ML100270054

Proposed License Amendment to Technical Specifications: Revised P-T Limit Curves and Relocation of Pressure-Temperatures (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

ML100270054
Person / Time
Site:	Pilgrim
Issue date:	01/24/2010
From:	Bronson K Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.10.005
Download: ML100270054 (62)

Text

{{#Wiki_filter:Entergy Nuclear Operations, Inc. Pilgrim Nuclear Power Station SEntergy 600 Rocky Hill Road Plymouth, MA 02360 Kevin. H. Bronson Site Vice President January 24, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc. Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Proposed License Amendment to Technical Specifications: Revised P-T Limit Curves and Relocation of Pressure-Temperatures (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

REFERENCES:

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

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

3. Pilgrim License Amendment No. 227, Extension of Pressure-Temperature Limits Specified in Technical Specifications (TAC No. MD4093), dated March 26, 2007

4. Structural Integrity Associates Topical Report (TR) SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", April 2007 LETTER NUMBER: 2.10.005

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests an Operating License amendment to the Pilgrim Operating License Technical Specifications to revise the Pressure-Temperature Curves and to include a Pressure and Temperature Limits Report in accordance with NRC Generic Letter 96-03 (Reference 1) and TSTF-419-A (Reference 2). By License Amendment No. 227 (Reference 3), NRC approved the use of current Pressure-Temperature (P-T) Curves through the end of Operating Cycle 18, in order for Entergy to submit and execute a plan to improve plant dosimetry and use of the RAMA calculated fluence values to the end of the anticipated period of the extended license.

Entergy Nuclear Operations, Inc Letter Number 2.10.005 Pilgrim Nuclear Power Station Page 2 of 3 The proposed license amendment uses the RAMA calculated fluence values to the end of the anticipated extended license. Benchmarking of the RAMA Code to the requirements of RG 1.190 was done using Monticello capsule data, and is applicable to Pilgrim, as documented in NRC ADAMS Accession Number ML090370930. The proposed change modifies Technical Specification (TS) Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Administrative Controls Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The PTLR includes revised 34 Effective Full Power Years (EFPY) P-T Curves, neutron fluence, and Adjusted Reference Temperature (ART) values. The PTLR is based on the NRC approved methodology and template provided in SIR-05-044-A (Reference 4). The proposed changes are described in the following attachments:

• Attachment 1 Description and Evaluation of the Proposed TS changes with 2 Enclosures
• Attachment 2 Marked-up pages of the current TS and Bases
• Attachment 3 Retyped TS and Bases pages
• Attachment 4 Pressure and Temperature Limits Report (PTLR)

The proposed Pilgrim PTLR amendment follows the previously approved following precedents:

   "   James A. FitzPatrick, Amendment No. 292, (TAC No. MD8556) dated October 3, 2008.

• Wolf Creek, Amendment No. 180 (TAC No. MD9217), dated January 27, 2009.
• Comanche Peak, Amendment No. 132 (TAC No. MC9500) dated February 22, 2007
• Calloway, Amendment No. 177 (TAC No. MD3053), dated December,5, 2006 Entergy requests NRC approval of the proposed TS amendment by March 1, 2011,4to support the restart from Refueling Outage (RFO)-18, with the amendment being implemented within 60 days from approval.

This letter contains no new regulatory commitments. If you have any questions regarding the subject matter, please contact Joe Lynch at 508 830 8403. I declare under penalty of perjury that the foregoing is true and correct. Executed on the

 .1"*- dayof ,,TN4-,L"                 ,2010.

Sincerely, Kevin Bronson, Pilgrim Site Vice President

Entergy Nuclear Operations, Inc Letter Number 2.10.005 Pilgrim Nuclear Power Station Page 3 of 3 : Description and Evaluation of the Proposed TS Changes (9 pages). : Marked-up pages of the Current TS and Bases (14 pages). : Retyped Proposed TS and Bases pages (9 pages). : Pressure and Temperature Limits Report (PTLR) (23 pages). CC: Mr. James S. Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North O-8C2 11555 Rockville Pike Rockville, MD 20852 Mr. Samuel J. Collins, Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 NRC Resident Inspector Pilgrim Nuclear Power Station Director, Mass Emergency Management Agency 400 Worcester Road Framingham, MA 01702 Director, Radiation Control Program Commonwealth of Massachusetts 90 Washington Street Dorchester, MA 02121

Attachment 1 to Entergy Letter 2.10.005 (9 pages) Description and Evaluation of the Proposed TS Changes Proposed License Amendment to Technical Specifications: Revised P-T Curves and Relocation of Pressure-Temperatures (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

1. DESCRIPTION

2. PROPOSED CHANGES

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES I

1.0 Description The proposed amendment modifies Technical Specifications (TS) Section 1.0, Definitions, Section 3.6, Primary System Boundary, Specification 3.6.A.2, and Section 5.0, Administrative Controls, Section 5.5, Programs and Manuals to include reference to the Pressure and Temperature Limits Report (PTLR). The proposed change adopts the NRC approved methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated April 2007 (References 1 and 3) for preparation of the pressure and temperature curves, in accordance with the, guidance of NRC Generic Letter 96-03, "Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits", dated January 31, 1996 (Reference 2) and incorporates NRC approved TSTF-419-A ("Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR") (Reference 4). The Pilgrim PTLR has been developed based on the methodology and template provided in SIR 044-A, and is enclosed for NRC review and approval. The Entergy proposed Pilgrim License Amendment follows the recently approved precedents, as follows

• James A. FitzPatrick, Amendment No. 292, (TAC No. MD8556) dated October 3, 2008.
• Wolf Creek, Amendment No. 180 (TAC No. MD9217), dated January 27, 2009.
• Comanche Peak, Amendment No. 132 (TAC No. MC9500) dated February 22, 2007,,
• Calloway Plant Unit 1 (TAC No. MD3053) 2.0 Proposed Changes The proposed modifications to the Pilgrim Technical Specifications are as follows (deletions are marked by single strike and insertions are indicated by bold letters):

a. A new Definition is added on Page 1-4, TS Section 1.0, DEFINITIONS, PRESSURE AND The PTLR is the Pilgrim-Specific document that provides TEMPERATURE the reactor vessel Pressure-Temperature (P-T) Curves, LIMITS REPORT including heatup and cooldown rates and fluence and (PTLR) Adjusted Reference Temperature limits for Specification 3.6.A. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.5.9

b. Technical Specification 3.6.A is revised (TS pages 3/4.6-1and 2) as follows:

1. The average rate of reactor coolant temperature change duiring normal heatup or cooldown shall not exceed 100oF/hr when averaged over a ono hou pcriod eXccPt;Yhen thc-vessel tcmperaturs arc*above l.OoF the limit in the PTLR.

2. The reactor vessel shall not be pressurized for hydrostatic and/or leakage tests, and subcritical or critical core operation shall not be conducted unless the reactor vessel temperatures are above those defined by the appropriate curves on Figus 3.6 1, 3.65 2, and 3.6 3. (Lincar ... cp..ati-n betccn cu spermitted).

ps Atsatdpessure, the reactor vcsscl bottom hedmyb vv nn1 ananda epraturcs below those W tcmpcaturs corcspnding to the adjacent rcact()r vessel shall as shoWn 1 of 9

in Figures 3.6 1 and 3.6 2 in the PTLR. In the event this requirement is not met, achieve stable reactor conditions with reactor vessel temperature above that defined by the appropriate curve and obtain an engineering evaluation to determine the appropriate course of action to take.

3. The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head is greater than 560F-the PTLR limit.

4. The pump in an idle loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are within 509P of each othc, the PTLR limit.

5. The reactor recirculation pumps shall not be started unless the coolant temperatures between the dome and the bottom head drain are within 1-4F the PTLR limit.

c. Technical Specification 5.5.9 is inserted as follows:

5.5.9 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 3.6.A

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 reason or supplement thereto.

The proposed TS Section 5.5.9 includes 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 document 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. 2 of 9

3.0 Background. The NRC Safety Evaluation Report (SER) of Structural Integrity Associates Topical Report (TR) SIR-05-044, permits Boiling Water Reactor (BWR) licensees to relocate their pressure-temperature (P-T) curves from the facility Technical Specifications (TS) to a Pressure and Temperature Limits Report (PTLR) utilizing the guidance in Technical Specification Task Force (TSTF) Traveler No. 419-A and in accordance with Generic Letter 96-03 (References 1, 2, 3, and 4). The proposed "Pilgrim Pressure and Temperature Limits Report (PTLR)" is based on the methodology and template provided in the NRC approved document SIR-05-044-A. The Pilgrim PTLR includes operating limits related to Reactor Coolant System (RCS) pressure versus temperature limits during heatup, cooldown and hydrostatic/class 1 leak testing. The curves 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.

4. 0 Technical Analysis 4.1 Pilgrim Fluence and Adjusted Reference Temperature (ART) Calculations, and Revised P-T Curves NRC approved the use of the current 48 EFPY P-T Curves through the end of Operating Cycle 18 by Pilgrim License Amendment No. 227 (Reference 5), in order to execute a plan to resolve RAMA fluence calculation benchmarking issues and to use current NRC approved- methodology to calculate fluence values to the end of extended license (Reference 7).

Previously, new fluence calculations were performed using RAMA methodology, but could not be benchmarked to the results of Pilgrim surveillance capsule data removed during refueling outage (RFO)-4. The RFO-4 Pilgrim surveillance capsule data yielded a Calculated-to-Measured (C/M) ratio that was outside of the tolerance of Regulatory Guide (RG) 1.190. Pilgrim has been a participant in the NRC approved BWRVIP Integrated Surveillance Program as authorized by License Amendment No. 209 (Reference 6). As such, Pilgrim opted ýto use the Monticello Nuclear Power, Plant, a BWR/3 class plant, benchmarking evaluation to produce benchmarked Pilgrim-specific fluence and ART values, and revised P-T curves. EPRI contractor, TransWare has performed their activation / fluence evaluation for the Monticello nuclear power plant capsule removed during their 2008 refueling outage. As a part of the Monticello evaluation, RAMA-calculated activation predictions were compared to measurements for the 3000 ISP capsule flux wires that were irradiated for 23 cycles (28.2 EFPY). Activation comparisons were performed for four copper, four iron, and four nickel flux wires that resided at various axial elevations in the capsule. The Monticello average C/M result for all copper activation measurements is 0.89, for all iron activation measurements is 1.08, and for all nickel activation measurements is 1.13. The resulting average C/M for all measurements in the Monticello capsule is 1.03 +/- 11% (Ia). The comparison results are well within the RG 1.190 guidelines and confirm that the RAMA fluence methodology provides an unbiased estimate of the fluence in the Monticello capsule data. In addition, the results obtained from the Monticello capsule comparisons using the RAMA Fluence Methodology are consistent with the observed comparisons from other BWRs (including BWR/2s, BWR/4s, and BWR/6s). The results of the Monticello flux wire activation analyses indicate that the RAMA fluence methodology does not exhibit a statistically significant bias, -unlike the results of the Pilgrim activation analysis discussed in Reference 5. Thus, from a RAMA Fluence Methodology perspective, the bias observed for the Pilgrim BWR/3 plant for RFO-4 capsule remains unique within the BWRs and can not be explained and will not be used for fluence, ART, or revising Pilgrim P-T curves. 3 of 9

Entergy has determined that Monticello reactor pressure vessel fluence calculation for a BWR/3 provides an acceptable benchmark for Pilgrim fluence data to support revised P-T Curves for Pilgrim Operating Cycle 18 and beyond. This information was discussed with the NRC Staff on or about October 17, 2008. The NRC staff concurred with the Entergy approach to use Monticello fluence for benchmarking Pilgrim RAMA fluence calculations (as documented in NRC ADAMS Accession Number ML090370920) and to submit Pilgrim revised P-T curves for NRC approval. The Monticello reactor vessel fluence calculation provides an acceptable benchmarking for RAMA calculated Pilgrim fluence, ART, and revised P-T curves. Unbiased RAMA methodology calculated fluence and ART data are used to develop the revised P-T curves included in the proposed PTLR. Even though the TransWare calculations for Pilgrim were performed in December 2005, prior to the Monticello benchmarking which was completed in 2008, there have been no material changes in the RAMA Code and Methodology that impacted the results of fluence and ART calculations and revised P-T Curves. Therefore, the unbiased fluence and ART values included in TransWare Report No. ENT-FLU-001-R-001, Rev.0, provides predictable Pilgrim reactor vessel fluence values using RAMA methodology in support of the proposed license amendment (Reference 11). The Pilgrim fluence calculation results show the vessel would experience peak fluence for 54 EFPY of 1.14x10 18 n/cm 2 at weld 1-338A/C location and 1.28x10' 8 n/cm 2 at the lower intermediate shell location, respectively. The N2 nozzle peak fluence at 54 EFPY was calculated to be 2.81xl 017 n/cm 2. Structural Integrity Report No. PNPS-22Q-301, ARTNDT and ART Evaluation, Rev. 1, (Reference

12) provides Pilgrim Adjusted Reference Temperature Calculations for 24, 34, 44, and 54 EFPYs (Table 3, 4, 5, and 6). These Tables show limiting material (plates and welds) characteristics and fluence values calculated at 24, 34, 44, and 54 EFPYs of operation.

Fluence is projected based on actual operation through 20.7 EFPY (April 2005). Fluence at intermediate exposure is linearly interpolated between 20.7 and 54 EFPY. Structural Integrity Report Nos. PNPS-03Q-301, Rev. 2 (Reference 13), Development of Pressure Test (Curve A) and PNPS-03Q-302, Rev. 2 (Reference 14), Development of Heatup/Cooldown (Curves B and C) are based on the requirements of Appendix G to 10 CFR 50 and Appendix G to ASME Section XI and the methodology in NRC approved Structural Integrity Report, SIR-05-044-A. As presented in PNPS-03Q-301, Rev. 2, RPV beltline, bottom head, and feedwater nozzle/upper vessel regions are evaluated. These regions bound all other regions with respect to brittle fracture, including the N2 nozzles through 34 EFPY. In addition, limiting stresses for the bottom head (CRD penetration) region were selected from all applicable design basis transients in order to accommodate any potential inadvertent bottom head cooldown events. The proposed P-T curves for hydrostatic and leak tests, subcritical heatup and cooldown, and critical core operation, were generated with fuel in the vessel for 24, 34, 44, and 54 EFPY operations, however only the curves applicable up to 34 EFPY are included in the PTLR. Pilgrim estimated EFPY will be approximately 26.4 at the end of the current operating cycle (18). 4.2 Pilgrim Pressure and Temperature Limits Report (PTLR) 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 the improved 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 4 of 9

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 are 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. The objective of the Structural Integrity Associates Report provides 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 methodology for developing reactor coolant system (RCS) pressure test, core not critical, and core critical P-T curves for BWRs. As discussed in the NRC Safety Evaluation Report (Section 2.1 of the Reference 1), 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 in-service leak rate testing and/or hydrostatic testing conditions). As discussed in the NRC 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 procedure for calculating P-T 'limits. Appendix A provides guidance for evaluating surveillance data. Appendix B provides a template PTLR. Section 2.0 of the PTLR 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 Attachment 1 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 PTLR's 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 PTLR 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 Pilgrim P-T curves considered three regions of the vessel: beltline, bottom head, and non-beltline (including flange) regions. The Pilgrim Pressure and Temperature Limits Report (P.TLR) based on the methodology and template provided in SIR-05-044-A is being submitted for review and approval. The pressure and temperature curves utilize the methodology of SIR-05-044-A. The PTLR will be a licensee-controlled document, subject to Technical Specification 5.5.9. Based upon the foregoing presentation, Pilgrim PTLR complies with the NRC GL 96-03 and TSTF-419-A and is developed in accordance with the NRC approved methodology. 5 of 9

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"), Specification 3.6.A.2, and revises 5.0 ("Administrative Controls"), to include section 5.5.9 to 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 Pilgrim 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-05-044-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 light water 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. 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. 6 of 9

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 pressure and temperature curves. Therefore, the proposed change does not involve a significant reduction in a margin of.safety. 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"). 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-desighed 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 I Criteria The NRC has established requirements in Appendix G of Part 50 to Title 10 of the Code of Federal Regulations, 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 XI, 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. Table 1 to 10 CFR Part 50, Appendix G provides the NRC staffs criteria for meeting the P-T limit requirements of ASME Code, Section XI, 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 light water reactors. Pilgrim Nuclear Power Plant demonstrates its compliance with the 'Appendix H through participation in the BWRVIP Integrated Surveillance Program (ISP) (Reference 6). In March 2001, the NRC staff issued RG 1.190, "Calculation 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. 7 of 9

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. 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 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. 7.0 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 for Boiling 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.

8 of 9

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. Pilgrim License Amendment No. 227, Extension of Pressure-Temperature Limits Specified in Technical Specifications (TAC No. MD4093), dated March 26, 2007

6. Pilgrim License Amendment No. 209, Revision to the Vessel Material Surveillance Program, (TAC No. MC1565), dated January 5, 2005

7. Entergy Letter No. 2. 07.078, License Renewal Application Commitment 47 Response, September 12, 2007

8. Entergy Letter No. 2.07.006, Proposed Change to the Applicability of Pilgrim's Pressure-Temperature Curves as described in Technical Specification Figures 3.6.1, 3.6.2, and 3.6.3, Revision 1, January 15, 2007

9. Entergy Letter No. 2.06.018, Proposed License Amendment to Change Technical Specification 3.6.A.2, Pressure Temperature Limits Curves, April 12, 2007

10. Entergy Letter No. 2.06.090, Pilgrim Response to NRC Request for Additional Information Related to Proposed License Amendment to Change P-T Curves (TAC No.

MD1218), dated October 16, 2006

11. TransWare Report No. ENT-FLU-001 -R-001, "Pilgrim Nuclear Power Station Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 and 54 EFPY", Rev.0, submitted by Entergy Letter No. 2.06.090, dated October 16, 2006.

12. Structural Integrity Report, File No. PNPS-22Q-301, "ARTNDT and ART Evaluation",

Rev. 1, dated November 17, 2009 (34 pages)

13. Structural Integrity Report, File No. PNPS-03Q-301, "Development of Pressure Test (Curve A) P-T Curves", Rev. 2, November 23, 2009 (37 pages)[Proprietary Document]

14. Structural Integrity Report File, No. PNPS-03Q-302, "Development of Heatup/ Cooldown (Curves B & C) P-T Curves", Rev.2, dated November 30, 2009 (55 pages) [Proprietary Document]

9 of 9

            )

Attachment 2 to Entergy Letter No. 2.10.005 Marked-up Pages of Current Technical Specifications and Bases (14 Marked-up Pages)

1. TS page 1-4

2. INSERT to TS page 1-4

3. License Page 3

4. TS page 3/4.6-1

5. TS page 3/4.6-2

6. TS page 3/4.6-9

7. TS page 3/4.6-10

8. TS page 3/4.6-11

9. TS BASES page B3/4.6-1

10. TS BASES page B3/4.6-2

11. TS BASES page B3/4.6-3

12. INSERT "A"4.6.A BASES

13. TS Page 5.0-11

14. INSERT "SPECIFICATION 5.5.9"

1.0 DEFINITIONS (Cont) OPERABLE - A system, subsystem, division, :omponent, or device shall be OPERABILITY OPERABLE or have OPERABILITY when it is capable of 4w performing its specified safety function(s) and when all necessary attendant instrumentation, coitrols, 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). OPERATING OPERATING means that a system or component is performing its intended functions in its required manner. OPERATING CYCLE Interval between the end of one refueling outage and the end of the next subsequent refueling outage. PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY means that the drywell INTEGRITY and pressure suppression chamber are intact and all of the following conditions are satisfied: I

1. All manue, containment isolation valves on lines AM 44fLC connected to the reactor coolant system or containment which are not required to be open during accident

    .-fft                               conditions are closed.

I elm VT~- 2. At least one door in each airlock is closed and sealed

3. All blind flanges and manways are closed.

lop o~e

4. All automatic primary containment isolation valves and all instrument line check valves are operable or at least one containment isolation valve in each line having an inoperable valve shall be deactivated in the isolated condition.

5. All containment isolation check valves are operable or at least one containment valve in each line having an inoperable valve is secured in the isolated position.

PROTECTIVE ACTION An action initiated by the protection system when a limit is reached. A PROTECTIVE ACTION can be at a channel or system level. PROTECTIVE FUNCTION A system PROTECTIVE ACTION which results from the PROTECTIVE ACTION of the channels monitoring a particular plant condition. REACTOR POWER REACTOR POWER OPERATION is any operation with the mode OPERATION switch in the "Startup" or "Run" position with the reactor critical and above 1% design power. REACTOR VESSEL Unless otherwise indicated, REACTOR VESSEL PRESSURES PRESSURE listed in the Techni=,a, Specifications are those measured by the 4W reactor vessei ste.--n space detectors. PNPS 1-4 Amendment No:, ' -

INSERT DEFINITION FOR "PTLR" The PTLR is the Pilgrim-Specific document that provides the reactor vessel Pressure-Temperature (P-T) Curves, including heatup and cooldown rates and fluence and Adjusted Reference Temperature limits for Specification 3.6.A. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.5.9

B. Technical Specifications . . , . - -- -, The Technical Specifipations contained in Appendix A, as revised through Amendment No.24j are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. Records ENO shall keep facility operating records in accordance with the requirements of the Technical Specifications. D. Equalizer Valve Restriction - DELETED E. Recirculation Loop Inoperable - DELETED F. Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated December 21, 1978 as supplemented subject to the following provision: ENO may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. G. Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Pilgrim Nuclear Power Station Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0" submitted by letter dated October 13, 2004, as supplemented by letter dated May 15, 2006. Amendment 226, 226, 2, 228. 220. 24O, 20404, 2 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Applicability: Applicabilit : W Applies to the operating status of the reactor Applies to the periodic examination and coolant system. testing requirements for the reactor cooling system. Oblective: Obiective: To assure the integrity and safe operation of To determine the condition of the reactor the reactor coolant system. coolant system and the operation of the safety devices related to it Secification: Specification: A. Thermal and Pressurization Limitations A. Thermal and Pressurization Limitations fi~

                                                                   ..  .....               i     I       II Ii * - .

1. The average rate of reactor coolant 1. During heatups and cooldowns.

temperature change during normal heatup with the reactor vessel temperature or cooldown shall not exceed J1# less than or equal to 460F, verify the RCS heatup and cooldown rates are within limits every 15

             ' V

• P4~ 8 R. 9L 4 minutes.

2. The reactor vessel shall not be 2. Reactor vessel shell temperatures, including pressurized for hydrostatic and/or leakage reactor vessel bottom head, and reactor tests, and subcritical or critical core coolant pressure shall be permanently operation shall not be conducted unless logged at least every 15 minutes whenever the reactor vessel temperatures are the shell temperature Is below 220OF and above those defined by the appropriate the reactor vessel is notvented.

curves ...-Fk .. . ... . $ w Revision 201 Amendment No. 3/4.6-1

LIMmNG CONDITON FOR OPERATION SURVILLANCE REQUIREMENTS 3.8 PRIMARY SYSTEM BOUNDARY (Cont) 4.8 PRIMARY SYSTEM BOUNDARY (Coan) A. Thermal and Pressurization Umitations A. Thermal and Pressurization Limitations (Cont) (Cont) in the event this requirement is not met, achieve stable reactor conditions with reactor vessel temperature above that defined by the appropriate curve and obtain an engineering evaluation to determine the appropriate course of action to take.

3. The reactor vessel head bolting studs 3. When the reactor vessel head bolting shall not be under tension unless the studs are tensioned and the reactor Is In temperature of the vessel head flange a Cold Condition, the reactor vessel shed and the head is tar **jer' temperature immediately below the head flange shall be permanently recorded.

A)

4. The pump In an Idle recirculation loop 4. Prior to anddurlng startup of an Idle shall not be started unless the recirculation loop, the temperature of the temperatures of the coolant within the reactor coolant In the operating and idle idle and operating recirculation Iopps are loops shall be permanently logged.

I within , P it

5. The reactor recirculation pumps shall not be started unless the coolant temperatures between the dome and the

5. Prior to starting a recirculatlon pump, the reactor coolant temperatures In the dome and In the bottom head drain shall be bottom head drain are within 14674,-, compared and permanently logged.

Am endm ent No. 27, -22, 42,82, 1 40,2 40, g4" 3/4.8-2

PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE UMITS HYDROSTATIC AND LEAK TESTS 1 I I - - ý I 1 [ I I * ] [ t -t--l l l l 1-r f J

                                                                                                                                                     -.. t --

I 1-I 1,200 o 21] SOTTOM 12 J1~ 2'2 1,100. Ii I I 1,000 900 800

                ~~1~~~~

t,-- Lu~. P.-- ) IiN 700 w 7. i

                                '§
                                                                                                                                                /

Lu 4600 12 (A, -i If 500 400- 7z~ FIGURE 3.6.1 300 Pilgrim Reactor Pressure Vessel Press Temperature Limits for Hydrostatic Leak Tess 200 The curve applies through Operating Cycle1 100 0 50 60 70 80 90

                                            -lit-it
                                                   , '     i T

• l ,, l ' t 71i~ 771.1. -

100 110 120 .130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 iH '4-{i VI 3~.sj+;.{7S~477 TEMPERATURE (-F) Amendment No. -28,82, 94,14,.4na - 3/4.6-9

PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS SUBCRITICAL HEATUP AND COOLDOWN K- S SS * - S- S .n.u-~ - - S- S - - S- S- S- S- S - S- S- P-

                                                                       --4.-

1,200 74,7 §47.711 ~ - N S. 1 .. -

    .1,100.

1,000-t - I. H.. BOTTOM HEAD

                                                                                     .1                                                          .-.. ~<L
                                      #I..
                                                                                                         -177 J7177                                                                                                  I I                                                                                                         -~1----~

4z:.z; ~

                                                                                  -it 9001--:

• 77.7. 17 --. - 1 i

                                                                                                                                          ----..                                                                 U-'

800B 1~~

    ~700-                   7-w                                                          L....

• 600 - .- - -.----..-.--.....-

a: 500 117.1 .1 A -....---.---..- S-.-.-

                                                                                                                                                   ~
                                                                                                                                                                                       ~
                                                                                                                                                                                       .7.1.   .                      I.

400 ~~7.!7477 - - I - - __ FIGiOR 3.6.2

                                         -     *-                   - I       I -  --  -

300- - - , ~- - - - ~- - Pilgrim Reactor Pressure Ve. PrureTemperature Limits for Subcritical Heatuorond Cooldown 200 The curve applies through Opertmg Cycle 18 I 1001 0 I 1 9- i i 7 i I " I i i I i~LiWt-itj7i~ '~JLj TKFYTT .~

                                                                                                                                                                                                                  !    I 50       60   70       80       90     100 110 120 130-140              150 160 170 180                   190 200 210 220 230 240 250 260'0 N TEMPERATURE ('F)

Amendment No. 23/4 I,AA, -1 y 3/4.6-10

"\

N K PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS CRITICAL CORE OPERATION 1,200 (NrN~ 1,100 (nfl 1 nnn i7 ... [7.......... .... -.. ~.. ......... -I I-. ... .. .... t. .. ...- -~.~... 800 ~ ~~1--- ----- . - . 0 700

                                                                  ...     --        -        -

• T....r7 7.... - -

ILU 600 uJS50O a-400-300-200- I ~FIGURE 3.6.31

                                              .. . ...---                                               Pilgrim Reactor Pressure Vessel Pressure&emperature 100-                                                                                                       Limits for Critical Core Operatio T-The                                                                             curve applies through OperatingCyl 0                                                                                                                                                               I 50 60   70     80    90     100 110 120                       130   140 150             160    170 180 190 200         210 220   230  240 250 260 27Q TEMPERATURE ('F)                            "

Amendment No. 4-40,19i~ 3/4.6-11

INSERT "A" 4.6.A BASES The Pressure and Temperature Limits Report (PTLR) provides Pressure-Temperature Curves for heatup and cooldown for the reactor vessel. This report is developed in compliance with NRC approved methodology prescribed in Structural Integrity Associates Topical Report (TR) SIR-05-044-A, "Pressure -Temperature Report Methodology for Boiling Water Reactors", April 2007. The PTLR provides Pressure Test Curve A, and Heatup and Cooldown Curves B and C for 34 EFPY operation. These curves are developed in compliance with the requirements of Appendix G to 10 CFR 50 and Appendix G to ASME Section XI and the methodology in NRC approved Structural Integrity Report, SIR-05-044-A. Pilgrim reactor pressure vessel beltline, bottom head, and feedwater nozzle/upper vessel regions are evaluated. These regions bound all other regions with respect to brittle fracture, including the N2 nozzles through 34 EFPY. In addition, limiting stresses for the bottom head (CRD penetration) region were selected from all applicable design basis transients in order to accommodate any potential inadvertent bottom head cooldown events. The P-T curves for hydrostatic and leak tests, subcritical heatup and cooldown, and critical core operation, were generated with fuel in the vessel for 24, 34, 44, and 54 EFPY operations, however only the curves applicable up to 34 EFPY are included in the PTLR. Pilgrim Adjusted Reference Temperature Calculations for 24, 34, 44, and 54 EFPYs (Table 3, 4, 5, and 6). These Tables show limiting material (plates and welds) characteristics and fluence values calculated at 24, 34, 44, and 54 EFPYs of operation. Fluence is projected- based on actual operation through 20.7 EFPY (April 2005). Fluence at intermediate exposure is linearly interpolated between 20.7 and 54 EFPY. The allowable rate of heatup and cooldown for the reactor vessel contained fluid is 100'F per hour averaged over a period of one hour. This rate has been chosen based on past experience with operating power plants. The associated time periods for heatup and cooldown cycles when the 100°F per hour rate is limiting, provides for efficient, but safe, plant operation. Specific analyses were made based on a heating and cooling rate of 100°F/hour applied continuously over a temperature range of 100OF to 5460F. Calculated stresses were within ASME Boiler and Pressure Vessel Code Section III stress intensity and fatigue limits even at the flange area where maximum stress occurs. The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow. This colder water is forced up when recirculation pumps are started. This will not result is stresses which exceed ASME Boiler and Pressure Vessel Code, Section III limits when the temperature differential is not greater than 145 0 F. The reactor coolant system is a primary barrier against the release of fission products to the environs. In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating donditions to which it can be subjected. Appendix G to 10CFR50 defines the temperature-pressurization restrictions for hydrostatic and leak tests, pressurization, and critical operation. These limits have been calculated for Pilgrim and are contained in the PTLR.

BASES 3/4.6 PRIMARY SYSTEM BOUNDARY A. Thermal and Pressurization Limitations

             -' e allowable rate of heatup and cooldown for the reactor vessel contained fluid is 100 per h ,r averaged over a period of one hour. This rate has been chosen based on past perience with_, perating power plants. The associated time periods for heatup and coold                  n cycles when e 1000F per hour rate is limiting provides for efficient, but safe, plant a ration.

0 Specific an ses were made based on a heating and cooling rate of F/hour applied continuously r a temperature range of 1000 F to 5460 F. Calculat stresses were within ASME Boiler and Pressu Vessel Code Section III stress intensity and fail e limits even at the flange area where maxim stress occurs. The manufacturer perfor ed detailed stress analysis as own in Amendment 17 of the FSAR. This analysis includes more vere thermal condition an those which would be encountered during normal heating and co ling operations. The coolant in the bottom of the ye el isa lower temperature than that in the upper regions of the vessel when there is no recircul flow. This colder water is forced up when recirculation pumps are started. This ill t result is stresses-which exceed ASME Boiler and Pressure Vessel Code, Section Ill Ii s when temperature differential is not greater than 1450 F The reactor coolant syst is a primary barrier agains e release of fission products to the environs. In order to ovide assurance that this barrier is aintained at a high degree of integrity, restrictio ave been placed on the operating co itions to which it can be subjected, M0Appen G to I0CFR50 defines the temperature-pressurization restri ons for hydrostatic and lea sts, pressurization, and critical operation. These limits have been alculated for Pilgrim dare contained in Figures 3.6-1, 3.6-2, and 3.6-3. Revision 2@MV...... . B3..... (\mendment No. -28, 175 B3/4.6-

BASES: 3/4.6 PRIMARY SYSTEM BOUNDARY (Cont) A.. ermal and Pressurization Limitations (Cont) The ttom head, defined as the spherical portion of the reactor vessel located below e lower c uumferential weld, was also evaluated. Reference transition temperatures Tndt) were deve ed for the bottom head and the resulting pressure vs. temperature rves plotted on Fi res 3.6-1 and 3.6-2. It has been determined that the bottom h d temperatures a allowed to lag the vessel shell temperatures (

Reference:

ructural Integrity Associat (SIA) Report SIR-00-108, dated September 11, 200/. The referenced analysis utilizes the ess results established in the Combustion Eng~ eering Inc., Pilgrim Reactor Vessel Design eport, No. CENC 1139, dated 1971, and mbines the stress analysis results, specific t he bottom head, with the pressuriza&in temperatures necessary to maintain fracture toughne requirements in accordance wi the ASME Boiler and Pressure Vessel Code, Section I, the criteria of 10CFR50, ppendix G, and the supplementary guidelines of Reg. uide 1.99, Rev. 2. For Pilgrim pressure-temperature restri ions, two Ioc ions in the reactor YVessel are limiting. The closure region controls at Io r press es and the beltline controls at higher pressures. The nil-ductility transition (NDT) temperaturei de i ed as the temperature below which ferritic steel breaks in a brittle rather than d tile ma r. Radiation exposure from fast neutrons (>1 Mev) above about 1017 nvt ay shift the T temperature of the vessel metal above the initial value. Impact tests fro the first material urveillance capsule removed at 4.17 EFPY indicated a maximum RT t shift of 55 degrees r the-weld specimens. The RTbdt of the closure regia is +10 degrees F. The initial RTd for e beltline weld and base metal are -48 degree Fand 0 degrees F, respectively. These R temperatures are basedbtupon unirradiated st data, adjusted for specimen orientation in acc dance with USNRC Branch Tecaht al Position MTEB 5-2. The closure and b am head regions are not exposed to neutron fluence (> ev) over the vessellI sufficient to cause a shift in RTd, The pressure-temperature Ii~ ations (Figures 3.6-1 3.6-2, and 3.6-3) of the closure and bottom head regions will therefore remain con ant throughout vessel life. Only the beltline region of the reactor vessel wil experien a-shift in RTnd with a resultant increase in pressure-temperature limits. The urves apply to 100% bolt preload condition but are conservative far lesser bolt read conditions. Revision 96, 1-5-, 1-7-, 2-*, B3/4.6-2

BASES: 3/4.6 PRIMARY SYSTEM BOUNDARY (Cont) A. Thermal and Pressurization Limitations (Cont) or critical core operation when the water level is within the normal range for power op ration and the pressure is less than 20% of the preservice system hydrostati est pres ure (313 psi), the minimum permissible temperature of the highly stres regions

         ,of the osure flange is RTnft + 60 degrees F = 70 degrees F; thus, a cutol mit of 70 degrees was chosen as shown on Figure 3.6-3 and as required by CFR50 Appendix G, paragra, IV.A.3. This same cutoff is conservatively included/* the limits for hydrostatic an leak tests and for non-critical operation, as sh n on Figures 3.6-1 and 3.6-2, respective in order to be consistent with the limits           critical operation.

The closure region is re limiting than the feedwat nozzle with respect to both stress intensity and RT n. Ther re, the pressure-temp ature limits of the closure are controlling. The minimum bolt-up temperature is inimum 'allowable nil ductility transition temperature (RTe ss) at pressures belo t 0% of pre-operational system hydrostatic test pressure that bolt pre-load str s cn be lied to the reactor vessel closure region. It his defined as the initi Tr of thehige tressed component of the reactor vessel that includes the yes I head, head flange and s 11adjacent to the vessel flange. The mxmm Rl ndt is + 10 degrees F. For conse Sma minimum bolt-up temperature of es F is chosen because this temperature ides sufficient margin between th owest service temperature at 20% of the pre-ope ona hydrostatic test ssure prior to pressurization. The adjust reference temperature shift is based on Regulatory Guide 1.99, Revi . n 2, dated M 1988; the analytical results of General Electric Report MOE 277-1285, Revisi 1, dated January 21, 1985, regarding projected neutron fluence; and tructural Inte ity Associates (SIA) report SIR-00-108 dated September 11, 2000, for Rnt~versus fa nce as a function of temperature and pressure. B. Coolant Chemistry The reactor vessel coolant chemistry requirements are discussed in Subsection 4.2 of the FSAR. A radioactivity concentration of 20 g Ci/ml total iodine can be reached if there is significant fuel failure or if there is a failure or a prolonged shutdown of the cleanup demnineralizer. Calculations performed, by the AEC staff for this activity level results in a radiological dose at the site boundary of 8 rem to the thyroid from a postulated rupture of a main steam line assuming a 5 second valve closing time and a coolant inventory release of 3 x 104 lbs. A reactor sample will be used to assure that the limit of Specification 3.6.B,. 1 is not exceeded. .. mendment 82- 44.-, 190 B3/4.6-3

Programs and Manuals 5.5 5.5 Programs and Manuals "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii)assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the MCREC System, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. Each Surveillance Requirement shall be performed within the specified SURVEILLANCE INTERVAL with a maximum allowable extension not to exceed 25 percent of the specified SURVEILLANCE INTERVAL. The SURVEILLANCE INTERVAL requirement is applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

AmendmentýNo._3 Amendr !o~/ 5.0-11

INSERT "SPECIFICATION 5.5.9" 5.5.9 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 3.6.A

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 periodand for any reason or supplement thereto.

Attachment 3 to Enterq¥ Letter No. 2.10.005 .Retyped Pages of Proposed Technical Specifications and Bases (9 Retyped Pages) TS Page 1-4 License Page 3 TS Page 3/4.6-1 TS Page 3/4.6-2 TS Page 3/4.6-9 TS Page 3/4.6-10 TS Page 3/4.6-11 TS BASES Page B3/4.6-1 TS Page 5.0.12

1.0 DEFINITIONS (Cont) OPERABLE - A system, subsystem, division, component, or device shall be OPERABILITY 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). OPERATING OPERATING means that a system or component is performing its intended functions in its required manner. OPERATING CYCLE Interval between the end of one refueling outage and the end of the next subsequent refueling outage. PRESSURE AND The PTLR is the Pilgrim-Specific document that provides the TEMPERATURE LIMITS reactor vessel Pressure-Temperature (P-T) Curves, including heat REPORT (PTLR) up and cool down rates and fluence and Adjusted Reference Temperature limits for Specification 3.6.A. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.5.9. PRIMARY PRIMARY CONTAINMENT INTEGRITY means that the drywell CONTAINMENT and pressure suppression chamber are intact and all of the INTEGRITY following conditions are satisfied:

1. All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

2. At least one door in each airlock is closed and sealed

3. All blind flanges and manways are closed.

4. All automatic primary containment isolation valves and all instrument line check valves are operable or at least one containment isolation valve in each line having an inoperable yalve shall be deactivated in the isolated condition.

5. All containment isolation check valves are operable or at least one containment valve in each line having an inoperable valve is secured in the isolated position.

PROTECTIVE ACTION An action initiated by the protection system when a limit is reached. A PROTECTIVE ACTION can be at a channel or system level. PROTECTIVE FUNCTION A system PROTECTIVE ACTION which results from the PROTECTIVE ACTION of the channels monitoring a particular plant condition. PNPS 1-4 Amendment No. 1-77,

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. C. Records ENO shall keep facility operating records in accordance with the requirements of the Technical Specifications. D. Equalizer Valve Restriction - DELETED E. Recirculation Loop Inoperable - DELETED F. Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated December 21, 1978 as supplemented subject to the following provision: ENO may make changes to the approved fire protection program without prior approval of the Commission only ifthose changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. G. Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Pilgrim Nuclear Power Station Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0" submitted by letter dated October 13, 2004, as supplemented by letter dated May 15, 2006. Revision ý279. Amendment 225, 226, 227, 228, 2-., 230, 234, 232, 233,

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Applicability: Applicability: Applies to the operating status of the reactor Applies to the periodic examination and coolant system. testing requirements for the reactor cooling system. Obiective: Obiective: To assure the integrity and safe operation of the reactor coolant system. To determine the condition of the reactor coolant system and the operation of the safety devices related to it. Specification: Specification A. Thermal and Pressurization Limitations A. Thermal and Pressurization Limitations

1. The average rate of reactor coolant temperature change during normal heatup 1. During heatups and cooldowns, with or cooldown shall not exceed the limit in the the reactor vessel temperature less PTIR. than or equal to 450'F, verify the RCS

2. heatup and cooldown rates are within The reactor vessel shall not be limits every 15 minutes.

pressurized for hydrostatic and/or leakage tests, and subcritical or critical core operation shall not be conducted unless 2. Reactor vessel shell temperatures, the reactor vessel temperatures are above including reactor vessel bottom head, and those defined by the appropriate curves in reactor coolant pressure shall be the PTLR. permanently logged at least every 15 minutes whenever the shell temperature is below 220°F and the reactor vessel is not vented. Revision 201 Amendment No. 28, 82, 140, 153, 4-7-4, 3/4.6-1

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SYSTEM BOUNDARY (Cont) A. Thermal and Pressurization Limitations A. Thermal and Pressurization Limitations (Cont) (Cont) In the event this requirement is not met, achieve stable reactor conditions with reactor vessel temperature above that defined by the appropriate curve and obtain an engineering evaluation to determine the appropriate course of action to take.

3. The reactor vessel head bolting studs 3. When the reactor vessel head bolting shall not be under tension unless the studs are tensioned and the reactor is in temperature of the vessel head flange a Cold Condition, the reactor vessel shell I limit.

nd the head is greater than the PTLR temperature immediately below the head flange shall be permanently recorded.

4. Prior to and during startup of an idle

4. The pump in an idle recirculation loop recirculation loop, the temperature of the shall not be started unless the reactor coolant in the operating and idle temperatures of the coolant within the loops shall be permanently logged.

idle and operating recirculation loops are within the PTLR limits.

5. Prior to starting a recirculation pump, the

5. The reactor recirculation pumps shall not reactor coolant temperatures in the be started unless the coolant dome and in the bottom head drain shall temperatures between the dome and the be compared and permanently logged.

bottom head drain are within the PTLR limits. Amendment No. 27, 28, 42, 82, 140, 209, 2--3 3/4.6-2

N Pilgrim Reactor Vessel Pressure-Temperature Limits Hydrostatic and Leak Rates is Relocated to PTLR and TS 5.5.9 Amendment No. 29, 82, 94, 140,4972, 217, 3/4.6-9

Pilgrim Reactor Vessel Pressure-Temperature Limits Subcritical Heat up and Cool down is Relocated to PTLR and TS 5.5.9 Amendment No. 28, *2,94, 140, 197, 22_7, 314.6-10

Pilgrim Reactor Vessel Pressure-Temperature Limits Critical Core Operation is Relocated to PTLR and TS 5.5.9 Amendment No. 28, 8.2, 91, 140, 19,227, 3/4,6-1I1

BASES: 3/4.6 PRIMARY SYSTEM BOUNDARY A. Thermal and Pressurization Limitations The Pressure and Temperature Limits Report (PTLR) provides Pressure-Temperature Curves for heatup and cooldown for the reactor vessel. This report is developed in compliance with NRC approved methodology prescribed in Structural Integrity Associates Topical Report (TR) SIR-05-044-A, "Pressure -Temperature Report Methodology for Boiling Water Reactors", April 2007. The PTLR provides Pressure Test Curve A, and Heatup and Cooldown Curves B and C for 34 EFPY operation. These curves are developed in compliance with the requirements of Appendix G to 10 CFR 50 and Appendix G to ASME Section Xl and the methodology in NRC approved Structural Integrity Report, SIR-05-044-A. Pilgrim reactor pressure vessel beltline, bottom head, and feedwater nozzle/upper vessel regions are evaluated. These regions bound all other regions with respect to brittle fracture, including the N2 nozzles through 34 EFPY. In addition, limiting stresses for the bottom head (CRD penetration) region were selected from all applicable design basis transients in order to accommodate any potential inadvertent bottom head cooldown events. The PJT curves for hydrostatic and leak tests, subcritical heatup and cooldown, and critical core operation, were generated with fuel in the vessel for 24, 34, 44, and 54 EFPY operations, however only the curves applicable up to 34 EFPY are included in the PTLR. Pilgrim Adjusted Reference Temperature Calculations were performed for 24, 34, 44, and 54 EFPYs (Tables 3 to 6). These Tables show limiting material (plates and welds) characteristics and fluence values calculated at 24, 34, 44, and 54 EFPYs of operation. Fluence is projected based on actual operation through 20.7 EFPY (April 2005). Fluence at intermediate exposure is linearly interpolated between 20.7 and 54 EFPY. The allowable rate of heatup and cooldown for the reactor vessel contained fluid is 100°F per hour averaged over a period of one hour. This rate has been chosen based on past experience with operating power plants. The associated time periods for heatup and cooldown cycles when the 100°F per hour rate is limiting provides for efficient, but safe, plant operation. Specific analyses were made based on a heating and cooling rate of 100 0 F/hour applied continuously over a temperature range of 100°F to 5460 F. Calculated stresses were within ASME Boiler and Pressure Vessel Code Section III stress intensity and fatigue limits even at the flange area where maximum stress occurs. The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow. This colder water is forced up when recirculation pumps are started. This will not result is stresses which exceed ASME Boiler and Pressure Vessel Code, Section III limits when the temperature differential is not greater than 1450F. The reactor coolant system is a primary barrier against the release of fission products to the environs. In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected. Appendix G to 10CFR50 defines the temperature-pressurization restrictions for hydrostatic and leak tests, pressurization, and critical operation. These limits have been calculated for Pilgrim and are contained in the PTLR. Revision 2-0-, B3/4.6-1

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

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

i) Limiting conditions for Operation Section 3.6.A.2 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 reason or supplement thereto.

(continued) Amendment No. 2S4, 5.0-12

Attachment 4 to Enterqy Letter No. 2.10.005 Pilgrim "Pressure-Temperature Limits Report" (PTLR) (23 Pages)

ENTERGY NUCLEAR OPERATIONS, INC. PILGRIM NUCLEAR POWER STATION Pilgrim Pressure and Temperature Limits Report (PTLR) For 34 Effective Full Power Years (EFPYs) Revision 0 Prepared by: l* IC)"e..- OP* YV- Date: 1 -1 /q vs Reviewed by:

• V "/// Date: /2/2///0

, Approved by: Date: Diret'r of Engineering Date: I/I°°/°2010 Concurred by: 4Manager of Liensing W 6-Z/IIJV

Pilgrim Pressure-Temperature Limits Report Revision 0 1.0 PURPOSE 2.0 APPLICABILITY 3.0 METHODOLOGY 4.0 OPERATING LIMITS 5.0 DISCUSSION 6.0 PRESSURE-TEMPERATURE CURVES Fig. 6.1, Curve A: PNPS Pressure Test Curve, 34 EFPY Fig. 6.2, Curve B: PNPS Heatup/Cooldown, Core Not Critical Curve, 34 EFPY Fig. 6.3, Curve C: PNPS Heatup/Cooldown, Core Critical Curve, 34 EFPY 7.0 TABULATED PRESSURE-TEMPERATURE VALUES FOR 34 EFPY Table 7.1 Values for Beltline Pressure Test Curve (Curve A), 34 EFPY Table 7.2 Values for Upper Vessel region Pressure test Curve A Table 7.3 Values for Bottom Head Pressure Test Curve (Curve A) Table 7.4 Values for Beltline Core Not Critical Curve (Curve B), 34 EFPY Table 7.5 Values for Upper Vessel region Core Not Critical Curve B Table 7.6 Values for Bottom Head Core Not Critical Curve (Curve B) Table 7.7 Values for Core Critical Curve (Curve C), 34 EFPY Table 7.8 PNPS ART Calculations for 34 EFPY 8.0 PILGRIM REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM

9.0 REFERENCES

Pilgrim Pressure-Temperature Limits Report Revision 0 1.0 PURPOSE The purpose of Pilgrim 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 I Leak Testing
• RCS Heatup and Cooldown Rates

           "  Reactor Pressure Vessel (RPV) to RCS coolait 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 of Technical Specification 5.5.9, "Reactor Coolant System (RCS) Pressure and Temperature Limits Reports (PTLR)."

2.0 APPLICABILITY The current revision of this report is applicable to the Pilgrim RPV for 34 Effective Full Power Years (EFPY) operations. Technical Specification 3.6.A is affected by the information contained in this report. The Pilgrim Pressure and Temperature Limits for 34 to 54 EFPY have been developed using References 9.4 and 9.11, but only the 34 EFPY limits are incorporated in this revision of the report. Future revisions to this report must be completed in accordance with the requirements of 10 CFR 50.59 review process, as applicable. 3.0 METHODOLOGY The limits in this report are derived as follows: 3.1 The methodology used is in accordance with Reference 9.1, which has been approved for BWR use by the NRC. 3.2 The neutron fluence is calculated in accordance with NRC Regulatory Guide (RG) 1.190 (Reference 9.5) using the RAMA Computer Code, as documented in Reference 9.2. 3.3 Consistent with Section 2.3 and Appendix A of Reference 9.1, the Adjusted Reference Temperature (ART) values for the beltline materials are calculated in accordance with NRC RG 1.99, Revision 2 (Reference 9.6) and supplemented by data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) in Reference 9.8, as documented in Reference 9.3. Pilgrim Pressure-Temperature Limits Report ,Revision 0 3.4 This revision of the Pressure and Temperature Limits Report is to incorporate the initial issue of the PTLR for 34 EFPY: The changes to the P-T curves, limits, or parameters within this PTLR are based upon new irradiation fluence data of the RPV, new material data from the ISP, and other plant design information contained in the Updated Final Safety Analysis Report (UFSAR). Changes may be made in accordance with 10 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance. Any revisions to fluence calculation methodology shall be submitted to the NRC for review and approval prior to incorporation of the fluence results into this PTLR. 4.0 OPERATING LIMITS The Pressure-Temperature (P-T) Curves included in this report represent reactor vessel steam dome pressure versus minimum reactor vessel coolant temperature, and incorporate 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 as Curve A; b) Core Not Critical Heatup/Cooldown Operation, referred to as Curve B, and c) Core Critical Heatup/Cooldown Operation, referred to as Curve C, in accordance with Reference 9.13. Temperature limits applicable to the Reactor Pressure Vessel are: Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing (Figure 6.1: Curve A): _ 250F/hourl. Normal Operating Heatup and Cooldown rate limit (Figure 6.2: Curve B - non-nuclear heating, and Figure 6.3: Curve C - nuclear heating): _<100°F/hour2. RPV bottom head coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: _ 145 OF. Recirculation loop coolant temperature to RPV coolant temperature AT limit during Recirculation Pump startup: :5 50 0 F. RPV head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head is (Figure 6.1: Curve A - Hydrostatic and Class 1 Leak Testing; Figure 6.2: Curve B - non-nuclear heating): _>600 F. RPV flange criticality temperature limit (Figure 6.3: Curve C - nuclear heating): > 70 0 F. Interpreted as the temperature change in any 1-hour period is less than or equal to 25TF. 2 Interpreted as the temperature change in any 1-hour period is less than or equal'to 100TF. Pilgrim Pressure-Temperature Limits Report Revision 0 5.0 DISCUSSION, The ART of the limiting beltline material is used to adjust the beltline P-T Curves to account for irradiated effects. RG 1.99, Revision 2 (Reference 9.6) provides the methods for determining the ART. The Pilgrim reactor vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of Pilgrim vessel plate and weld materials (Reference 9.3). Cu and Ni values were used with Tables 1 and 2 of RG 1.99, Revision 2 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99, Revision 2 for welds and plates, respectively. Where applicable, data from the BWRVIP ISP (Reference 9.8) are used to adjust CFs, as reported in Reference 9.3. TransWare Report, No.ENT-FLU-001-R-001, Revision 0, (Reference 9.2) provides Pilgrim reactor vessel fluence values using the NRC-approved RAMA methodology. Pursuant to NRC approval of implementation of the BWRVIP ISP at Pilgrim (Pilgrim License Amendment 209, Reference 9.7), Entergy has performed fluence calculations using a bias factor of 1.0, which correlates to the benchmarked Monticello surveillance capsule activation data results achieved in 2008 (Reference 9.9). Therefore, the Pilgrim fluence without the bias factor is the representative fluence for Pilgrim Nuclear Power Station. Even though the TransWare calculations for Pilgrim were performed in December 2005, which was prior to the completion of the Monticello benchmarking in 2008, there have been no material changes in the RAMA Code and associated methodology that impact the previously completed fluence calculations for Pilgrim. Therefore, Reference 9.2 provides the applicable Pilgrim reactor vessel fluence values using the RAMA methodology without the need for a bias factor. The Pilgrim fluence calculation results show the reactor vessel will experience peak ID (at the clad/base metal interface) fluence at 34 EFPY of 7.53x1 017 n/cm 2 at the Lower Intermediate Weld 1-338A/C locations, and 8.42x1 017 n/cm 2 at the Lower Intermediate Shell Plates, respectively. Even though the peak fluence occurs in the Lower Intermediate Shell Plates, Lower Shell Plate 337-01 C has a higher ART due to material chemistry effects (peak fluence = 6.96x1 017 n/cm 2 at 34 EFPY). In addition, the N2 nozzle peak fluence at 34 EFPY was calculated to be 1.90x1017 n/cm 2 (Reference 9.12), and the N1 6A/B instrument ("drill-hole" style) nozzle fluence at 34 EFPY was calculated to be 3.52x1015 n/cm 2 (Reference 9.3). Both nozzles were evaluated for their impact on the Pilgrim P-T curves, and the limits presented in this PTLR incorporate any effects of these nozzles accordingly. All fluence values at 34 EFPY are linearly interpolated from the data reported in Reference 9.2. Reference 9.3 provides ART calculations at 24, 34, 44, and 54 EFPY using extrapolated fluence based on operation to 20.7 EFPY (Reference 9.2). Material properties, initial RTNDT and CFs for all vessel plates, welds and nozzles exposed to a fluence greater than 1.0x1017 n/cm 2 are obtained from Reference 9.10, except where data is available from the ISP program. The Lower Shell Longitudinal Weld material CF has been adjusted based on ISP data reported in Reference 9.8. References 9.4'and 9.11 provide the development of the P-T Curves documented in this report. Those calculations describe the methodology used from Reference 9.1, as well as the relevant requirements of Appendix G to 10 CFR 50 (Reference 9.13) and Appendix G to ASME Section XI (Reference 9.14). As presented in the calculations, the RPV beltline, bottom head (including penetrations), non-beltline vessel regions were evaluated for brittle fracture, including the effects of the N2 and N1 6A/B nozzles. Pilgrim Pressure-Temperature Limits Report Revision 0 In accordance with Section 2.4 of Reference 9.1, the P-T curves for the Core Not Critical" (Curve B) and Core Critical (Curve C) operating conditions 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 constraints,-and is well above the P-T curve limits. For Core Not Critical (Curve B) and Core Critical (Curve C) conditions, the P-T curves specify a coolant heatup and cooldown temperature rate of 5100 0 F/hr for which the curves are applicable. For Hydrostatic Pressure and Leak Test (Curve A) conditions, a coolant heatup and cooldown temperature rate of <25 0 F/hr must be maintained based on the requirements of Reference 9.1. The P-T limits and corresponding limits of either Curves A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure test conditions, the limits of Curve B may be conservatively used during pressure testing ifthe pressure test heatup and cooldown rate limits cannot be maintained. The only computer code used in the determination of the PNPS P-T curves was the ANSYS finite element computer program for the feedwater nozzle (non-beltline) and recirculation inlet nozzle (N2) stresses (various code versions as identified in References listed below). These analyses were performed to determine through-wall thermal and pressure stress distributions for the Pilgrim feedwater nozzles due to a step-change thermal transient (Reference 9.4), and for the recirculation inlet nozzles (N2) due to a shutdown transient at 100°F/hr (Reference 9.12). The ANSYS program was controlled under the vendor's 10 CFR 50 Appendix B Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems. The following inputs were used as input to the finite element analysis:

               "   With respect to operating conditions, stress distributions for a thermal shock of 456 0 F represents the maximum thermal shock for the feedwater nozzle during normal operating conditions. The stress results for a 4560 F shock are appropriate for use in developing the non-beltline P-T curves based on the limiting feedwater nozzle, as a shock of 456 0 F is representative of the Turbine Roll transient that occurs in the feedwater nozzle as part of the 100°F/hr startup transient. Therefore, these stresses represent bounding stresses in the feedwater nozzle associated with 100°F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region.

The thermal stress distributions used for the recirculation inlet nozzle were those generated for a plant shutdown transient at 100F/hr. Among the series of events analyzed this event produced the maximum hoop stress (Reference 9.12).

• Heat transfer coefficients used in the finite element analyses were those used in the original RPV stress report thermal analyses, the feedwater nozzle Pilgrim Pressure-Temperature Limits Report Revision 0 repair design specification and the reactor vessel cyclic load analysis.

(Reference 9.17) A three-dimensional finite element model was constructed for both the recirculation inlet and the feedwater nozzle evaluations. The models include a full 3600 representation of the nozzle, a section of the safe end, and the thermal sleeve. The local RPV shell included in the model extends 300 circumferentially and 40 inches longitudinally on each side of the nozzle centerline. The temperature dependent material properties used in the analyses are obtained from the ASME Code (the RPV shell, SA-533, Grade B Class 1; the nozzle and safe end, SA-508, Class 2; and the cladding and the nozzle-to-safe end weld,assumed to be Type 304 stainless steel) (References 9.12 and 9.18) Pilgrim Pressure-Temperature Limits Report Revision 0 6.0 PRESSURE-TEMPERATURE CURVES Pilgrim Pressure-Temperature Limits Report Revision 0 Figure 6.1 1,200_ 1,1000 . -.......

                            .....       ......                              i . . ... .if -.. . . ...

1,000Izl.. ......

                             -1........ ......... ......... ..........
           " ' 08  . ................... ......                 ..........

9 00 .....

     -U
     .. 70-0-..-.. ........................................................................

8 00.. .......

                                                             ..0 ..........----.-...................

S00

                          ,,, 3 0                             _.,!... ...i.....           *.                    ....

OI I otmHa 60- - -  ::::100

                               ---------            : : i~jJ7?--U-e                                                                                 -- Vesse----

10 0........................ ............

                                                                                                 ..... * T.....
                                                                                                                ..  .                         U p e...........

e s ... l ..- ......... 050 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F) PNPS Pressure Test Curve (Curve A), 34 EFPY Pilgrim Pressure-Temperature Limits Report Revision 0 Figure 6.2 1,200 .................

                  -- V 1,100
                  ~~~V*t       I F.

1,000 [4

   .9   900 A              .L~                                                                  ....................

w 800 0 LU 700 I ........ ku 600 0 500 Z 400 w w 300 4:; 200 .......

                           ............                        1 Beltline 4--

Bottom Head 100 ...... UpperVessel 0 ........ ..- 1. 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE ff) PNPS Heatup/Cooldown, Core Not Critical Curve (Curve 13), 34 EFPY _10-

Pilgrim Pressure-Temperature Limits Report Revision 0 Figure 6.3 1,200 1,100 1,000 7-- 900 C. #=Z 800 0.

   -j     700 U)

C', LU 600 0 I-500 F-Uj 400 ILl 300 i .... 200 100

                     -j  --- ----------
                                  ~    ~     ~ ~

0 0 50 100 150 200 1250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F) PNPS Heatup/Cooldown, Core Critical Curve (Curve C) 34EFPY

                                           -  11-

Pilgrim Pressure-Temperature Limits Report Revision 0 7.0 TABULATED PRESSURE AND TEMPERATURE VALUES FOR 34 EFPY CURVES Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.1. 34 EFPY Values for Beltline Pressure Test Curve A (Figure 6.1) Pressure-TemperatureCurve Calculation (PressureTest = Curve A) Input" Plant = Component = Vessel thickness, t = inches, so \t = 2.352 \,inch Vessel Radius, Ri = inches ARTNDT = °F =-=> 34 EFPY~ Cooldown Rate, CR = °F/hr KIT = ksi*inchl1 2 (From Appendix G, for cooldown rate above) AT1/4t = OF (no thermal for pressure test) Safety Factor = (for pressure test) Mm = (From Appendix G, for inside surface axial flaw) Temperature Adjustment = OF Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure for a full vessel at 70°F) Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kim P for P-T Curve P-T Curve (°F) ('F) (ksi*inchl1 2) (ksi*i nch 1 2) (psig) (°F) (psig) 60.0 60.0 47.52 31.68 0 60.0 0 60.0 60.0 47.52 31.68 706 60.0 688 65.0 65.0 49.03 32.69 729 65.0 710 70.0 70.0 50.69 33.80 754 70.0 735 75.0 75.0 52.53 35.02 781 75.0 763 80.0 80.0 54.57 36.38 811 80.0 793 85.0 85.0 56.81 37.87 844 85.0 826 90.0 90.0 59.30 39.53 881 90.0 " 863 95.0 95.0 62.04 41.36 922 95.0 904 100.0 100.0 65.07 43.38 967 100.0 949 105.0 105.0 68.43 45.62 1017 105.0 999 110.0 110.0 72.13 48.09 1072 110.0 1,054 115.0 115.0 76.22 50.82 1133 115.0 1,115 120.0 120.0 80.75 53.83 1200 120.0 1,182 125.0 125.0 85.75 57.17 1275 125.0 1,256 130.0 130.0 91.28 60.85 1357 130.0 1,338 135.0 135.0 97.39 64.92 1448 135.0 1,429 140.0 140.0 104.14 69.42 1548 140.0 1,530 145.0 145.0 111.60 74.40 1659 145.0 1,640 Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.2. Values for Upper Vessel Region Pressure Test Curve A (Figure 6.1) Pressure-Temr erature Curve Calculation (PressureTest = Curve A) Inputs Plant = Component = ,based on FW nozzle) Nozzle comer thickness, t = nches 1/4t = nches ARTNDT = IF =,= E Conservatively use flange. Cop = hird-order polynomial fit coefficient on eiper-psig basis Clp = hird-order polynomial fit coefficient on a per-psig basis C2p = third-order polynomial fit coefficient on a per-psig basis C3p = third-order polynomial fit coefficient on a per-psig basis Safety Factor = Temperature Adjustment = Pressure Adjustment = Unit Pressure = Flange RTNDT Fluid Calculated Adjusted Adjusted Temperature 1/4t Pressure Temperature Pressure for T Temperature Kic KIm P for P-T Curve P-T Curve (°F) (°F) (ksi*inchl1 2) (ksi*inchl1 2) (psig) (OF) (psig) 60.0 60.0 89.56 59.71 0 60.0 0 60.0 60.0 89.56 59.71 313 60.0 313 100.0 100.0 158.63 105.76 313 100.0 313 100.0 100.0 158.63 105.76 1062 100.0 1062 105.0 105.0 171.83 114.55 1151 105.0 1151 110.0 110.0 186.40 124.27 1248 110.0 1248 115.0 115.0 200.00 133.33 1339 115.0 1339 Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.3. Values for Bottom Head Pressure Test Curve A (Figure 6.1) Pressure-TemperatureCurve Calculation (PressureTest, Curve A) InIp u Plant = Component penetration Portior Vessel thickness, t = ches, so <t = 2.693 <inch Vessel Radius, Ri = inches (maximum) ARTNDT = °F ..... EFP*,.,* s Cooldown Rate, CR = °F/hr KIT = ksi*inch1/ 2 (From Appendix G, for cooldown rate above) AmlI/4t =

                                                                     'F (no thermal for pressure test)

Safety Factor = Mm = (From Appendix G, for inside surface axial flaw) Temperature Instrument Error = oF Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure for a full vessel at 700F) Fluid Calculated Adjusted Adjusted Temperature 1/4t Pressure Temperature Pressure for T Temperature KIc Kim P for P-T Curve P-T Curve (OF) (°F) (ksi*inch1/ 2) (ksi*inch1/ 2 ) (psig) (°F) (psig) 60.0 60.0 71.74 47.83 0 60.0 0 60.0 60.0 71.74 47.83 815 60.0 797 65.0 65.0 75.80 50.53 861 65.0 843 70.0 70.0 80.28 53.52 912 70.0 894 75.0 75.0 85.23 56.82 968 75.0 950 80.0 80.0 90.70 60.47 1030 80.0 1,012 85.0 85.0 96.75 64.50 1099 85.0 1,081 90.0 90.0 103.43 68.95 1175 90.0 1,157 95.0 95.0 110.82 73.88 1259 95.0 1,241 100.0 100.0 118.98 79.32 1352 100.0

• 1,333 105.0 105.0 128.00 85.33 1454 105.0 1,436 110.0 110.0 137.97 91.98 1568 110.0 1,549 115.0 115.0 148.99 99.33 1693 115.0 1,674 Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.4. 34 EFPY Values for Beltline Core Not Critical Curve B (Figure 6.2)

Pressure-TemrnperatureCurve Calculation (Heatup/CoolIdow, Core Not Critical= Curve B) luts: Plant = Pilgrim Component = *Belflne Vessel thickness, t = ,5.5312* inches, so 4t = 2.352 - -inch Vessel Radius, R = 113.91 inches 0 ARTNDT = F ====> 34 E'FP 0 Cooldown Rate, CR = F/hr 1 2 ksi*inch / (From Appendix G, for cooldown rate above) ATI/4 1 00 'F (conservatively neglect) Safety Factor 2 00 >' (for heatup/cooldown) Mm 2.178 (From Appendix G, for inside surface axial flaw) Temperature Adjustment 00 F Height of Water for a Full Vessel 507.5 inches Pressure Adjustment 18.3 psig (hydrostatic pressure for a full vessel at 70°F) Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kim P for P-T Curve P-T Curve ('F) 0 1 2 2 ( F) (ksi-inch ) (ksi*inch1l ) (psig) (°F) (psig) 0.0 0.0 0 10 0 60.0 0 0.0 0.0 37.51 15.33 342 60.0 323 5.0 5.0 37.97 15.56 347 60.0 328 10.0 10.0 38.47 15.81 352 60.0 334 15.0 15.0 39.02 16.08 359 60.0 340 20.0 20.0 39.64 16.39 365 60.0 347 25.0 25.0 40.31 16.73 373 60.0 355 (, 30.0 30.0 41.06 17.10 381 60.0 363 35.0 35.0 41.89 17.51 391 60.0 372 40.0 40.0 42.80 17.97 401 60.0 382 45.0 45.0 43.81 18.48 412 60.0 394 50.0 50.0 44.93 19.03 424 60.0 406 55.0 55.0 46.16 19.65 438 60.0 420 60.0 60.0 47.52 20.33 453 60.0 435 60.0 60.0 47.52 20.33 453 60.0 435 65.0 65.0 49.03 21.09 470 65.0 452 70.0 70.0 50.69 21.92 489 70.0 470 75.0 75.0 52.53 22.84 509 75.0 491 80.0 80.0 54.57 23.85 532 80.0 514 85.0 85.0 56.81 24.98 557 85.0 539 90.0 90.0 59.30 26.22 585 90.0 566 95.0 95.0 62.04 27.59 615 95.0 597 100.0 100.0 65.07 29.11 649 100.0 631 105.0 105.0 68.43 30.78 686 105.0 668 110.0 110.0 72.13 32.64 728 1.10.0 709 115.0 115.0 76.22 34.68 773 115.0 755 120.0 120.0 80.75 36.95 824 120.0 805 125.0 125.0 85.75 39.45 880 125.0 861 130.0 130.0 91.27 42.20 941 130.0 923 130.0 130.0 91.29 42.22 941 130.0 923 135.0 135.0 97.40 45.27 1009 135.0 991 140.0 140.0 104.15 48.65 1085 140.0 1,066 145.0 145.0 111.61 52.38 1168 145.0 1,150 150.0 150.0 119.86 56.50 1260 150.0 1,241 155.0 155.0 128.97 61.06 1361 155.0 1,343 160.0 160.0 139.04 66.09 1474 160.0 1,455 165.0 165.0 150.18 71.66 1598 165.0 1,579 170.0 170.0 162.48 77.81 1735 170.0 1,717

                                                                  -  16-

Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.5. Values for Upper Vessel Region Core Not Critical Curve B (Figure 6.2) Pressure-TemperatureCurve Calculation (Heatup/Cooldovi, Core Not Critical= Curve B) InjoyS Plant = 'iPIgrim' Component =ýtUpper VesseI (bas ed on FW nozzle) Nozzle comer thickness, t = , 7.5693 inchees 1/4t = 1.69n inchees ARTNoT = 10 F = All>EFPYs Conservatively use flange. Cp= 42.9728 third--order polynomial fit coefficient on a per-psig basis C p = .- 10.9363 third--order polynomial fit coefficient on a per-psig basis C2p= 17036 third--orderpolynomial fit coefficient on a per-psig basis C3p= -01214,~ third- -order polynomial fit coefficient on a per-psig basis COT= 8*t25,61 third- -order polynomial fit coefficient for thermal ClT = -2361.09 third-

                                                                                -orderpolynomial fit coefficient for thermal C2T =       1387.73     third-
                                                                                -orderpolynomial fit coefficient for thermal CaT  =T                 third- -order polynomial fit coefficient for thermal KIT =        11 .1       ksi'ir 1 2 nch' (using thermal coefficients above)

Safety Factor = !< 2.00 Temperature Adjustment 0. - Pressure Adjustment .0 psig Unit Pressure =.

• 1,565 psig Flange RTNOT  ; 10.0 -F Fluid Calculated Adjusted Adjusted Temperature 1/4t Pressure Temperature Pressure for T Temperature Kic K,,. P for P-T Curve P-T Curve 2 1

(*F) (°F) (ksi*lnch" ) (ksi*inch 2) (ps9g) (*F) (pslg) 10.0 10.0 53.93 2.84 0 60.0 0 10.0 10.0 53.93 2.84 48 60.0 48 15.0 15.0 56.11 3.13 52 60.0 52 20.0 20.0 58.52 3.47 58 60.0 58 25.0 25.0 61.19 3.88 65 60.0 65 30.0 30.0 64.13 4.35 73 60.0 73 35.0 35.0 67.38 4.91 82 60.0 82 40.0 40.0 70.98 5.57 93 60.0 93 45.0 45.0 74.95 6.34 106 60.0 106 50.0 50.0 79.34 7.25 121 60.0 .121 55.0 55.0 84.20 8.31 139 60.0 139 60.0 60.0 89.56 9.56 160 60.0 160 65.0 65.0 95.49 11.03 185 65.0 185 60.0 60.0 89.56 9.56 160 60.0 160 60.0 60.0 89.56 9.56 160 60.0 160 65.0 65.0 95.49 11.03 185 65.0 185 70.0 70.0 102.04 12.74 213 70.0 213 75.0 75.0 109.28 14.73 247 75.0 247 80.0 80.0 117.28 17.05 285 80.0 285 85.0 85.0 126.12 20.38 341 85.0 313. 90.0 90.0 135.90 25.27 423 90.0 313 95.0 95.0 146.70 30.67 513 95.0 313 100.0 100.0 158.63 36.63 613 100.0 313 105.0 105.0 171.83 43.23 724 105.0 313 110.0 110.0 186.40 50.52 846 110.0 313 115.0 115.0 202.52 58.58 981 115.0 313 120.0 120.0 220.32 67.48 1130 120.0 313 125.0 125.0 240.00 77.32 1295 125.0 313 130.0 130.0 261.71 88.17 1476 130.0 313 130.0 130.0 261.80 69.92 1171 130.0 1,171 135.0 135.0 285.84 79.80 1336 135.0 1,336 140.0 140.0 312.41 90.92 1522 140.0 1,522 145.0 145.0 341.78 103.39 1731 145.0 1,731 150.0 150.0 374.23 117.38 1966 150.0 1,966 155.0 155.0 410.10 133.04 2228 155.0 2,228 160.0 160.0 449.74 150.56 2521 160.0 2,521 165.0 165.0 493.54 170.12 2849 165.0 2,849 170.0 170.0 541.96 191.96 3214 170.0 3,214 Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.6. Values for Bottom Head Core Not Critical Curve B (Figure 6.2) Pressure-TemperatureCurve Calculation (Heatup/Cooldown,Core Not Critical= Curve B) Inputs Plant = Component = (Penetration Portion) Vessel thickness, t = nches, so At = 2.693 ,,inch Vessel Radius, RI = nches (maximum) ARTNDT = F ======> EFPYs Cooldown Rate, CR = IF/hr 1 2 KIT = ksi*inch (From Appendix G, for cooldown rate AT1 /4t = IF (conservatively neglect) Safety Factor = Mm = (From Appendix G, for inside surface axial flaw) Temperature Instrument Error = IF Height of Water for a Full Vessel = inches Pressure Adjustment psig (hydrostatic pressure for a full vessel at 70' Fluid Calculated Adjusted Adjusted Temperature 1/4t Pressure Temperature Pressure for T Temperature KIc KIm P for P-T Curve P-T Curve (IF) (ks*inch1/2) 2 (*F) (ks*inch 1 ) (psig) (°F) (psig) 0.0 0.0 44.81 15.66 0 60.0 0 0.0 0.0 44.81 15.66 267 60.0 249 5.0 5.0 46.03 16.27 277 60.0 259 10.0 10.0 47.38 16.95 289 60.0 270 15.0 15.0 48.87 17.69 301 60.0 283 20.0 20.0 50.52 18.52 316 60.0 297 25.0 25.0 52.34 19.43 331 60.0 313 30.0 30.0 54.35 20.43 348 60.0 330 35.0 35.0 56.58 21.54 367 60.0 349 40.0 40.0 59.04 22.77 388 60.0 370 45.0 45.0 61.75 24.13 411 60.0 393 50.0 50.0 64.76 25.63 437 60.0 419 55.0 55.0 68.08 27.29 465 60.0 447 60.0 60.0 71.74 29.13 496 60.0 478 60.0 60.0 71.74 29.13 496 60.0 478 65.0 65.0 75.80 31.15 531 65.0 513 70.0 70.0 80.28 33.39 569 70.0 551 75.0 75.0 85.23 35.87 611 75.0 593 80.0 80.0 90.70 38.61 658 80.0 640 85.0 85.0 96.75 41.63 709 85.0 691 90.0 90.0 103.43 44.97 766 90.0 748 95.0 95.0 110.82 48.66 829 95.0 811 100.0. 100.0 118.98 52.75 899 100.0 881 105.0 105.0 128.00 57.26 976 105.0 957 110.0 110.0 137.97 62.24 1061 110.0 1,042 115.0 115.0 148.99 67.75 1155 115.0. 1,136 120.0 .120.0 161.17 73.84 1258 120.0 1,240 125.0 125.0 174.63 80.57 1373 125.0 1,355 130.0 130.0 189.47 87.99 1500 130.0 1,481 130.0 130.0 189.53 88.02 1500 130.0 1,482 135.0 135.0 205.97 96.24 1640 135.0 1,622 140.0 140.0 224.14 105.33 1795 140.0 1,777 145.0 145.0 244.22 115.37 1966 145.0 1,948 150.0 150.0 266.42 126.47 2155 150.0 2,137 155.0 155.0 290.95 138.73 2364 155.0 2,346 160.0 160.0 318.05 152.28 2595 160.0 2,577 165.0 165.0 348.01 167.26 2850 165.0 2,832 170.0 170.0 381.12 183.82 3133 170.0 3,114

                                                               -  18-

Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.7. 34 EFPY Values for Core Critical Curve C (Figure 6.3) Pressure-TemperatureCurve Calculation (Core Critical= Curve C) Inpu2 Plant = EFPY = Curve A Leak Test Temperature = 'F (at 1,100 psig) Hydro Test Pressure = psig Flange RTNDT Curve B Curve B Curve B Curve B Curve B Curve B Curve B Curve B Temperature Pressure for Temperature Pressure for Temperature Pressure for Minimum Minimum Curve C Curve C Beltline Beitline Bottom Head Bottom Head Upper Vessel Upper Vessel Temperature Pressure Temperature Pressure (°F) (psig) ('F) (psig) ('F) (psig) (°F) (psig) (°F) (psig) 0.0 0 0.0 0 10.0 0 10.0 0 70.0 0 0.0 323 0.0 249 10.0 48 10.0 48 70.0 48 5.0 328 5.0 259 15.0 52 15.0 52 70.0 52 10.0 334 10.0 270 20.0 58 20.0 58 70.0 58 15.0 340 15.0 283 25.0 65 25.0 65 70.0 65 20.0 347 20.0 297 30.0 73 30.0 73 70.0 73 25.0 355 25.0 313 35.0 82 35.0 82 75.0 82 30.0 363 30.0 330 40.0 93 40.0 93 80.0 93 35.0 372 35.0 349 45.0 106 45.0 106 85.0 106 40.0 382 40.0 370- 50.0 121 50.0 121 90.0 121 45.0 394 45.0 393 55.0' 139 55.0 139 95.0 139 50.0 406 50.0 419 60.0 160 60.0 160 100.0 160 55.0 420 55.0 447 65.0 185 65.0 185 105.0 185 60.0 435 60.0 478 60.0 160 60.0 160 100.0 160 60.0 435 60.0 478 60.0 160 60.0 160 100.0 160 65.0 452 65.0 513 65.0 185 65.0 185 105.0 185 70.0 470 70.0 551 70.0 213 70.0 213 110.0 213 75.0 491 75.0 593 75.0 247 75.0 247 115.0 247 80.0 514 80.0 640 80.0 285 80.0 285 120.0 285 85.0 539 85.0 691 85.0 313 85.0 313 125.0 313 90.0 566 90.0 748 90.0 313 90.0 313 130.0 313 95.0 597 95.0 811 95.0 313 95.0 313 135.0 313 100.0 631 100.0 881 100.0 313 100.0 313 140.0 313 105.0 668 105.0 957 105.0 313 105.0 313 145.0 313 110.0 709 110.0 1,042 110.0 313 110.0 313 150.0 313 115.0 755 115.0 1,136 115.0 313 115.0 313 155.0 313 120.0 805 120.0 1,240 120.0 313 120.0 313 160.0 313 125.0 861 125.0 1,355 125.0 313 125.0 313 165.0 313 130.0 923 130.0 1,481 130.0 313 130.0 313 170.0 313 130.0 923 130.0 1,482 130.0 1,171 130.0 923, 170.0 923 135.0 991 135.0 1,622 135.0 1,336 135.0 991 175.0 991 140.0 1,066 140.0 1,777 140.0 1,522 140.0 1,066 180.0 1,066 145.0 1,150 145.0 1;948 145.0 1,731 145.0 1,150 185.0 1,150 150.0 1,241 150.0 2,137 150.0 1,966 150.0 1,241 190.0 1,241 155.0 1,343 155.0 2,346 155.0 2,228 155.0 1,343 195.0 1,343 160.0 1,455 160.0 2,577 160.0 2,521 160.0 1,455 200.0 1,455 165.0 1,579 165.0 2,832 165.0 2,849 165.0 1,579 205.0 1,579 170.0 1,717 170.0 3,114 170.0 3,214 170.0 1,717 210.0 1,717 Pilgrim Pressure-Temperature Limits Report Revision 0 Table 7.8. PNPS ART Calculations for 34 EFPY Pilgrim RPV MaterialART Calculations (34 EFPY) 2 (NOTE: This table covers all RPV materialsv,#h an exposedfluence, E > 1 MeV, of greaterthan 1.Ox 10 17 n/cm .) interpolated Estimated Chemistry Chemistry Ad] stments For 114t Description Piece Code Heat Initial RTNoT Factor ARTNw Margin Terms ARTNO No. No. No. (*F) Cu(wt%) NI(wt%) (°F) (°F) oA(°F) oI(°F) ('F) Lower Shell #1 337-01A G-3109-2 C-2957-2 0 0.10 0.47 65.0 19.0 9.5 0.0 38.0 co Lower Shell #2 337-01B G-3109-1 C-2957-1 -3 0.10 0.48 65.0 19.0 9.5 0.0 35.0 "Lower3Shei#3 1# 337-01C 'G-3109-3 ,<C-2973-1, * -4 0.11 0.63 74.5 21.8'* 01.9 0.0 39.6 Lower-Int. Shell #1 337-03A G-3108-3 C-2945-2 -12 0.10 0.66 65.6 21.2 10.6 0.0 30.5 S Lower-Int. Shel #2 337-03B G-3108-1 C-2921-2 -30 0.14 0.60 100.0 32.4 16.2 0.0 34.8 Lower-In.Shell#3 337-03C G-3108-2 C-2945-1 -7 0.10 0.65 65.5 21.2 10.6 0.0 35.4 Estimated Chemistry Chemistry Adjustments For 1/4t Description Seam Heat Flux Type & Initial RTaor Factor ARTNr Margin Terms ARTmT

                                    ".2     No.           No.              LotNo.              (°F)    Cu(wt%) Ni(wt%)             (°F)          ('F)    o         o'(°F) o(°F)       (°F)
     ,L. Int. Shell Long Weld #1          1-338A    27204/12008     UnLI~ide1092 #3774          -48       0.219        0.996      231.1          715       28.0       0.0     78.58 L. Int.Shell Lort.ed2              138        70/20            id     0237               -48       0.219        0.996      2,31,.1        44.9      22.4       0.0     4,1.8 w

lit. Shell Lng~reid' L. Int./L. Shell Girth Weld 1-338C 1-344 27204/12008 21935 id 02#74 Linde 1092 #3869 4

                                                                                                -50 0.219 0.183
                                                                                                                     .20.96 0.704
                                                                                                                               ~231.1 172.2
                                                                                                                                             '70.5 46.4 28.0 23.2 20.0 0.0 78.5 42.8 LowerShell Long. Weld #1            2-338A        27204       Linde 1092 #3714            -34       0.203        1.018      300.2          66.0      14.0       0.0     60.0 Lower Shell Long. Weld #2           2-338B        27204       Linde 1092 #3714            -34       0.203        1.018      300.2          78.2      14.0       0.0     72.2 Lower Shell Long. Weld #3           2-338C        27204       Linde 1092 #3714            -34       0.203        1.018      300.2          68.6      14.0       0.0     62.6  ,

Fluence Information (see Note 2): Wall Thickness (inches) Fluence at ID Attenuation, 1/4t Fluence 0 1/4t Fluence Factor, FF 2 0 24 2 Full 13) 1/4t (n/cm ) e ." x (n/cm ) t) f(0.28-.O01og Location Lower Shell #1 5.531 1.383 6.96E+17 0.718 4.99E+17 0.293

*n             Lower Shell #2                            5.531              1.383          6.96E+17             0.718                  4.99E+17                     0.293' I*.            Lower Shell#3                             5.531              1.383          6.96E+17             0.7,18                 4.99E+17                     0.293 Lower-Int. Shell #1                          5.531              1.383          8.42E+17             0.718                  6.04E+17                     0.324 Lower-Int. Shell #2                          5.531              1.383          8.42E+17             0.718                  6.04E+17                     0.324 Lower-Int. Shell #3                          5.53               1.383          8.42E+17             0.718                  6.04E+17                     0.324 L. Int. Shell Long. Weld #1                       5.531              1.383          7.53E+17             0.718                  5.40E+17                     0.305
 . L. Int. Shell Long. Weld #2                       5.531              1.383          3.40E+17             0.718                  2.44E+17                     0.194
.4J    L. Int. Shell Long. Weld #3                       5.531              1.383          7.53E+17             0.718                  5.40E+17             ,       0.305 w       L. Int./L. Shell Girth Weld                      5.531              1.383          5.99E+17             0.718                  4.30E+17                     0.269 Lower Shell Long. Weld #1                          5.531              1.383          4.19E+17             0.718                  3.01E+17                     0.220 Lower Shell Long. Weld #2                          5.531              1.383          5.63E+17             0.718                  4.04E+17                     0.260 Lower Shell Long. Weld #3                          5.531              1.383          4.48E+17             0.718                   3.21E+17                    0.229 Notes:   1. Material information taken from SIA Report No. SIR-00-082, Revision 0, "Updated Evaluation of Reactor Pressure Vessel Materials Properties for Pilgrim Nuclear Power Station," August 2000, Tables 3-1 through 3-12.

2. Fluence values from Transware Report No. ENT-FLU-001-R-001, Revision 0, "Pilgrim Nuclear Power Station Reactor Pressure Vessel Fluence Evaluation," interpolated from Tables 7-1 and 7-2.

3. RPV minimum thickness = 5 17/32" per Section 3.3.2 of SIR-00-082, Revision 0.

C

                                                                                   - 20  -

Pilgrim Pressure-Temperature Limits Report Revision 0 8.0 PILGRIM REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H (Reference 9.15), Pilgrim has developed Reactor Vessel Material Surveillance Program Requirements (Reference 9.7), the first surveillance capsule at Pilgrim was removed in Refueling Outage (RFO) No. 4 after 4.17 EFPY of operation. The surveillance capsule 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 El 85. The surveillance capsule activation and fluence results are discussed in Reference 9.2. Pilgrim has replaced the original RPV material surveillance program with participation in the BWRVIP ISP (Reference 9.7). 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 Pilgrim reactor vessel. Representative surveillance capsule materials for the Pilgrim limiting beltline plate and weld are in another representative plant with a capsule withdrawal schedule controlled by the BWRVIP ISP. There are two remaining Pilgrim specimen capsules which will remain in place to serve as backup surveillance material for the BWRVIP program, or as otherwise needed. The activation results of the RFO-4 Surveillance Capsule have not provided a comparable bench marking results with the RAMA fluence methodology. As such, Pilgrim has used Monticello benchmarking to predict the Pilgrim fluence. Pilgrim remains as a participant in the NRC-approved BWRVIP Integrated Surveillance Program and will follow those requirements. The Pilgrim Reactor Vessel Material Surveillance Program is discussed in the Pilgrim UFSAR, Appendix M (Reference 9.16). Pilgrim Pressure-Temperature Limits Report Revision 0

9.0 REFERENCES

9.1 Structural Integrity Associates Report No. SIR-05-044-A, Revision 0, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," April 2007 9.2 TransWare Enterprises Report No. ENT-FLU-001 -R-001, "Pilgrim Nuclear Power Station Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 and 54 EFPY", Rev.0, dated October 6, 2006 9.3 Structural Integrity Associates Calculation No. PNPS-22Q-301, Revision 1, "ARTNDT and ART Evaluation." 9.4 Structural Integrity Associates Calculation No. PNPS-03Q-302, Revision 2, "Development of Heatup/Cooldown (Curves B & C) P-T Curves." 9.5 U. S. Nuclear Regulatory Commission Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001. 9.6 U. S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988. 9.7 Pilgrim License Amendment 209, Revision to the Vessel Material Surveillance Program, (TAC No. MC1 565), dated January 5, 2005. 9.8 EPRI Technical Report No. 1020231, "BWRVIP-135 Revision 2: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations," October 2009. 9.9 EPRI Technical Report No. 1016567, "BWRVIP-1 99: BWR Vessel and Internals Project Testing and Evaluation of the Monticello 3000 Capsule," December 2008 9.10 Structural Integrity Associates Report No. SIR-00-082, Revision 0, "Updated Evaluation of Reactor Pressure Vessel Materials Properties for Pilgrim Nuclear Power Station," August 2000. 9.11 Structural Integrity Calculation No. PNPS-03Q-301 Revision 2, "Development of Pressure Test (Curve A) P-T Curves," 9.12 Structural Integrity Associates Calculation No. PNPS-22Q-302, Revision 0, "N2 Nozzle Evaluation." 9.13 10 CFR 50 Appendix G. 9.14 2004 Edition of ASME Section XI Nonmandatory Appendix G. 9.15 10 CFR 50 Appendix H. 9.16 Pilgrim Nuclear Power Station UFSAR Appendix M. 9.17 Structural Integrity Associates Calculation No. PNPS-10Q-301 Revision 0, "Design Input for RPV and Nozzles Finite Element Analyses and Flaw Evaluation Readiness," (M1239) Pilgrim Pressure-Temperature Limits Report Revision 0 9.18 Structural Integrity Associates Calculation No. PNPS-10Q-304 Revision 0, "Finite Element Stress Analysis of RPV Feedwater Nozzle," (M1232) 9.19 Structural Integrity Associates Calculation No. PNPS-10Q-303 Revision 0, "Finite Element Stress Analysis of RPV Recirculation Inlet Nozzle," (M1230)

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