JAFNPP 3.4.9-5 Amendment

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

.1.

SR 3.4.9.8 -- -- -

-- -NOTES-------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 100OF with any reactor vessel head I bolting stud tensioned.

Verify, when the reactor vessel head bolting studs are 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> under tension, reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.

I JAFNPP 3.4.9-6 Amendment

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

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:

i) Limiting Conditions for Operation Section 3.4.9 "RCS Pressure and Temperature (P/T) Limits" ii) Surveillance Requirements Section 3.4.9 "RCS Pressure and Temperature (P/T) 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 document:

i) SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors"

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

JAFNPP 5.6-3 Amendment

JAFP-08-0034 Attachment 4 Proposed Technical Specification Bases Changes (Mark Up)

(Information Only)

Pages B 3.4.9-1 B 3.4.9-2 B 3.4.9-3 B 3.4.9-4 B 3.4.9-6 B 3.4.9-7 B 3.4.9-9 B 3.4.9-10

RCS P/T Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The PTLR This Sp, , ,fie;tien contains P/T limit curves for heatup, cooldown, inservice leakage and hydrostatic testing, and criticality and also limits the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation.

The curves are used for operational guidance during heatup or cooldown maneuvering. Pressure and temperature are monitored and compared to the applicable curve to ensure that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure.

Therefore, the LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.

Reference 1 requires an adequate margin to brittle failure during normal operation, abnormal operational transients, and system inservice leakage and hydrostatic tests. It mandates the use of the ASME Code,Section III, Appendix G (Ref. 2).

The nil-ductility transition (NDT) temperature, RTNDT, is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner. The RTNDT increases as a function of neutron exposure at integrated neutron exposures greater than approximately 1017 nvt with neutron energy in excess of I MeV.

The actual shift in the RTNDT of the vessel material is determined periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance (continued)

JAFNPP B 3.4.9-1 Revision 0

RCS P/T Limits B 3.4.9 BASES BACKGROUND with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4), and the (continued) BWR Vessel and Internals Project (VIP) Integrated Surveillance Program (ISP) (Ref.13). The operating P/T limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive locations.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. However, the P/T limit curves reflect the most restrictive of the heatup and cooldown curves.

The P/T criticality limits include the Reference 1 requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference -7 14 establishes the methodology for determining the P/T limits. Referenee 8

(continued)

JAFNPP B 3.4.9-2 Revision 16

RCS P/T Limits B 3.4.9 BASES APPLICABLE app..vcd thc uF.v.. and ,limit... U..cd by this SAFETY ANALYSES Speeieatiep, Since the P/T limits are not derived from (continued) any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 9).

LCO The elements of this LCO are:

a. RCS pressure and temperature are within the limits specified in Figuc 3.4.9 :1 or FigU.. 3.4.9 2, as appli.ablc.the PTLR. In addition, RCS temperature change averaged over a one hour period is: <

100°F when the RCS pressure and temperature are on or to the right of curve C Figure 3.4.9 1 Or FIgUrc 3.4.9 2, as applioablc in-the PTLR, as applicable, during inservice leak and hydrostatic testing; <

20°F when the RCS pressure and temperature are to the left of curve C Figurc 3.4.9 1 or Figr,, 3.4.9 2, as appli-abic in the PTLR, during inservice leak and hydrostatic testing; and < 100OF during other heatup and cooldown operations;

b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is 464I within the limits soecified in the PTLR during recirculation pump startup;

c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is 5--O 2F within the limits specified in the PTLR during recirculation pump startup;

d. RCS pressure and temperature are within the limits specified in Figurc 3.4.9 . or Figur 3.4.9 2, as appli.abl. the PTLR, prior to achieving criticality; and

e. The reactor vessel flange and the head flange temperatures are 902F within the limits specified in the PTLR when tensioning the reactor vessel head bolting studs and when any stud is tensioned.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

(continued)

JAFNPP B 3.4.9-3 Revision 0

RCS P/T Limits B 3.4.9 BASES LCO The limits on the rate of change of RCS temperature, (continued) influenced by RCS flow and RCS stratification, control the thermal gradient through the vessel wall. For this reason, both RCS temperature and RPV metal temperatures are used as inputs for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing P/T limit curves.

Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.

P/T limit curves are provided for plant operations through 24 EFPY,-,Fig&*e-

.4.9 1) a-nd 32 EFPY (Fig*-.e.4.9-2.,in the PTLR. Curves A, ABH (bottom head), and ANB (non-beltline) establish the minimum temperature for hydrostatic and leak testing, Curves B and BBH (bottom head) establish limits for plant heatup and cooldown when the reactor is not critical or during low power physics tests, and Curve C establishes the limits when the reactor is critical. In addition, ART is the adjusted reference temperature.,

Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;

b. he length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to' become more pronounced); and

c. The existence, size, and orientation of flaws in the vessel material.

APPLICABILITY The potential for violating a P/T limit exists at all times. For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.

(continued)

JAFNPP B 3.4.9-4 Revision 0

RCS P/T Limits B 3.4.9 BASES ACTIONS B.1 and B.2 (continued) reduced pressure and temperature. With the reduced pressure and temperature conditions, the likelihood of propagation of undetected flaws is decreased.

Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.

Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 2120 F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code, Section Xl, Appendix E (Ref. 6),

may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within RGS prczzurc and t.mpera... .im;its well as within RCS tcmpeat... " hang. the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are as I

undergoing planned changes. This (continued)

JAFNPP B 3.4.9-6 Revision 0

RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)

REQUIREMENTS is accomplished by monitoring the bottom head drain, recirculation loop, and RPV metal temperatures. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations. The limits Of Figur.. 3.4.9 1 and 3.4.9 2 in the PTLR are met when operation is on or to the right of the applicable curve.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given~in the relevant plant procedure for ending the activity are satisfied. In general, if two consecutive temperature readings taken > 30 minutes apart are within 5 0 F of each other the activity can be considered complete.

This SR is modified by a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing. Unlike steady-state operation, these intentional operational transients may be characterized by large pressure and temperature changes, and performance of this SR provides assurance that RCS pressure and temperature remain within acceptable regions of the P/T limit curves as well as within RCS temperature change limits.

SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality.

Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.

Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

SR 3.4.9.3. SR 3.4.9.4. and SR 3.4.9.5 Differential temperatures within the specified limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In (continued)

JAFNPP B 3.4.9-7 Revision 0

RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.6. SR 3.4.9.7. and SR 3.4.9.8 (continued)

REQUIREMENTS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.

The flange temperatures .must be verified to be above the limits within 30 minutes before and while tensioning the reactor vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied.

When any reactor vessel head bolting stud is tensioned with RCS temperature I -0 800 F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When any reactor vessel head bolting stud is tens~ioned with RCS temperature < 420 100 0 F, monitoring of the flange temperature is required every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within specified limits.

The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.

SR 3.4.9.6 is modified by a Note which requires the SR to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.9.7 is modified by a Note which states that the SR is not required to be performed until 30 minutes after RCS temperature is <400 80°F in MODE

4. SR 3.4.9.8 is modified by a Note which states that the SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is _ 420 IO0°F in MODE 4. These Notes are necessary to specify when the reactor vessel flange and head flange temperatures are required to be within specified limits.

REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

3. ASTM E 185-82, July 1982.

4. 10 CFR 50, Appendix H.

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

6. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.

(continued)

JAFNPP B 3.4.9-9 Revision 0

RCS P/T Limits B 3.4.9 BASES REFERENCES 7. GE-NE-B1100732-01, Revision 1, Plant FitzPatrick RPV (continued) Surveillance Materials Testing and Analysis of 1200 Capsule at 13.4 EFPY, February 1998, including Errata and Addenda Sheets dated June 17, 1999 and December 3, 1,999.

8. Letter from Guy Vissing (NRC) to James Knubel (NYPA) Issuance of Amendment No. 258 to James A. FitzPatrick Nuclear Power Plant, November 29, 1999.

9. 10 CFR 50.36(c)(2)(ii).

10. UFSAR, Section 14.5.7.2.

11. GE-NE-208-04-1292, Evaluation of Idle Recirculation Loop Restart Without Vessel Bottom Temperature Indication for FitzPatrick Nuclear Power Plant, December 1992..

12. JAF-RPT-RWR-02076, Verification of Alternative Operating Conditions for Idle Recirculation Loop Restart Without Vessel Bottom Temperature Indication, June 25, 1995.

13. UFSAR, Section 4.2.7.

14. SIR-05-044-A. "Pressure-Temperature Limits Report Methodology For Boiling Water Reactors". dated April 12. 2007.

JAFNPP B 3.4.9-10R Revision 6

JAFP-08-0034 Attachment 5 Pressure and Temperature Limits Report (PTLR)

ENTERGY NUCLEAR OPERATIONS, INC.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT REPORT PRESSURE AND TEMPERATURE LIMTS REPORT (PTLR) UP TO 32 EFFECTIVE FULL-POWER YEARS REVISION 0 Prepared by: t YC. U-1_r&, LN Date: C_-)

Reviewed by: Date: 4./0o8 Approved by: Date: (

Concurred by: Date: ______/__

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

" Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing

" RCS Heatup and Cooldown rates

• Reactor Pressure Vessel (RPV) to RCS coolant AT requirements during Recirculation Pump startups

" RPV bottom head coolant temperature to RPV coolant AT requirements during Recirculation Pump startups

• RPV head flange boltup temperature limits This report has been prepared in accordance with the requirements ofTechnical Specification 5.6.7, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."

2.0 APPLICABILITY This report is applicable to the JAF RPV for up to 32 Effective Full-Power Years (EFPY).

The following JAF Technical Specification (TS) is affected by the information contained in this report:

• Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 ("RCS Pressure and Temperature (P/T) Limits"),

The JAF Reactor Vessel Pressure and Temperature Limits for 32 to 54 EFPY have been developed per Reference 6.4. Only the 32 EFPY Limits are incorporated in this revision of the PTLR. Future revisions of the. PTLR must be revised per the 10CFR50.59 Review process as applicable.

Revisioni 0 Page 2 of 22

JAF Pressure-Temperature Limits Report 3.0 METHODOLOGY The limits in this report were derived as follows:

1) The methodology used is in accordance with Reference 6.1, which has been approved for BWR use by the NRC.

2) The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 Reference 6.5, using the RAMA computer code, as documented in Reference 6.2.

3) The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2.

Reference 6.6, and supplemented by data from the BWRVIP Integrated Surveillance Program in Reference 6.14, as documented in Reference 6.3.

4) This revision of the pressure and temperature limits is to incorporate the following changes:

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

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

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

The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not Revision 0 Page 3 of 22

JAF Pressure-Temperature Limits Report critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C, Reference 6.10.

Complete P-T curves were developed for 32 EFPY for JAF, as documented in Reference 6.4. The JAF P-T curves for 32 EFPY are provided in Figures 1 through 3, and a tabulation of the curves is included in Tables 1 through 3. For both Curves A and B in Tables 1 and 2, respectively, three tables are included for the beltline, bottom head, and upper vessel/feedwater nozzle regions.

Other temperature limits applicable to the Reactor Pressure Vessel are:

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

& Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C - nuclear heating): < 100OF/hour 2 .

0 RPV bottom head coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: < 145 OF.

0 Recirculation loop coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: _<50 0 F.

0 RPV head installation temperature limit (Figure 1: Curve A - Hydrostatic and Class 1 Leak Testing; Figure 2: Curve B.- non-nuclear heating): _ 60 0 F.

• RPV flange and adjacent shell temperature limit: _ 601F.

5.0 DISCUSSION The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide (RG) 1.99 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.

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

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

Revision 0 Page 4 of 22

JAF Pressure-Temperature Limits Report The vessel beltline copper and nickel values were obtained from the evaluation of the JAF vessel plate and weld materials (Reference 6.3). The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds and plates, respectively. Because JAF participates in the BWRVIP Integrated Surveillance Program, there are two exceptions to the use of Paragraph 1.1 in setting the CF. Plate C3278-2 uses a fitted CF based on JAF and Supplemental Surveillance Program (SSP) test data, and Lower Intermediate Longitudinal Welds 1-233-A, B & C use an adjusted CF based on ISP surveillance results, see References 6.3 and 6.14.

2 The peak RPV ID fluence used in the P-T curve evaluation for 32 EFPY is 2.01x1 018 n/cm for JAF from Reference 6.2, which was calculated using methods that comply with the guidelines of RG 1.190 (Reference 6.5). An ID fluence value of 1.686x1018 n/cm 2 applies to the limiting beltline lower shell plate C3376-2 for JAF. This fluence value was adjusted for the limiting lower shell plates based upon an attenuation factor of 0.682 for a postulated 1/4t flaw. As a result, the 1/4t 32 EFPY fluence for the limiting beltline lower shell plate is 1.15x10 18 n/cm 2 for JAF. An ID fluence value of 1.544x1018 n/cm 2 applies to the limiting beltline lower intermediate shell weld 27204/12008. This fluence value was adjusted based upon an attenuation factor of 0.724 for a postulated 1/4t flaw. As a result, the 1/4t 32 EFPY fluence for the limiting beltline lower intermediate shell weld is 1.1 19x1018 n/cm 2 for JAF.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4t and 3/4t locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4t location (inside surface flaw) and the 3/4t location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4t location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4t location. This approach is conservative because irradiation effects cause the allowable toughness at 1/4t to be less than that at 3/4t for a given metal temperature. This approach causes no operational Revision 0 Page 5 of 22

JAF Pressure-Temperature Limits Report difficulties, since the BWR is at steam saturation conditions during normal operation, which is well above the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of < 100°F/hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams, Reference 6.7. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of < 20°F/hr must be maintained.

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

The initial RTNDT, the chemistry (weight-percent copper and nickel) and adjusted reference temperature at the 1/4 thickness location for all RPV beltline materials significantly affected byfluence (i.e., fluence > 1017 n/cm 2 for E > 1 MeV) are shown in Table 4 for 32 EFPY.

The initial RTNDT values shown in Table 4 (obtained from Reference 6.8) have been previously approved for use by the NRC (Reference 6.13).

Per Reference 6.3 and in accordance with Appendix A of Reference 6.1, the JAF representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). The representative heats of weld and plate material in the ISP are identical to the heats in the JAF vessel for two materials, plate heat number C3278-2 and weld heat number 13253/12008.

For the RPV Plate number C3278-2, capsule test data is available both from tested JAF surveillance capsules and from Supplemental Surveillance Program (SSP) capsules. The BWRVIP ISP Source Book Implementation "Procedure 1"approach was used to obtain a Revision 0 Page 6 of 22

JAF Pressure-Temperature Limits Report fitted chemistry factor of 89.56°F for this material (Reference 6.14). The data was determined to be credible; however, the scatter in the surveillance data exceeds the credibility criteria, so a standard full margin (oa = 17.0 °F) was used in the ART calculations shown in Table 4 (Reference 6.3).

For the RPV welds 2-233A, B & C (heat number 27204/12008), the representative weld material (27204) is not the same heat number as the target weld (27204/12008).

Therefore, BWRVIP ISP Implementation "Procedure 2" was utilized since the heat number of this material is different than the heat number of the ISP Representative Material.

For RPV welds 1-233A, B &C (heat number 13253/120008), BWRVIP ISP Implementation "Procedure 1" was utilized since the heat number of this material is identical to the BWRVIP ISP Representative Material and the surveillance data is deemed to be credible.

The capsule data from the ISP surveillance data report specifies a fitted chemistry factor of 323.68°F for the 13253/12008 weld heat (Reference 6.14). Since the surveillance weld chemistry differs slightly from the vessel best-estimate weld chemistry, an adjusted chemistry factor of 326.940F was calculated using the ratio procedure. Because the surveillance data is credible, the margin term was cut in half (OA, = 14.0 OF) in calculations of the ART values for the JAF vessel welds made with this heat number as shown in Table 4 (Reference 6.3).

Revision 0 Page 7 of 22

JAF Pressure-Temperature Limits Report

6.0 REFERENCES

6.1 Structural Integrity Associates Report No. SIR-05-044-A, Revision 0, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," April 2007 6.2 TransWare Enterprises Inc. Report No. ENT-FLU-002-R-001, Revision 0, "JAF Reactor Pressure Vessel Fluence Evaluation at End Of Cycle 17 and 54 EFPY," October 2007.

6.3 Structural Integrity Associates, Inc. Calculation No. FITZ-1OQ-301, Revision 0, "Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts,"

February, 2008.

6.4 Structural Integrity Associates, Inc. Calculation No. FITZ-10Q-302, Revision 0, "Revised P-T Curves," February, 2008.

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

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

6.7 JAF drawing 5.15-1A, GE drawing 729E762 RO, Reactor Thermal Cycles 6.8 General Electric Report GENE-B1100732-01, Revision 1, "Plant Fitzpatrick RPV Surveillance Materials Testing and Analysis of 120 degree capsule at 13.4 EFPY,"

February 1998.

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

6.10 Part 50 of Title 10 of the Code of Federal Regulations, Appendix G, 'Fracture Toughness Requirements," January 2005.

6.11 Part 50 of Title 10 of the Code of Federal Regulations, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," January 2005.

6.12 JAF TS Amendment #285, Changes to the Reactor Vessel Material Surveillance Program, dated July 26, 2006.

6.13 JAF License Renewal Final SER, dated January 24, 2008, Open Item 4.2.2-1 (P-T Limits).

6.14 EPRI Technical Report No. 1013400, "BWRVIP-135 Revision 1: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations," June 2007.

Revision 0 Page 8 of 22

JAF Pressure-Temperature Limits Report Figure 1: FitzPatrick Pressure Test (Curve A) --- 32 EPFY 1,600 - " I I_ -

' ~I 1,500 1 ,40 0 - . ..

.I...... .............

I 4/

1,3 0 0 -. .....

0 1,200 .............

[.....

1<10 ............ ..... ....

. .......-1 uJFO1 ,0 0 0 I... .... )

S- - Bottom Head O 800 ......... Upper Vessel

• - Beltline LU 700- .I.... ........

I- 600- .. I.....

LU 50 0 .......... I............

C,)

LU400-1-0 0 . ............. ..... . ... .... .. . .......... .. ................. ................

300 - ... ..... ....... ......

O_1_ ___ _ ___ __

200 Bolt-up... ....

Ternp 00 L¶1 0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Revision 0 Page 9 of 22

JAF Pressure-Temperature Limits Report Figure 2: Normal Operation Core Not Critical (Curve B) --- 32 EFPY 1,600 1,500 1,400 1,300 1,200 U 1,100 01,000

-j LU 0~900

-. 800 0

< 700 Zf z-600 500 400 n 300 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Revision 0 Page 10 of 22

JAF Pressure-Temperature Limits Report Figure 3: Normal Operation - Core Critical (Curve C) --- 32 EFPY 1,600 1,500 1,400 1,300

• 1,200 CL

<1,100 w

o 1,000 I--

_j Co 900 o) uJ

" 800

UJ 700

.z

• _ 600

._j w 500 Co Co 400 ry 300 Criticality Temp 90*F , .

200 ............ ... .............

..I - .. ..............

100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (ff)

Revision 0 Page 11 of 22

JAF Pressure-Temperature Limits Report Table la: JAF Pressure Test (Curve A) --- Beltline (32EFPY)

Plant = Fitz~patrick Component = Beltline Vessel thickness, t = 5.375 inches Vessel Radius, R = 110.375 inches ART = 1096 'F .======> 32 EFPY KIT = 0*00 (no thermal effects)

Safety Factor =  : 1 50 Mm = 2.47 Temperature Adjustment = 0°0 'F (applied after bolt-up, instrument uncertaii nty)

Height of Water for a Full Vessel = 825.20 inches

'Pressure Adjustment = 298* psig (hydrostatic pressure head for a full ves sel at 70'F)

Pressure Adjustment = 0.0 psig (instrument uncertainty)

Gauge Adjusted Fluid Temperature Pressure for Temperature Kic Kim for P-T Curve P-T Curve (OF) (ksi*inchl/2) (ksi*inchl/2) (°F) (psig) 60.0 40.89 27.26 60.0 0 60.0 40.89 27.26 60.0 589 62.0 41.20 27.47 62.0 593 64.0 41.53 27.69 64.0 598 66.0 41.87 27.91 66.0 603 68.0 42.22 28.15 68.0 609 70.0 42.59 28.39 70.0 614 72.0 42.97 28.65 72.0 620 74.0 43.37 28.92 74.0 626 76.0 43.79 29.19 76.0 632 78.0 44.22 29.48 78.0 639 80.0 44.67 29.78 80.0 646 82.0 45.14 30.09 82.0 653 84.0 45.63 30.42 84.0 660 86.0 46.13 30.76 86.0 668 88.0 46.66 31.11 88.0 676 90.0 47.21 31.47 90.0 684 92.0 47.78 31.85 92.0 693 94.0 48.38 32.25 94.0 702 96.0 49.00 32.66 96.0 711 98.0 49.64 33.09 98.0 721 100.0 50.31 33.54 100.0 731 102.0 51.01 34.01 102.0 742 104.0 51.74 34.49 104.0 753 106.0 52.49 35.00 106.0 764 108.0 53.28 35.52 108.0 776 110.0 54.10 36.07 110.0 788 112.0 54.95 36.64 112.0 801 114:0 55.84 37.23 114.0 815 116.0 56.77 37.84 116.0 829 118.0 57.73 38.48 118.0 843 120.0 58.73 39.15 120.0 858 122.0 59.77 39.85 122.0 874 124.0 60.85 40.57 124.0 890 126.0 61.98 41.32 126.0 908 128.0 63.16 42.10 128.0 925 130.0 64.38 42.92 130.0 944 132.0 65.65 43.77 132.0 963 134.0 66.98 44.65 134.0 983 136.0 68.36 45.57 136.0 1,004 138.0 69.79 46.53 138.0 1,026 140.0 71.28 47.52 140.0 1,048 142.0 72.84 48.56 142.0 1,072 144.0 74.46 49.64 144.0 1,096 146.0 76.14 50.76 146.0 1,122 148.0 77.89 51.93 148.0 1,148 150.0 79.72 53.14 150.0 1,176 152.0 81.61 54.41 152.0 1,204 Revision 0 Page 12 of 22

JAF Pressure-Temperature Limits Report Table la: JAF Pressure Test (Curve A) --- Beltline (32EFPY), continued Gauge Adjusted Fluid Temperature Pressure for Temperature KIc Kim for P-T Curve P-T Curve

(*F) (ksi*inchl/2) (ksi*inchl/2) (°F) (psig) 154.0 83.59 55.73 154.0 1,234 156.0 85.65 57.10 156.0 1,265 158.0 87.79 58.52 158.0 1,298 160.0 90.01 60.01 160.0 1,331 162.0 92.33 61.55 162.0 1,366 164.0 94.75 63.16 164.0 1,403 166.0 97.26 64.84 166.0 1,441 168.0 99.87 66.58 168.0 1,480 170.0 102.59 68.39 170.0 1,522 172.0 105.42 70.28 172.0 1,564 174.0 108.37 72.25 174.0 1,609 Revision 0 Page 13 of 22

JAF Pressure-Temperature Limits Report Table Ib: JAF Pressure Test (Curve A) --- Bottom Head (All EFPY)

Component = (penetrations portion)

Bottom Head thickness, t = inches Bottom Head Radius, R = inches ART = 'F ...... => All EFPY Kit = (no thermal effects)

Safety Factor =

Stress Concentration Factor = (bottom head penetrations)

Mm =

0 Temperature Adjustment = F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment = psig (instrument uncertainty)

Gauge Adjusted Fluid Temperature Pressure for Temperature Kic Kim for P-T Curve P-T Curve

(°F) (ksi*inchlIZ) (ksi*inchllZ) (6F) (psig) 60.0 89.56 59.71 60 0 60.0 89.56 59.71 60 987 62.0 91.86 61.24 62 1,013 64.0 94.25 62.84 64 1,040 66.0 96.75 64.50 66 1,068 68.0 99.34 66.23 68 1,098 70.0 102.04 68.03 70 1,128 72.0 104.85 69.90 72 1,160 74.0 107.77 71.85 74 1,193 76.0 110.82 73.88 76 1,228 78.0 113.98 75.99 78 1,264 80.0 117.28 78.19 80 1,301 82.0 120.71 80.47 82 1,340 84.0 .124.28 82.86 84 1,381 86.0 128.00 85.33 86 1,423 88.0 131.87 87.91 88 1,467 90.0 135.90 90.60 90 1,513 92.0 140.09 93.39 92 1,560 94.0 144.45 96.30 94 1,610 Revision 0 Page 14 of 22

JAF Pressure-Temperature Limits Report Table Ic: JAF Pressure Test (Curve A) --- Upper Vessel (All EFPY)

Component = (based on FW nozzle)

ART = 'F ...... => All EFPY Vessel Radius, R = inches Nozzle corner thickness, t' = inches, approximate Kit = (no thermal effects)

KIp-applied = ksi*inchl/2 Crack Depth, a = inches Safety Factor =

Temperature Adjustment = *F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment = psig (instrument uncertainty)

Reference Pressure = psig (pressure at which the FEA stress coefficients are valid) [GREEN]

Unit Pressure = psig (hydrostatic pressure)

Flange RTNDT = °F ======> All EFPY Gauge P-T P-T Fluid Curve Curve Temperature Ki. Kip Temperature Pressure 11 2 (psiq)

('F) (ksi*inch ) (ksi*inchllz) (6F) 60.0 79.34 52.90 60 0 60.0 79.34 52.90 260 313 62.0 81.23 54.15 120 313 64.0 83.19 55.46 120 957 66.0 85.23 56.82 120 982 68.0 87.35 58.23 120 1007 70.0 89.56 59.71 120 1033 72.0 91.86 61.24 120 1060 74.0 94.25 62.84 120 1089 76.0 96.75 64.50 120 1118 78.0 99.34 66.23 120 1149 80.0 102.04 68.03 120 1181 82.0 104.85 69.90 120 1215 84.0 107.77 71.85 120 1249 86.0 110.82 73.88 120 1285 88.0 113.98 75.99 120 1323 90.0 117.28 78.19 120 1362 92.0 120.71 80.47 120 1403 94.0 124.28 82.86 120 1445 96.0 128.00 85.33 120 1489 98.0, 131.87 87.91 120 1535 100.0 135.90 90.60 120 1583, 102.0 140.09 93.39 120 1633 Revision 0 Page 15 of 22

JAF Pressure-Temperature Limits Report Table 2a: JAF Core NotCritical (Curve B) --- Beltine (32 EFPY)

Component =

Vessel thickness, t = inches Vessel Radius, R = inches ART = *F======> 32 EFPY ksi*inch"2 Kit =

Safety Factor =

Mm =

Temperature Adjustment = °F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment = psig (instrument uncertainty)

Heat Up and Cool Down Rate = °F/Hr Gauge Adjusted Fluid Temperature Pressure for Temperature Ki. KIm for P-T Curve P-T Curve (OF) (ksi*inchl' 2) (ksi-inch112) (°F) (psig) 60.0 40.89 17.25 60.0 0 60.0 40.89 17.25 60.0 362 62.0 41.20 17.41 62.0 365 64.0 41.53 17.57 64.0 369 66.0 41.87 17.74 66.0 373 68.0 42.22 17.92 ,68.0 377 70.0 42.59 18.10 70.0 381 72.0 42.97 18.30 72.0 385 74.0 43.37 18.50 74.0 390 76.0 43.79 18.70 76.0 394 78.0 44.22 18.92 78.0 399 80.0 44.67 19.14 80.0 404 82.0 45.14 '19.38 82.0 410 84.0 45.63 19.62 84.0 415 86.0 46.13 19.87 86.0 421 88.0 46.66 20.14 88.0 427 90.0 47.21 20.41 90.0 433 92.0 47.78 20.70 .92.0 440 94.0 48.38 21.00 94.0 446 96.0 49.00 21.31- 96.0 454 98.0 49.64 21.63 98.0 461 100.0 50.31 21.96 100.0 468 102.0 51.01 22.31 102.0 476 104.0 51.74 22.68 104.0 485 106.0 52.49 23.06 106.0 493 108.0 53.28 23.45 108.0 502 110.0 54.10 23.86 110.0 511 112.0 54.95 24.29 112.0 521 114.0 55.84 24.73 114.0 531 116.0 56.77 25.19 116.0 542 118.0 57.73 25.67 118.0 553 120.0 58.73 26.17 120.0 564 122.0 59.77 26.69 122.0 576 124.0 60.85 27.24 124.0 588 126.0 61.98 27.80 126.0 601 128.0 63.16 28.39 128.0 614 130.0 64.38 29.00 130.0 628 132.0 65.65 29.63 132.0 642 134.0 66.98 30.30 134.0 657 136.0 68.36 30.99 136.0 673 138.0 69.79 31.70 138.0 689 140.0 71.28 32.45 140.0 706 142.0 72.84 33.23 142.0 724 144.0 74.46 34.04 144.0 742 146.0 76.14 34.88 146.0 761 148.0 77.89 35.75 148.0 781 150.0 79.72 36.67 150.0 802 152,0 81.61 37.62 152.0 823 154.0 83.59 38.60 154.0 846 156.0 85.65 39.63 156.0 869 158.0 87.79 40.70 158.0 893 160.0 90.01 41.82 160.0 919 Revision 0 Page 16 of 22

JAF Pressure-Temperature Limits Report Table 2a: JAF Core Not Critical (Curve B) --- Beltine (32 EFPY), continued Gauge Adjusted Fluid Temperature Pressure for Temperature KI. KIm for P-T Curve P-T Curve (OF)

(ksi*inch ... (ksi-inchvz 111°F)

(Dsia) 162.0 92.33 42.97 162.0 945 164.0 94.75 44.18 164.0 972 166.0 97.26 45.44 166.0 1,001 168.0 99.87 46.74 168.0 1,031 170.0 102.59 48.10 170.0 1,061 172.0 105.42 49.52 172.0 1,093 174.0 108.37 50.99 174.0 1,127 176.0 111.44 52.53 176.0 1,162 178.0 114.63 54.12 178.0 1,198 180.0 117.96 55.79 180.0 1,236 182.0 121.41 57.52 182.0 1,275 184.0 125.01 59.32 184.0 1,316 186.0 128.76 61.19 186.0 1,358 188.0 132.66 63.14 188.0 -1,402 190.0 136.72 65.17 190.0 1,448 192.0 140.95 67.28 192.0 1,496 194.0 145.34 69.48 194.0 1,546 196.0 149.92 71.77 196.0 1,598 198.0 154.68 74.15 198.0 1,652 Revision 0 Page 17 of 22

JAF. Pressure-Temperature Limits Report Table 2b: JAF Core Not Critical (Curve B) --- Bottom Head (All EFPY)

Component = (penetrations portion)

Bottom Head thickness, t = inches Bottom Head Radius, R = inches ART = 'F ======>

12 All EFPY Kit =

ksi*inch Safety Factor =

Stress Concentration Factor = (bottom head penetrations)

Mrn =

Temperature Adjustment = 'F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment = psig (instrument uncertainty)

Heat Up and Cool Down Rate = °F/Hr Gauge Adjusted Fluid Temperature Pressure for Tern perature Ki. Kim for P-T Curve P-T Curve

(*F) (ksi*inch1 2 ) (ksi*inch t z) (6F) (psig) 60.0 89.56 39.01 60 0 60.0 89.56 39.01 60 634 62.0 91.86 40.16 62 654 64.0 94.25 41.36 64 674 66.0 96.75 42.60 66 695 68.0 99.34 43.90 68 718 70.0 102.04 45.25 70 741 72.0 104.85 46.65 72 764 74.0 107.77 48.11 74 789 76.0 110.82 49.64 76 815 78.0 113.98 51.22 78 842 80.0 117.28 52.87 80 870 82.0 120.71 54.58 82 899 84.0 124.28 56.37 84 930 86.0 128.00 58.23 86 962 88.0 131.87 60.16 88 994 90.0 135.90 62.18 90 1,029 92.0 140.09 64.27 92 1,064 94.0 144.45 66.45 94 1,102 96.0 148.99 68.72 96 1,140 98.0 153.72 71.09 98 1,180 100.0 158.63 73.54 100 1,222 102.0 163.75 76.10 102 1,266 104.0 169.08 78.77 104 1,311 106.0 174.63 81.54 106 1,358 108.0 180.40 84.43 108 1,408 110.0 186.40 87.43 110 1,459 112.0 192.66 90.56 1,12 1,512 114.0 199.16 93.81 114 1,567 116.0 205.94 97.20 116 1,625 Revision 0 Page 18 of 22

JAF Pressure-Temperature Limits Report Table 2c: JAF Core Not Critical (Curve B) --- Upper Vessel (All EFPY)

Component = (based on FW nozzle)

ART = 'F ======> All EFPY Vessel Radius, R = inches Nozzle corner thickness, t = inches, approximate Kit = ksi*inchkl KID-applied=

ksi*inch 12 Crack Depth, a = inches Safety Factor =

Temperature Adjustment *F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment = psig (instrument uncertainty)

Reference Pressure = psig (pressure at which the FEA stress coefficients are valid) [GREEN]

Unit Pressure = psig (hydrostatic pressure)

Flange RTNDT =

• F ======> All EFPY Gauge P-T P-T '

Fluid Curve Curve.

Temperature Kic Kip Temperature Pressure

(*F1 (ksi*inch112 )

12 (ksi*inchll 2 ) (°F) (psig) 60.0 79.34 7.01 60 0' 60.0 79.34 17.19 60 276 62.0 81.23 17.86 62 288-64.0 83.19 18.55 64 301 66.0 85.23 19.31 64 313 68.0 87.35 20.37 150 313 70.0 89.56 20.87 150 342 72.0 91.86 21.72 150 357 74.0 94.25 22.62 150 373.

76.0 96.75 23.55 150 389 78.0 99.34 24.53 150 407, 80.0 102.04 25.57 150 425 82.0 104.85 26.65 150 445!

84.0 107.77 27.79 150 465 86.0 110.82 28.98 150 486 88.0 113.98 30.24 150 508 90.0 117.28 31.55 150 532, 92.0 120.71 32.91 150 556 94.0 124.28 34.35 150 582 96.0 128.00 35.86 150 609 98.0 131.87 37.45 150 637 100.0 135.90 39.10 150 666 102.0 140.09 40.84 150 697 104.0 144.45 42.65 150 730 106.0 148.99 44.56 150 763 108.0 153.72 46.55 150 799 110.0 158.63 48.62 150 836 112.0 163.75 50.80 150 875 114.0 169.08 53.08 150 915 116.0 174.63 55.46 150 958

'118.0 180.40 57.96 150 1002 120.0 186.40 60.57 150 1048 122.0 192.66 63.30 150 1097 124.0 199.16 66.15 150 1148 126.0 205.94 69.13 150 1201 128.0 212.99 72.25 150 1256 130.0 220.32 75.50 150 1314 132.0 227.96 78.89 150 1375 134.0 235.91 82.45 150 1438 136.0 244.18 86.16 150 1504 138.0 252.79 90.04 150 1573 140.0 261.75 94.08 150 1645 Revision 0 Page 19 of 22

JAF Pressure-Temperature Limits Report Table 3: JAF Core Critical (Curve C) --- (32 EFPY)

'Cure A Leak Test Temperature = 138.o0 F Curve A Pressure = 1,025.0* psig Unit Pressure = 1,563: psig (hydrostatic pressure)

Flange RTNDT = 30.0 °F Adjusted Adjusted P-T Curve P*T Curve Temperature Pressure 90.00 0 90.00 50 90.00 100 90.00 150 90.00 200 95.17 250 103.92 300 104.00 312 190.00 313 190.00 350 190.00 400 190.00 450 190.00 500 190.00 550 190.00 600 190.00 650 190.00 700 190.00 750 190.00 800 194.36 850 198.52 900 202.36 950 205.94 1000 209.26 1050 212.39 1100 215.33 1150 218.11 1200 220.73 1250 223.23 1300 225.61 1350 227.89 1400 230.06 1450 232.15 1500 234.14 1550 236.07 1600 Revision 0 Page 20 of 22 ,

JAF Pressure-Temperature Limits Report Table 4: JAF ART Calculations for 32 EFPY [6.3]

(NOTE: This table covers all RPV materials with an exposed fluence, E> I MeV, of greaterthan 1.0x10 17 n/cm 2 )

&. No Het~

Cod yp & otChiemistry Chemnistry Adjustments For 1/4t Desritin od N. ea N. lu Tpe& otNo.. Inta TT(F Factor AR'TNT ~Marin Terms ARTINOT Lower Shell #1 G-3415-1R C3394-1 - -10.0 0.11 0.56 73.60 32.8 16.4 0.0 55.6 Lower Shell #3 G-3415-2 C3103-2 - -2.0 0.14 0.57 98.65 43.9 .17.0 0.0 75.9 Lower-nt. Shell #1 G-3413-7 03368-1 -10.0 0.12 0.50 81.00 40.2 17.0 0.0 64.2 Lower-Int. Shell #2 G-3414-2 C3278-2 -10.0 0.11 0.61 89.56 44.4 17.0 0.0 68.4 Lower-Int. Shell #3 G-3414-1 C3301-1 -18.0 0.18 0.57 131.15 65.1 17.0 0.0 81.1 De'scription~ Code. No. Heat No.~ Flux Typo & Lot No. Initial RTNj[DT ('F) Factor ~ARNDT Margin T~erms~ARTrjDT L... nt....Shell........

Lg. W .. 1-233-A.13253/12008.Flux.1092.Lot.3947.-50 0.873 .94. 122.9 .0.0.. 100.9.

L. Int. Shell Long. Weld #1 1-233-A 13253/12008 Flux 1092 Lot 3947 -50.0 0.21 0.873 326.94 122.9 14.0 0.0 100.9 L. Int. Shell Long. Weld #2 1-233-B 13253/12008 Flux 1092 Lot 3947 -50.0 0.21 0.873 326.94 113.2 14.0 0.0 91.2 L. Int./L. Shell Girth Weld 1-240 305414 Flux 1092 Lot 3947 -50.0 0.337 0.609 209.11 95.8 28.0 0.0 101.8 Lower Shell Long. Weld'#2 2-233-B 27204/12008 Flux 1092 Lot 3774 -48.0 0.219 0.996 231.06 84.1 28.0 0.0 92.1 Lower Shell Long. Weld #3 2-233-C 27204/12008 Flux 1092 Lot 3774 -48.0 0.219 0.996 1 231.06 80.6 28.0 0.0 88.6

,,WallThickness ,, (in) Flu.nce at ID Attenuation, 14t tuFnce at 114t Fluence Factor, FF 2

Location Full

.nicm 11 ).e86+18A.6 0. 145o f.15E+16 Lower Shell #1 6.375 1.594 1.686E+18 0.682 1.15E+18 0.445 Lower Shell #2 6.375 1.594 1.686E+18 0.682 1.15E+18 0.445 Lower Shell #3 6.375 1.594 1.686E+1 8 0.682 1.15E+18 0.445 Lower-Int. Shell #1 5.375 1.344 2.01E+18 0.724 1.456E+18 0.496 Lower-Int. Shell #2 5.375 1.344 2.01E+I 8 0.724 1.456E+18 0.496 Lower-Int. Shell #3 5.375 1.344 2.01 E+1 8 0.724 1.456E+18 0.496 5.375 1.344 1.118E+ 18 0.724 8.096E+17 0.376 L. Int. Shell Long. Weld #1 L. Int. Shell Long. Weld #2 5.375 . 1.344 9.49E+17 0.724 6.874E+17 0.346 u L. Int. Shell Long. Weld #3 5.375 1.344 1.118E+18 0.724 8.096E+17 0.376 z L. Int./L. Shell Girth Weld 5.375 1.344 .1.686E+18 0.724 1.221E+18 0.458 Lower Shell Long. Weld #1 5.375 1.344 1.544E+18 0.724 1.119E+18 0.440 Lower Shell Long. Weld #2 5.375 1.344 1.048E+18 0.724 7.592E+17 0.364 Lower Shell Lonq. Weld #3 5.375 1.344 9.632E+17 0.724 6.977E+17 0.349 Page 21 of 22

JAF Pressure-Temperature Limits Report APPENDIX A JAF Reactor Vessel Material Surveillance Programs In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements (Reference 6.11), the first surveillance capsule at JAF was removed at 5.98 EFPY and the second capsule was removed at 13.4 EFPY. Both surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel core beltline region. The flux wires and test specimens removed from the capsules were tested according to the latest version of ASTM E185.

The methods and results of testing are presented in Reference 6.8, as required by 10 CFR 50, Appendices G and H (References 6.10 and 6.11).

JAF has replaced the original RPV material surveillance program with the BWRVIP ISP (Reference 6.12). This program meets the requirements of 10 CFR 50, Appendix H, and has been approved by NRC. Under the ISP, there are no further capsules scheduled for removal from the JAF reactor vessel. Representative surveillance capsule materials for the JAF limiting beltline plate and weld are in another representative plant, with a capsule withdrawal schedule controlled by the BWRVIP ISP.

There are two remaining JAF specimen capsules (the capsule removed in 1996 was re-constituted and returned to the vessel in 1998) which will remain in place to serve as backup surveillance material for the BWRVIP program, or as otherwise needed.

Page 22 of 22