ML090640301

License Amendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report

ML090640301
Person / Time
Site:	Nine Mile Point
Issue date:	03/03/2009
From:	Polson K Constellation Energy Group, Nine Mile Point
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML090640301 (35)

Text

Keith J. Poison P.O. Box 63 Vice President-Nine Mile Point Lycoming, New York 13093 315.349.5200 315.349.1321 Fax 0 Constellation Energy-Nine Mile Point Nuclear Station March 3, 2009 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 1, Docket No. 50-220 License Amendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report Pursuant to 10 CFR 50.90, Nine Mile Point Nuclear Station, LLC (NMPNS) hereby requests an amendment to the Nine Mile Point Unit 1 (NMP1) Renewed Facility Operating License DPR-63. The proposed amendment would modify Technical Specification (TS) Section 3.2.1, "Reactor Vessel Heatup and Cooldown Rates," and Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization," by replacing the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature (P-T) limit curves with references to the Pressure and Temperature Limits Report (PTLR). In addition, a new definition for the PTLR would be added to TS Section 1.0, "Definitions," and a new section addressing administrative requirements for the PTLR would be added to TS Section 6.0, "Administrative Controls."

Relocation of the P-T limit curves to the PTLR is consistent with the guidance provided in NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." The proposed TS changes are consistent with the guidance provided in GL 96-03 as supplemented by Technical Specification Task Force (TSTF) traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR."

New P-T limit curves have been developed for NMP 1 that are valid for up to 46 Effective Full Power Years (EFPY) of core operation. These new curves have been developed using the NRC-approved analytical methods described in Structural Integrity Associates Report No. SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors." The NMP1 PTLR containing these new P-T limit curves has been prepared based on the methodology and template provided in SIR-05-044-A.

Document Control Desk March 3, 2009 Page 2 The Enclosure provides a description and technical bases for the proposed changes, existing TS pages and associated TS Bases pages marked up to show the proposed changes, a copy of the NMP1 PTLR, and a supporting calculation. NMPNS has concluded that the activities associated with the proposed amendment represent no significant hazards consideration under the standards set forth in 10 CFR 50.92.

The enclosed submittal contains no regulatory commitments.

The P-T limit curves currently contained in the NMP1 TS are valid for 28 Effective Full Power Years (EFPY) of core operation. Current projections indicate that NMP1 will reach 28 EFPY of core operation in January 2010. Therefore, to support continued plant operation, approval of the proposed license amendment is requested by December 11, 2009, with 30 days allowed for implementation.

Pursuant to 10 CFR 50.91(b)(1), NMPNS has provided a copy of this license amendment request, with Enclosure, to the appropriate state representative.

Should you have any questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.

Very truly yours,

Document Control Desk March 3, 2009 Page 3 STATE OF NEW YORK TO WIT:

COUNTY OF OSWEGO I, Keith J. Poison, being duly sworn, state that I am Vice President-Nine Mile Point, and that I am duly authorized to execute and file this license amendment request on behalf of Nine Mile Point Nuclear Station, LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Nine Mile Point employees and/or consultants. Such information has been reviewed in accordance with company practice andfI believe it to be reliable.

Subscribed and sworn,-!this before me,ofa Notary

',,M day Yv.ý, Public in2009.

and for the State of New York and County of WITNESS my Hand and Notarial Seal: /,- C .

Notary Public My Commission Expires:

SANDRA A. OSWALD S,-Notary Public. State of New York No. 01OS6032276 Date Qualified in Oswego CouP.ty Commission Expires 10A-3 0-KJP/DEV

Enclosure:

Evaluation of the Proposed Change cc: S. J. Collins, NRC R. V. Guzman, NRC Resident Inspector, NRC J. P. Spath, NYSERDA

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Description of the Proposed Change 2.2 Background

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

ATTACHMENTS

1. Nine Mile Point Unit 1 - Proposed Technical Specification Changes (Mark-up)

2. Nine Mile Point Unit 1 - Changes to Technical Specification Bases (Mark-up)

3. Nine Mile Point Unit 1 Pressure and Temperature Limits Report (PTLR)

4. Structural Integrity Associates, Inc. Calculation No. 0800297.301, Revision 1, Revised Pressure-Temperature Curves Nine Mile Point Nuclear Station, LLC March 3, 2009

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Facility Operating License DPR-63 for Nine Mile Point Unit 1 (NMP1).

The proposed amendment modifies the Technical Specifications (TS) by replacing the reactor vessel heatup and cooldown rate limits and the pressure and temperature (P-T) limit curves, contained in TS Sections 3.2.1 and 3.2.2, with references to the Pressure and Temperature Limits Report (PTLR).

Relocation of the P-T limit curves to the PTLR is consistent with the guidance provided in NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits" (Reference 1). The TS changes are consistent with the guidance provided in GL 96-03 as supplemented by Technical Specification Task Force (TSTF) traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR" (Reference 2).

New P-T limit curves have been developed for NMP1 that are valid for up to 46 Effective Full Power Years (EFPY) of core operation. These new curves have been developed using the analytical methods described in Structural Integrity Associates Report No. SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (Reference 3). The NMP1 PTLR containing these new P-T limit curves has been prepared based on the methodology and template provided in SIR-05-044-A.

The PTLR is provided in Attachment 3.

2.0 DETAILED DESCRIPTION 2.1 Description of the Proposed Change The proposed change includes the following TS revisions:

• TS Section 1.0, Definitions: New Definition 1.18, "Pressure and Temperature Limits Report," is added. The wording for this definition is consistent with that in TSTF-419-A.

• TS Section 3.2.1, Reactor Vessel Heatup and Cooldown Rates: The currently stated heatup and cooldown rates of 100'F in any one hour period are replaced with a reference the PTLR.

• TS Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization: The current P-T limit curves, Figures 3.2.2.a through 3.2.2.e, and associated Tables 3.2.2.a through 3.2.2.e, are deleted. The references to Figures 3.2.2.a through 3.2.2.e that are contained in TS Sections 3.2.2.a, 3.2.2.b, and 3.2.2.c are replaced with references to the PTLR. TS Section 3.2.2.d is revised by replacing the specified minimum reactor vessel head flange and head temperature of 100'F with a reference to the PTLR. In addition, in TS Section 3.2.2.b, the following obsolete phrase is deleted: "except when performing low power physics testing with the vessel head removed at power levels not to exceed 5 row(t)."

• TS Section 6.0, Administrative Controls: New Section 6.6.7, "Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR)," is added. The format and content of new Section 6.6.7 are consistent with that in TSTF-419-A. This new section: (1) identifies the individual TSs that address reactor coolant system P-T limits; (2) references the NRC-approved topical report that documents PTLR methodologies; and (3) requires that the PTLR and any revision or supplement thereto be submitted to the NRC.

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE provides the existing TS pages marked-up to show the proposed changes. Marked-up pages showing corresponding changes to the TS Bases are provided in Attachment 2 for information only. The TS Bases changes will be processed in accordance with the NMP1 TS Bases Control Program (TS 6.5.6).

2.2 Background The NRC safety evaluation transmitted by letter dated February 6, 2007 (Reference 4) discussed the Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report SIR-05-044, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 0, dated December 2005, that was submitted for NRC review and acceptance for referencing in subsequent licensing actions. The BWROG provided this report to support the ability of BWR licensees to relocate their P-T curves and the associated numerical limits (such as heatup and cooldown rates) from the facility TS to a PTLR, a licensee-controlled document, using the guidelines provided in GL 96-03 (Reference 1). The NRC safety evaluation concluded that SIR-05-044 satisfies the criteria in Attachment 1 to GL 96-03 and provides adequate methodology for BWR licensees to calculate P-T limit curves, and that by using this methodology and following the PTLR guidance in GL 96-03, as amended by TS Task Force (TSTF) traveler TSTF-419, BWR licensees can relocate the P-T limit curves and the associated heatup/cooldown rates from the TS to a PTLR. SIR-05-044 was issued as a final report (-A) in April 2007 (Reference 3).

TSTF-419-A (Reference 2) amends the Standard TS (NUREGs-1430, -1431, -1433, and -1434) by: (1) revising the definition for the PTLR to delete references to the TS Limiting Conditions for Operation (LCO) for the P-T limits; and (2) revising the reviewers note in STS 5.6.6.b to require identification, by number and title, of the NRC-approved topical reports that document PTLR methodologies, or the NRC safety evaluation for a plant-specific methodology by NRC letter and date. In addition, the revised STS 5,6.6.b reviewers' note includes a requirement that the PTLR contain the complete identification for each TS-referenced topical report used to prepare the PTLR, including the report number, title, revision, date, and any supplements. TSTF-419-A does not change the requirements associated with the review and approval of the methodology or the requirement to operate within the limits specified in the PTLR. Any change to a methodology not previously approved by the NRC would continue to require NRC review and approval prior to use.

New P-T limit curves have been developed for NMP1 that are valid for up to 46 Effective Full Power Years (EFPY) of core operation. These new curves have been developed using the analytical methods described in SIR-05-044-A. The NMP1 PTLR containing these new P-T limit curves has been prepared based on the methodology and template provided in SIR-05-044-A. The PTLR is provided in Attachment

3. The changes identified in TSTF-419-A have also been incorporated into the PTLR and the proposed TS changes.

3.0 TECHNICAL EVALUATION

10 CFR 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. These limits are defined by P-T limit curves for normal operations (including heatup and cooldown operations of the RCS, normal operation of the RCS with the reactor being in a critical condition, and anticipated operational occurrences) and during pressure testing conditions (i.e., inservice leak rate testing and/or hydrostatic testing conditions). Historically, the P-T limit curves for BWRs have been contained in the TS, necessitating the submittal of license amendment requests to update the curves. This caused both the NRC and licensees to expend resources that could otherwise be devoted to other activities.

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE Generic Letter 96-03 allows plants to relocate their P-T curves and associated numerical limits (such as heatup and cooldown rates) from the plant TS to a PTLR, which is a licensee-controlled document. As stated in the GL, during the development of the improved standard technical specifications (STS), a change was proposed to relocate the P-T limits currently contained in the plant TS to a PTLR. As one of the improvements to the STS, the NRC staff agreed with the industry that the curves may be relocated outside the plant TS to a PTLR so that the licensee could maintain these limits efficiently. One of the prerequisites for having the PTLR option is that the P-T curves and limits be derived using methodologies approved by the NRC, and that the associated licensing topical reports describing the approved methodologies be referenced in the plant TS. The purpose of SIR-05-044-A is to provide BWRs with an NRC-approved report that can be referenced in plant TS to establish BWR fracture mechanics methods for generating P-T curves/limits, thereby allowing BWR plants to adopt the PTLR option.

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 TS. Report SIR-05-044-A, prepared by Structural Integrity Associates, provides the current methodology for developing RCS P-T limit curves and other associated numerical limits for BWRs. The NMP1 P-T curves have been developed in accordance with the SIR-05-044-A methodology, as documented in calculation 0800297.301, "Revised Pressure-Temperature Curves,"

which is provided in Attachment 4.

• Section 2.0 of SIR-05-044-A provides the fracture mechanics methodology and its basis for developing P-T limits. Application of this methodology for NMP1 is described in Section 2.0 of calculation 0800297.301.

• Section 3.0 of SIR-05-044-A provides a step-by-step procedure for calculating P-T limit curves. This section indicates that typically three reactor pressure vessel (RPV) regions are evaluated with respect to P-T limits: (1) the beltline region; (2) the bottom head region; and (3) the non-beltline region. This section also notes that P-T limits may be developed for other RPV regions to provide additional operating flexibility. The NMP1 P-T curves consider three regions of the vessel; beltline, bottom head, and non-beltline (feedwater nozzle / upper vessel). No other RPV regions are considered.

Design inputs and assumptions utilized for NMP1 are described in Section 3.0 of calculation 0800297.301, and a description of the calculations performed for NMPl is provided in Section 4.0 of calculation 0800297.301.

0 Appendix B of SIR-05-044-A provides a template for the PTLR. The NMP1 PTLR, provided in Attachment 3 to this submittal, has been prepared in accordance with the template.

The methodology used to calculate RPV neutron fluence values utilized in the development of the NMP1 P-T limit curves is in accordance with the recommendations of Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 5).

This methodology was previously submitted to the NRC for review and approval by Nine Mile Point Nuclear Station, LLC (NMPNS) letter dated November 15, 2002 (Reference 6), as supplemented by NMPNS letters dated January 15, 2003 (Reference 7), July 31, 2003 (Reference 8), and September 15, 2003 (Reference 9). By letter dated October 27, 2003 that issued NMP1 License Amendment No. 183 (Reference 10), the NRC documented their determination that the NMPNS neutron fluence calculation methodology was acceptable.

The following proposed TS revisions associated with relocation of the P-T limits to a PTLR are consistent with the guidance provided in GL 96-03 as supplemented by TSTF-419-A:

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE

" New Definition 1.18, "Pressure and Temperature Limits Report," is added.

• In TS Section 3.2.1, Reactor Vessel Heatup and Cooldown Rates, the currently stated heatup and cooldown rates of 100'F in any one hour period are replaced with a reference the PTLR.

" In TS Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization, the current P-T limit curves (Figures 3.2.2.a through 3.2.2.e, and associated Tables 3.2.2.a through 3.2.2.e) are deleted, and the references to these figures in TS Sections 3.2.2.a, 3.2.2.b, and 3.2.2.c are replaced with references to the PTLR. TS Section 3.2.2.d is also revised by replacing the specified minimum reactor vessel head flange and head temperature of 100°F with a reference to the PTLR.

• In TS Section 6.0, Administrative Controls, new Section 6.6.7, "Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR)," is added. This new section: (1) identifies the individual TSs that address reactor coolant system P-T limits (i.e., TS Sections 3.2.1, 3.2.2, and 4.2.2); (2) references the NRC-approved topical report that documents PTLR methodologies (i.e.,

SIR-05-044-A); and (3) requires that the PTLR and any revision or supplement thereto be submitted to the NRC.

TS Section 3.2.2.b currently specifies that the applicable P-T limits shall be satisfied during RPV heatup and cooldown when the reactor is critical "except when performing low power physics testing with the vessel head removed at power levels not to exceed 5 mw(t)." This exception, originally added to the TS in 1974, is not presently needed, and operation under the stated conditions (reactor critical with the vessel head removed at power levels not in excess of 5 mw(t)) would not be consistent with any of the Reactor Operating Conditions defined in TS Section 1.1. Therefore, the stated exception is being deleted from TS Section 3.2.2.b.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The NRC has established requirements in 10 CFR 50 Appendix G in order to protect the integrity of the RCPB in nuclear power plants. Appendix G requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code were used to generate the P-T limits. 10 CFR Part 50 Appendix G also requires that applicable surveillance data from 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.

NRC regulatory guidance related to P-T limit curves is found in RG 1.99, Revision 2 (Reference 11) and Standard Review Plan (NUREG-0800) Section 5.3.2 (Reference 12). Adoption of the NRC-approved methodology described in SIR-05-044-A for the preparation of NMP1 P-T limit curves ensures that the requirements of 10 CFR 50 Appendix G will be satisfied.

10 CFR Part 50 Appendix H provides criteria for the design and implementation of RPV material surveillance programs for operating light water reactors. NMP1 demonstrates its compliance with the requirements of 10 CFR Part 50 Appendix H through participation in the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) (Reference 13).

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE Regulatory Guide 1.190 (Reference 5) describes methods and assumptions acceptable to the NRC for determining RPV neutron fluence. By letter dated October 27, 2003 (Reference 10), the NRC documented their determination that the NMPNS neutron fluence calculation methodology met the guidance of RG 1.190 and was acceptable.

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

(1) Have NRC-approved methodologies to reference in the TS, (2) Develop a PTLR to contain the P-T limit curves, associated numerical limits, and any necessary explanation, and (3) Modify applicable sections of the TS accordingly.

The NRC-approved methodology of SIR-05-044-A has been adopted for preparation of NMP1 P-T limit curves. The NRC review documented in Reference 4 concluded that SIR-05-044-A satisfies the criteria in of GL 96-03 and provides adequate methodology for BWR licensees to calculate P-T limit curves. A PTLR has been prepared for NMP 1 based on the methodology and template provided in SIR-05-044-A, and is provided in Attachment 3 to this submittal. Proposed revisions to applicable sections of the TS have been prepared and are provided in Attachment I to this submittal. These proposed TS changes are consistent with the guidance provided in GL 96-03, as supplemented by TSTF-419-A.

4.2 Precedent The NRC has approved similar license amendments to relocate P-T limit curves to a PTLR. Recent examples for boiling water reactor plants include:

" Oyster Creek Nuclear Generating Station (License Amendment No. 269 issued by NRC letter dated September 30, 2008 - ADAMS Accession No. ML082390685).

• James A. Fitzpatrick Nuclear Power Plant (License Amendment No. 292 issued by NRC letter dated October 3, 2008 - ADAMS Accession No. ML082630385).

4.3 Significant Hazards Consideration Nine Mile Point Nuclear Station, LLC (NMPNS) is requesting revisions to Nine Mile Point Unit 1 (NMP1) Technical Specifications (TS). The proposed changes would replace the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature (P-T) limit curves with references to the Pressure and Temperature Limits Report (PTLR). Relocation of the P-T limit curves to the PTLR is consistent with the guidance provided in NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." New P-T limit curves have been developed for NMP 1 that are valid for up to 46 Effective Full Power Years (EFPY) of core operation. These new curves have been developed using the analytical methods described in Structural Integrity Associates Report No. SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors." In a safety evaluation dated February 6, 2007, the NRC concluded that SIR-05-044-A provides adequate methodology for BWR licensees to calculate P-T limit curves.

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE NMPNS 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:

I1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes modify the TS by replacing references to existing reactor vessel heatup and cooldown rate limits and P-T limit curves with references to the PTLR. The proposed amendment also adopts the NRC-approved methodology of SIR-05-044-A for the preparation of NMP1 P-T limit curves. In 10 CFR 50 Appendix G, requirements are established to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. Implementing the NRC-approved methodology for calculating P-T limit curves and relocating those curves to the PTLR provide an equivalent level of assurance that RCPB integrity will be maintained, as specified in 10 CFR 50 Appendix G.

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

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

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

Response: No.

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

and installed equipment is not being operated in a new or different manner. Thus, no new failure modes are introduced.

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

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

Response: No.

The proposed changes do not affect the function of the RCPB or its response during plant transients. By calculating the P-T limits using NRC-approved methodology, adequate margins of 6 of 8

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE safety relating to RCPB integrity are maintained. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined, there are no changes to setpoints at which protective actions are initiated, and the operability requirements for equipment assumed to operate for accident mitigation are not affected.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, NMPNS concludes that the proposed amendment does not involve a 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.

4.4 Conclusions 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 issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

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

2. Technical Specification Task Force Traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," Rev. 0, dated August 4, 2003

3. Structural Integrity Associates Report No. SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 0, April 2007

4. 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 7 of 8

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE

5. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001

6. Letter from J. T. Conway (NMPNS) to Document Control Desk (NRC), dated November 15, 2002, "License Amendment Request Pursuant to 10 CFR 50.90: Revision of Reactor Pressure Vessel Pressure-Temperature Limits and Request for Exemption from Requirements of 10 CFR 50.60 (TAC Nos. MB6687 and MB6703)"

7. Letter from B. S. Montgomery (NMPNS) to Document Control Desk (NRC), dated January 15, 2003, "Transmittal of Neutron Transport Calculations Benchmarking Report (TAC Nos. MB6687 and MB6703)"

8. Letter from P. E. Katz (NMPNS) to Document Control Desk (NRC), dated July 31, 2003, "Request for Additional Information (RAI) - Amendment Application Re: Pressure-Temperature Limit Curves (TAC Nos. MB6687 and MB6703)"

9. Letter from W. C. Holston (NMPNS) to Document Control Desk (NRC), dated September 15, 2003, "Transmittal of Revised Neutron Transport Calculations Benchmarking Report (TAC Nos. MB6687 and MC033 1)"

10. Letter from P. S. Tam (NRC) to P. E. Katz (NMPNS), dated October 27, 2003, "Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment Re: Pressure-Temperature Limit Curves and Tables (TAC No. MB6687)"

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

12. NUREG-0800, NRC Standard Review Plan, Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock"

13. Letter from P. S. Tam (NRC) to J. A. Spina (NMPNS), dated November 8, 2004, "Nine Mile Point Nuclear Station Unit Nos. 1 and 2 - Issuance of Amendments Re: Implementation of the Reactor Pressure Vessel Integrated Surveillance Program (TAC Nos. MC1758 and MC 1759)"

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ATTACHMENT 1 NINE MILE POINT UNIT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

The current versions of the following Technical Specification pages have been marked-up by hand to reflect the proposed changes:

6 81 83 through 94 358 Nine Mile Point Nuclear Station, LLC March 3, 2009

1.16 Dose Equivalent 1-131 Dose Equivalent 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The dose conversion factors used for this calculation shall be the Committed Effective Dose Equivalent dose conversion factors listed in Table 2.1 of Federal Guidance Report No. 11, EPA, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

1.17 Recently Irradiated Fuel Recently irradiated fuel is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.18 ed) 1.19 (Deleted) li ect-1 1.20 (Deleted) 1.21 (Deleted)

AMENDMENT NO. !42, 472, 176,444- 66

INSERT 1 (forTS page 6; New Definition 1.18) 1.18 Pressure and Temperature Limits Report (PTLR)

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.6.7.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

-t 3.2.1 REACTOR VESSEL HEATUP AND COOLDOWN RATES Applicability:

Applies to the reactor vessel heating or cooling rate.

Objective:

To assure that thermal stress resulting from reactor heatup and cooldown are within allowable code limits.

Specification:

During the startup and shutdown operations of the reactor, the reactor vessel!L incroaeedl more theni 4-9(3F in any oncehoupori, docrzr ncr mcA Ile ta O~

-Teq+/-,4,urp Lc-MLS Reoret~ shT1).

AMENDMENT NO. 8-142 81

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT REQUIREMENT 3.2.2 MINIMUM CONDITION LIMITINGREACTOR VESSEL OPERATION FOR TEMPERATURE FOR 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR SURVEILLANCE 3.2,2 MINIMUM REACTOR VESSEL TEMPERATURE FOR 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION PRESSURIZATION Applicability: Applicability:

Applies to the minimum vessel temperature required Applies to the required vessel temperature for for vessel pressurization. pressurization.

Obiective: Objective:

To assure that no substantial pressure is imposed on To assure that the vessel is not subjected to any the reactor vessel unless its temperature is substantial pressure unless its temperature is greater considerably above its Nil Ductility Transition than its Nil Ductility Transition Temperature (NDTT).

Temperature (NDTT).

Specification: Specification: "/

a. During reactor vessel heatup and cooldown when a. Reactor vessel temperature and pressure shall be the reactor isand critical, the not pressure vessel reactors~ monitored and controlled to ass re that the temperature shall h {jt -

pressure and temperature limitsfare met.

b. During reactor vessel heatup and cooldown when the reactor is critical, the reactor vessel temperature and pressure shall xcpt-when p~f~mn o oe hzc

" 1n . ith

. h.... l Q . d romolo

. po... i"' 44,c_. PTL-,.

lo'.. .. agnotood AMENDMENT NO. 442-, 484- 83 83

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION 1- SURVEILLANCE REQUIREMENT

c. During leakage and hydrostatic testing, the reactor vessel temperature and pressure shall i~. na4aveA~j~a~44~ YLd QInd- coodoWn O f8r tho pRWpese of leakcage arel hydrostatic t@&ting, the reactfr Yesse4-to ngpe-rauwe and peroccuro she!! satisfy tthe rgirMeat-, of F~ig- xgre 3.:2.2. ;and-- .2 f8r non critical heatup Candoldowi1 , '-spectvely.

d. The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head are equal to or greater than AMENDMENT NO. !42,4!64, 183, 484-) 844

es99 pflwo 4 Vele-"e V

-- i EATUP - CORE NOT CRITIC AL

\

HI 15000 Minim mn Te /erature f Boltup:

00OF 1000 ----- -----

REACTOR PRESSURE (psig) 500 n

100 150 200 250 300 35 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (fF)

(reactor vessel belt , e downcomer water temperature is measured at recirc\tion loop suction)

(instrument unce ainties have been included in this figure)

FIGURE 3.2.2.a IVImllmVIUiVI DE.LI LIl'lr- LJ.VVI  %,%IVIr-I1VVR1 'l% I ElVIlMrrl'l/- urnl rwn F

PRESSURIZATION DURING HEATUP AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL) (HEATING RATE*< 100'F/HR)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 142, 164,4.8.3- 85

C LIMIT FOR NON-CRITICAL OPERATION HEATUP AT UP TO 100 0F/HR REACTOR VESSEL BELTLINE REACTOR PRESSURE (psig) DOWNCOMER WATER IN TOP DOME TEMPERATURE (OF) 0 100 298 100 298 102 298 107 298 11 298 11)4-IS #"t G2)LETED) 298 22 8 127 29 132 300 137 304 142 311 147 319 152 329 157 340 162 354 167 369 172 387 177 406 182 406 182 429 187 454 192

.483 197 515 202 547 207 582 212 622 217 665 222 713 227 767 232 840 238 840 238 895 242 969 47 10 0 140 25 (reactor essel beltline downcomer water temperature is measured at recirculation loop ction)

(instrument uncertainties have been included in this table)

TABLE 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEAT-UP (CORE NOT CRITICAL) (HEATING RATE _ 100°F/HR)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION AMENDMENT NO. 142,164, 183 86

COOLDOWN - CORE NOT CRI TICAL T-t$

ýf~ LPAT u rn T mperature Mi

/or Boltup:

1000 F REACTOR PRESSURE (psig) 500 ----- .............- .- - - --.. -

Il 0 50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE OWNCOMER WATER TEMPERATURE (°F)

(reactor vessel beltlin owncomer water temperature is measured at recirculati n loop suction)

(instrument uncertai ies have been included in this figure)

FIGURE 3.2.2.b MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL) (COOLING RATE 5 100°F/HR)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 142-,-64, 183 87

LIMIT FOR NON-CRITICAL OPERATION COOLDOWN AT UP TO 100 0 F/HR REACTOR VESSEL BELTLINE REACTOR PRESSURE (psig) DOWNCOMER WATER O O __..__ _TEMPERATURE ('F) 0 100 205 100 209 100 X 213 100 10 22* 0 23100 242 100 250 *100 2S4 100 258 102 268 107 278 112 290 117 302 122 316 127 332 132 349 137 360 140 360 160 455 160 471 163 471 163 498 167 532 172 570 177 613 182 659 187 701 192 737 197 777 202 820 207 869 212 9227 119 22 1119 232 (react vessel beltline downcomer water temperature is measured at recirculation loop sucti n)

(instrument uncertainties have been included in this table)

TABLE 3.2.2.b MINIMUM TEMPERATURE FOR PRESSURIZATION .DURING COOLDOWN (CORE NOT CRITICAL) (COOLING RATE : 100°F/HR)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION AMENDMENT NO. 142 , 163883 88

HEATUP - CORE CRITICS

~TP-i- e~&tG ' 2tzTiED I

I 1500

-L. XM, imum emperatu re for Boltup:

100° F 1000 .7-7 REACTOR PRESSURE (psig)

Water Level Must 500 - Be in Range For Power Operation If Core Is Critical Below 222 0 F A

0 5 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (_F)

(reactor vessel eltline downcomer water temperature is measured at recirc ation loop suction)

(instrument certainties have been included in this figure)

FIGURE 3.2.2.c MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (HEATING RATE5<1000 F/HR) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 142,164, 183 89

LIMIT FOR POWER OPERATION (CORE CRITICAL)

HEATUP AT UP TO 100°F/HR REACTOR VESSEL BELTLIN REACTOR PRESSURE (psig) DOWNCOMER WATER IN TOPDOME TEMPERATURE (OF) 0 100 298 .100 298 172 300 -177 304 *M E -2&LE--3*DI 18 3 /92 329 197 340 202 354 207 360 212 360 217 360 222 a 406 222 a 429 227 454 232 483 237 515 242 547 247 582 252 622 257 665. 262 713 267 767 272 840 278 840 278 895 282 969 287 1050 292 1140 297 (reactor vessel bel ne downcomer water temperature is measured at recirculatio loop suction)

('water vel must be in range for power operation if core is critical below 2 'F)

(instrument uncertainties have been included in this table)

TABLE 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEATUP (CORE CRITICAL) (HEATING RATE _ 100"F/HR)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION I

AMENDMENT NO. 4-42,-164, 183 90

COOLDOWN - CORE CRIT ICAL

-14)S eAr&G "PELETED~

/Mini urn 500 Teperature f rBoltup:

100OF 1000 REACTOR PRESSURE -- - ------ - --

(psig)

W ater Level Must 500 . . . . . . . . ... .....

Be in Range For Power Operation If Core Is Critical Below 203°F Al 0 50 100 150 200 250 300 50 400 REACTOR VESSEL BELTLINE OWN COM ER WATER TEMPERATURE (°F)

(reactor vessel b tline downcomer water temperature is measured at re *rculation loop suction)

(instrument u ertainties have been included in this figure)

FIG URE 3.2.2. d MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (COOLING RATE!<100-F/HR) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 1-4 --_-64, 183 91

LIMIT FOI L POWER OPERATION (CORE CRITICAL)

CJUOLING AT UP TOI1f0t0ttl(

REACTOR VESSEL BELTLINE REACTOR PRESSURE (psig) DOWNCOMER WATER IN TOP DOME TEMPERATURE (OF) 0 100 205 100 209 102 213 IRLF7 D107 8 T-S22 f-c L T 2'112 117 229 122 235 127 2421 250 254 140 258 142 268 147 278 152 290 157 302 162 316 167 332 172 349 177 360 180 360 200 360 203 a 471 203 a 498 207 532 212 570 217 613 222 659 227 701 232 737 237 777 242 820 8692 247 922 22 98226 10* 267

• 19 272 1199

• 277 (reacto vessel beltline downcomer water temperature is measured at recirculation loop suction*

is critical below 203'F)

('water level must be in range for power operation if core (instrument uncertainties have been included in this table)

TABLE 3.2.2.d MINIMUM TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE CRITICAL) (COOLING RATE : 100'F/HR)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS UPC kUKUt UJrnKA I MUIN I'

AMENDMENT NO. 142, !464, 183 92

LEAK/HYDRO TEST - CORE NOT CRITICAVL 1r;nnTWS A& ELT Mi imum em perature for Boltup:

1 00'F 1000 .. ..a.a ...a. a.

REACTOR PRESSURE (psig) 500 A

0 50 100 150 200 250 300 50 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (-F)

(reactor vesse eltline downcomer water temperature is measured at recircula on loop suction)

(instrument ncertainties have been included in this figure)

FIGURE 3.2.2.e MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR

/

PRESSURIZATION DURING IN-SERVICE HYDROSTATIC TESTING AND LEAK TESTING (CORE NOT CRITICAL) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 142,!64,183 93

LIMIT FOR IN-SERVICE TEST (CORE NOT CRITICAL, FUEL IN VESSEL)

REACTOR VESSEL BELTLI CTOR PRESSURE (psig) DOWNCOMER WATE)L TOP DOME TEMPERATURE 0 100 3 10 360 40e G 'DELLz-EjE 0 130 704 135 722 140 742 145 764 150 788 155 815 160 844 165 877 170 913 175 953 180 997 185 1046 190 1100 195 1160 200 (reactor vessel beltline ncomer water temperature is measured at recircul on loop suction)

(i rument uncertainties have been included in this table)

TABLE 3.2.2.e MINIMUM TEMPERATURE FOR PRESSURIZATION DURING LEAKIHYDROSTATIC TESTING (CORE NOT CRITICAL)

FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION AMENDMENT NO. !42, !464,183 94

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

1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," U.S. Supplement, (NRC approved version specified in the COLR).

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

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

6.6.6 Special Reports Special reports shall be submitted within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. - f. (Deleted)

g. Sealed Source Leakage In Excess Of Limits, Specification 3.6.5.2 (Three months).

h. Accident Monitoring Instrumentation Report, Specification 3.6.11.a (Table 3.6.11-2, Action 3 or 4) (Within 14 days following the event).

AMENDMENT NO. 142, 157, 162, 181, +4-,35) 358

INSERT 2 (for TS page 358; New Section 6.6.7) 6.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

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

1. Limiting Condition for Operation Section 3.2.1, "Reactor Vessel Heatup and Cooldown Rates."

2. Limiting Condition for Operation Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization."

3. Surveillance Requirement Section 4.2.2, "Minimum Reactor Vessel Temperature for Pressurization."

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:

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

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

ATTACHMENT 2 NINE MILE POINT UNIT 1 CHANGES TO TECHNICAL SPECIFICATION BASES (MARK-UP)

The current versions of Technical Specifications Bases pages 82 and 95 have been marked-up by hand to reflect the proposed changes. These Bases pages are provided for information only.

Nine Mile Point Nuclear Station, LLC March 3, 2009

BASES FOR 3.2.1 REACTOR VESSEL HEATUP AND COOLDOWN Design calculations reported in Volume I, Section V-A, 4.0 (page V-6)* have demonstrated that the heatup and cooldown rate of 100 0 F/hr considered in the fatigue analysis will result in stresses well within code limits. A series of calculations have demonstrated that various extreme heatup and cooldown transients result in thermal strains well within the ASME Code limits stated in Volume I, Section V-C, 3.0 (p.

V-19)* Cooldown incidents include: failure of the pressure regulator leading to a cooldown of 215 0 F in 5.5 minutes (Appendix E-l, 3.15 (p. E-45))*, inadvertent opening of a single solenoid-actuated pressure relief valve leading to a cooldown of 1050°F/hr sustained for 10 minutes (Vol. I, Section V-B, 1.3 (p. V-1 1))*, and finally, opening all six of the solenoid-actuated relief valves leads to a cooldown of 250°F in 7.5 minutes (Volume IV, Section I-B)*. Reactor vessel heatup of 3000F/hr (Volume IV, Section 1-B)* also demonstrates stresses well within the code requirements. vI_ spr Ih_ heau_

___,' ad c.uldwn

______f__

_ a._r__ _de, badier , to ,.-

• FSAR AMENDMENT NO. 142 82

INSERT A (for TS page 82)

The maximum allowable reactor vessel heatup and cooldown rates during normal startup and shutdown operations are specified in the Pressure and Temperature Limits Report (PTLR). These limits affect the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing pressure-temperature (P-T) limit curves that are contained in the PTLR. Thus, operation within the specified limits for rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P-T limit curves.

BASES FOR 3.2.2 AND 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION res 92.2.a, 3.2.2.b .2.2.c, and 3K.2 I2are plots of ppr psure versus te erature for heat and cooldown ra s of up to 1000 F/hr.

n/n(Specifica n .21). Fiu322e is the plot f pressure versu temperature for akage and hydra atic testing. When ee 6p rate to the -nimum test te perature for leaka(? and hydrostatic esting is maintai d < 100 F/hr, the t ermal gradient acro thee el wall is ne ;igible, howeve , if the heatup rate xceeds 100OF/hr, hermal soak is r quired. These cu es are base on c ulations ress in/ten y factors acco ing to Appedix G )f Section XI of t ASME Biea PrsueVslod199Ed ition id Code Case

,0. In a ition, tempera, re shifts due to fa neutron fluence twenty-eight eff tive full power ye s of operation wer incorporated into igure . Thes shits re calculated usin the procedure pr sented in Regul ry Guide 1.99, R ision 2. Reactor ssel flange/react flf, ge boltup is g erned by other crit ia as stated in S cification 3,2.2. The pressure re ings on the figure ave been adjus dt tfor instrume uncertainties and t reflect the calcul ed elevation he difference betwe the pressure se ing instrument lo tions he pressure s sitive area of the re beltline region The temperatur readings on the fi res have been a 'sted to account fr instrument

!rtainties.

The reactor vessel head flange and vessel flange in combination with the double "0"ring type seal are designed to provide a leak-tight seal when bolted together. When the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flanges. Ith thc RIF-

, I i. , e minimum vessel flange and head flange temperature for bolting is established .'n TL igures .2.a, 3.2.2.b, . .2.c, 3.2.2,d,,d 3.2.2.e hay ncorporated a mperature s due to the cculated fast n ron fluence.

Then ron flux at th essel wall is culated base on Regulatory ide 1.190 pliant metholusing a plantecific model v dated to x monitors in lled inside the essel. The c es are a licae for u to tw~6t -ei ht effee full ower rs of operatioe AMENDMENT NO. 142, !64, Revision 9,-12 (A1,,-r84) 95

INSERT B (for TS page 95)

The Pressure and Temperature Limits Report (PTLR) establishes the methodology for determining pressure-temperature (P-T) limits and contains the P-T limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing. The heatup curve provides limits for both heatup and criticality. Each P-T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The pressure values on the curves have been adjusted to account for instrument uncertainties and to reflect the calculated elevation head difference between the pressure sensing instrument locations and the pressure-sensitive area of the core beltline region. The temperature values on the curves have been adjusted to account for instrument uncertainties.

10 CFR 50, Appendix G, requires the establishment of P-T limits for material fracture toughness requirements of the reactor coolant pressure boundary materials. 10 CFR 50, Appendix G requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic and leak tests, and mandates that the P-T limits be at least as conservative as limits obtained by following the methods of Appendix G to Section XI of the ASME Code. The operating limits specified in the PTLR provide a margin to brittle failure of the reactor vessel and ensure that the requirements of 10 CFR 50 Appendix G are satisfied.

The P-T limit curves are established based on limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P-T limit curves, different locations are more restrictive. In addition, heatup operations represent a different set of restrictions than cooldown operations because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. The P-T limit curves reflect the most restrictive results from consideration of heatup and cooldown operations. The criticality limits include the 10 CFR 50, Appendix G requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for inservice leakage and hydrostatic testing.