ML123380336
| ML123380336 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/21/2012 |
| From: | Landgon K Constellation Energy Nuclear Group, EDF Group, Nine Mile Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML123380336 (85) | |
Text
{{#Wiki_filter:Attachment 6 to the Enclosure transmitted herewith contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When separated from the Enclosure, this document is decontrolled. Ken Landgon P.O. Box 63 Vice President-Nine Mile Point Lycoming, New York 13093 315.349.5200 315.349.1321 Fax CENGS. a joint venture of Constellation negy" *g*eDF Energy, &1N NINE MILE POINT NUCLEAR STATION November 21, 2012 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk
SUBJECT:
Nine Mile Point Nuclear Station Unit No. 2, Docket No. 50-410 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 2 (NMP2) Renewed Facility Operating License NPF-69. The proposed amendment would modify Technical Specification (TS) Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits," 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 1.1, "Defmnitions," and a new section addressing administrative requirements for the PTLR would be added to TS Section 5.0, "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."
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Document Control Desk November 21, 2012 Page 2 New P-T limit curves have been developed for NMP2 that are valid for a peak internal diameter (ID) fluence of 9.60E+17 n/cm2 , which corresponds to a projected 32 Effective Full Power Years (EFPY) of core operation. These new curves have been developed using Boiling Water Reactor Owners' Group (BWROG) Licensing Topical Report (LTR) NEDC-33178P-A, which has been approved by the NRC in a letter from T.R. Bount, USNRC to D. Coleman, BWROG, "Final Safety Evaluation for Boiling Water Reactors Owners' Group Licensing Topical Report NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (TAC No. MD2693)," dated April 27, 2009. The Enclosure provides a description and technical bases for the proposed changes with attachments for existing TS pages and associated TS Bases pages marked up to show the proposed changes, a copy of the draft NMP2 PTLR, and responses to previous requests for additional information (RAI) to other licensees related to this issue which have been customized for NMP2. 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 P-T limit curves currently contained in the NMP2 TS are valid for a peak vessel fluence corresponding to 22 Effective Full Power Years (EFPY) of operation based on a rated thermal power of 3467 MWth. Conservative projections for operation at extended power uprate conditions (3988 MWth) indicate that NMP2 will reach the peak vessel fluence corresponding to the current P-T curves in December 2013. Therefore, to support continued plant operation, approval of the proposed license amendment is requested by November 22, 2013, with implementation within 60 days. Pursuant to 10 CFR 50.91(b)(1), NMPNS has provided a copy of this license amendment request, with Enclosure absent the proprietary information, to the appropriate state representative. There are no commitments in this submittal. to the Enclosure contains information considered to be proprietary as defined by 10 CFR 2.390. The Electric Power Research Institute (EPRI) and General Electric Hitachi (GEH), as the owners of the proprietary information, have executed the affidavits provided in Attachment 5 to the Enclosure detailing the reasons for withholding the proprietary information. On behalf of EPRI and GEH, NMPNS hereby requests that proprietary information in Attachment 6 to the Enclosure be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390. Information that is not considered proprietary is provided in Attachment 4 to the Enclosure. Should you have any questions regarding the information in this submittal, please contact John J. Dosa, Director Licensing, at (315) 349-5219. Very truly yours,
Document Control Desk November 21, 2012 Page 3 STATE OF NEW YORK TO WIT: COUNTY OF OSWEGO I, Ken Langdon, 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 and I believe it to be reliable. Subscribed and sworn before me, a Notary Public in and for the State of New York and County of (k/~ 5 ,this 2-1 day of NJc,/ercah ,2012. WITNESS my Hand and Notarial Seal: *. Notary Public My Commission Expires: SUsaM. Doran Date otaryPublic In fthState of Now York Date Oswego County Reg. No. 01D06029220 My Commission Expires 9/12/2013 KL/KJK
Enclosure:
Evaluation of the Proposed Change cc: Regional Administrator, Region 1, NRC Project Manager, NRC Resident Inspector, NRC A. L. Peterson, NYSERDA (without Attachment 6 to the Enclosure)
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
3.0 TECHNICAL EVALUATION
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions
5.0 ENVIRONMENTAL CONSIDERATION
6.0 REFERENCES
ATTACHMENTS
- 1. Nine Mile Point Unit 2 - Proposed Technical Specification Changes (Mark-up)
- 2. Nine Mile Point Unit 2 - Changes to Technical Specification Bases (Mark-up)
- 3. Nine Mile Point Unit 2 Pressure and Temperature Limits Report (PTLR) (Draft)
- 4. Responses to Requests for Additional Information - Non-Proprietary Version
- 5. Affidavits Justifying Withholding of Proprietary Information
- 6. Responses to Requests for Additional Information - Proprietary Version Nine Mile Point Nuclear Station, LLC November 21, 2012
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 1.0
SUMMARY
DESCRIPTION This evaluation supports a request to amend Renewed Facility Operating License NPF-69 for Nine Mile Point Unit 2 (NMP2). The proposed amendment would modify Technical Specification (TS) Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits," 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 1.1, "Definitions," and a new section addressing administrative requirements for the PTLR would be added to TS Section 5.0, "Administrative Controls." Relocation of the P-T limit curves to the PTLR is consistent with the guidance provided in NRC approved General Electric Hitachi Nuclear Engineering (GEH) Licensing Topical Report, NEDC-33178P-A, Revision 1, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," (Reference 1). This topical report uses the guidelines provided in NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," (Reference 2). 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," (Reference 3). New P-T limit curves have been developed for NMP2 that are valid for a peak internal diameter (ID) fluence of 9.60E+17 n/cm2 , which corresponds to a projected 32 Effective Full Power Years (EFPY) of core operation. Attachment 1 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 NMP2 TS Bases Control Program (TS 5.5.10). The draft NMP2 PTLR is provided in Attachment 3. Attachment 6 to the enclosure contains proprietary responses to the requests for additional information (RAI) related to this issue which have been customized for NMP2. Attachment 4 to the enclosure is a non-proprietary version. Affidavits requesting withholding of the proprietary information are included as Attachment 5 2.0 DETAILED DESCRIPTION ,2.1 Description of the Proposed Change The proposed change includes the following TS revisions:
" TS Section 1.1, Definitions: A new Definition, "Pressure and Temperature Limits Report," is added.
The wording for this definition is consistent with that in TSTF-419-A.
- TS Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits," The currently specified heatup and cooldown rates, temperature limits, and associated figures are replaced with a reference to the PTLR.
- TS Section 5.0, Administrative Controls: New Section 5.6.7, "Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR)," is added. The format and content of new Section 5.6.7 is 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 1 of 8
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE documents PTLR methodologies; and (3) requires that the PTLR and any revision or supplement thereto be submitted to the NRC. 2.2 Background The NRC safety evaluation transmitted by letter dated April 27, 2009 (Reference 4) discussed the Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report NEDC-33178P, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," Revision 1, dated January 2009 (Reference 5) 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. The NRC safety evaluation concluded that NEDC-33178P 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. NEDC-33178P was issued as a final report (-A) in July 2009 (Reference 1). The NRC safety evaluation dated April 27, 2009 (Reference 4) included a limitation and condition requiring the licensee to identify the report used to calculate the neutron fluence and document that the plant-specific neutron fluence calculation will be performed using an approved neutron fluence calculation methodology. NMP2 maintains an NRC approved RG 1.190 fluence monitoring program (Reference 6) and reviews actual fluence on a routine basis. The fluence projections have been confirmed to be conservative using the NMP2 fluence methods (Reference 7). TSTF-419-A (Reference 3) 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. Subsequent to TSTF-419-A approval, the NRC stated a new position in a letter to the TSTF, dated November 2, 2009 (Reference 8), describing concerns with TSTF-419-A. The NRC's concerns were related to not including the revision and date of the NRC-approved topic reports that document PTLR methodologies in STS 5.6.6, per the guidance in GL 96-03.
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 2 of 8
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 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. Generic Letter 96-03 allows plants to relocate their P-T curves and associated numerical limits (such as heatup and cooldown rates) from the plant TS to a PTLR, which is a licensee-controlled document. As stated in GL 96-03, during the development of the improved 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. 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. The purpose of NEDC-33178P-A, Revision 1, 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, and other associated numerical limits, thereby allowing BWR plants to adopt the PTLR option. The NMP2 P-T curves have been developed in accordance with the methodology and template in NEDC-33178P-A, Revision 1, as documented in the PTLR provided in Attachment 3. NEDC-33178P-A, Revision 1, does not include development or licensing of vessel fluence methods. The methodology used to calculate RPV neutron fluence values utilized in the development of the NMP2 P-T limit curves in the PTLR is in accordance with RG 1.190 methods. NMP2 maintains an NRC approved RPV neutron fluence calculation methodology documented by MPM Technologies, Inc; in MPM-40278 1, Revision 1, September 2003 (Reference 6), which has been previously approved by the NRC with the issuance of Nine Mile Point Unit 1 License Amendment No. 183 (Reference 7). The fluence values are in compliance with the recommendations of Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 9). 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:
- In TS Section 1.1, a new definition for "Pressure and Temperature Limits Report," is added.
- In TS Section 3.4.11, RCS Pressure and Temperature (PIT) Limits, LCO 3.4.11 is updated to include reference to the PTLR. The Surveillance Requirements (SRs) for Section 3.4.11 are updated as follows to reference the PTLR.
" SR 3.4.11.1.a: The reference to Figures 3.4.11-1, 3.4.11-2, 3.4.11-3, 3.4.11-4, and 3.4.11-5 is replaced with a reference to the PTLR.
" SR 3.4.11 .b: The currently stated heatup and cooldown rates of < 100°F in any one hour period are replaced with a reference to the PTLR.
" SR 3.4.11.1 c: The currently stated temperature change limit of < 20'F and the reference to Figure 3.4.11-2 or Figure 3.4.11-3 are replaced with a reference to the PTLR.
- SR 3.4.11.2: The reference to Figures 3.4.11-4 and 3.4.11-5 is replaced with a reference to the PTLR.
" SR 3.4.11.3 and SR 3.4.11.5: The currently stated temperature difference of_< 145'F is replaced with a reference to the PTLR.
- SR 3.4.11.4 and SR 3.4.11.6: The currently stated temperature difference of _<50'F is replaced with a reference to the PTLR.
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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE
" SR 3.4.11.7, SR 3.4.11.8, and SR 3.4.11.9: The currently stated temperature difference of> 70'F is replaced with a reference to the PTLR.
- The P-T limit curves, Figures 3.4.11-1 through 3.4.11-5, are deleted.
In TS Section 5.6, "Reporting Requirements," new Section 5.6.7, "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 3.4.11); (2) references the NRC-approved topical report that documents PTLR methodologies (i.e., NEDC-33178P-A); and (3) requires that the PTLR and any revision or supplement thereto be submitted to the NRC. In the Reference 10 letter, the NRC requested additional information concerning the Grand Gulf Nuclear Station (GGNS) Unit 1 license amendment request pertaining to the implementation of a PTLR. Responses to these questions for NMP2 are provided in Attachment 4 (non-proprietary) and Attachment 6 (proprietary).
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 NEDC-33178P-A for the preparation of NMP2 P-T limit curves ensures that the requirements of 10 CFR 50 Appendix G will be satisfied. 10 CFR 50 Appendix H provides criteria for the design and implementation of RPV material surveillance programs for operating light water reactors. NMP2 demonstrates its compliance with the requirements of 10 CFR 50 Appendix H through participation in the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) (Reference 13). Regulatory Guide 1.190 (Reference 9) describes methods and assumptions acceptable to the NRC for determining RPV neutron fluence. By letter dated October 27, 2003 (Reference 7), the NRC documented their determination that the benchmarking of NMP Unit 1 and Unit 2 neutron transport calculations 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 4 of 8
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE (3) Modify applicable sections of the TS accordingly. The NRC-approved methodology of NEDC-33178P-A has been adopted for preparation of NMP2 P-T limit curves. The NRC review documented in Reference 4 concluded that NEDC-33178P-A satisfies the criteria in Attachment 1 of GL 96-03 and provides adequate methodology for BWR licensees to calculate P-T limit curves. A draft PTLR has been prepared for NMP2 based on the methodology and template provided in NEDC-33178P-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 1 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 amendment requests 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).
- Nine Mile Point Unit 1 (License Amendment No. 204 issued by NRC letter dated January 21, 2010 -
ADAMS Accession No. ML093370002). 4.3 Significant Hazards Consideration Nine Mile Point Nuclear Station, LLC (NMPNS) is requesting revisions to the Nine Mile Point Unit 2 (NMP2) 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 NMP2 that are valid for a peak internal diameter fluence of 9.60E+17 n/cm2, which corresponds to a projected 32 Effective Full Power Years (EFPY) of core operation. 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:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No. The proposed amendment modifies 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 NEDC-33178P-A for the preparation of NMP2 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 5 of 8
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 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 amendment does not adversely affect accident initiators or precursors, and does 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 amendment does 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 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 amendment does 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 amendment does 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 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 amendment does 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. 6 of 8
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 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. Letter from D. Coleman, BWROG to Chief, Information Management Branch, USNRC, dated July 29, 2009, "Submittal of GE BWROG Topical Report NEDC-33178P-A, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves,"
ML092370486
- 2. Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," January 31, 1996
- 3. Technical Specification Task Force Traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," Revision 0, August 4, 2003, MLO 12690234
- 4. Letter from T. R. Bount, USNRC to D. Coleman, BWROG, dated April 27, 2009, "Final Safety Evaluation for Boiling Water Reactors Owners' Group Licensing Topical Report NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (TAC No. MD2693)," ML091100139
- 5. Boiling Water Reactors Owners' Group Licensing Topical Report NEDC-33178P, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," Revision 1, January 2009
- 6. MPM Technologies, Inc, MPM-402781, Revision 1, "Benchmarking of Nine Mile Point Unit 1 and Unit 2 Neutron Transport Calculations," September 2003
- 7. 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 Amendments Re: Pressure-Temperature Limit Curves and Tables (TAC NO. MB6687)," ML032760696 7 of 8
ENCLOSURE EVALUATION OF THE PROPOSED CHANGE
- 8. Letter from S. L. Rosenberg, USNRC to Technical Specification Task Force, dated November 2, 2009, "Technical Specification Task Force (TSFT) Traveler 363, Revision 0, "Revise Topic Report References in ITS 5.6.6, COLR," ML09215106
- 9. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Revision 0, March 2001
- 10. Letter from M. A. Krupa (Entergy Operations, Inc.) to USNRC, dated February 23, 2011, "Request for Additional Information Regarding Extended Power Uprate," ML110540540
- 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. MC 1758 and MC1759)"
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ATTACHMENT 1 NINE MILE POINT UNIT 2 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: 1.1-5 3.4.11-1 through 3.4.11-10 5.6-4 Nine Mile Point Nuclear Station, LLC November 21, 2012
Definitions 1.1 1.1 Definitions (continued) MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a. Described in Chapter 14, Initial Test Program of the FSAR;
- b. Authorized under the provisions of 10 CFR 50.59; or
- c. Otherwise approved by the Nuclear Regulatory Commission. ._
RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3988 MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. (continued) NMP2 1.1-5 Amendment 94,444
INSERT 1 (for TS page 1.1-5; New Definition) Pressure and The PTLR is the unit specific document that provides Temperature Limits the reactor vessel pressure and temperature limits, Report (PTLR) 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 5.6.7.
RCS P/T Limits 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 RCS Pressure and Temperature (P/T) Limits LCO 3.4.11 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculatig .o temperature requirements shall be maintained withi limitn - I Al ee-. APPLICABILITY: At all times. c ACTTNms 1~ I CONDITION REQUIRED ACTION COMPLETION TIME A. ------- NOTE------- A. 1 Restore parameter(s) 30 minutes to within limits. Required Action A.2
- shall be completed if AND this Condition is entered. A. 2 72 hours
----------------- Determine RCS is acceptable for continued operation.
Requirements of the LCO not met in MODE 1, 2, or 3.
~-..----------..----------i 12 hours B. Required Action and B.1 Be in MODE 3.
associated Completion AND Time of Condition A not met. B.2 36 hours Be in MODE 4. (continued) NMP2 3.4.11-1 3.4.11-1Amendment-9i-
RCS P/T Limits 3.4.11 l~rTTfl~Jc~ ntt, s~l'- frnntinu~d'~
' - - ..----. T
,nrT TnM(Z frontinuedl CONDITION REQUIRED ACTION COMPLETION TIME
-t C - --------- NOTE---------
Required Action C.2 C.1 Initiate action to Immediately shall be completed if restore parameter(s) this Condition is to within limits. entered.
------------------ AND Requirements of the C.2 Determine RCS is Prior to acceptable for entering MODE 2 LCD not met in other or 3 than MODES 1, 2, operation.
and 3. tfIWOVTI I ANCE REQUIREMENTS SURVEILLANCE FREQUENCY i SR 3.4.11.1 ------------------ NOTE---------------- Only required to be performed during RCS heatup and cooldown operations, and RCS system leakage and hydrostatic testing. Verify: 30 minutes RCS pressure and RCS temperature are witjlin the applicable limit-sspecified RCS temperature change duri I l~eakage and$ hydrost~atic tes S%. O - , 1 *- _I .. *__, (continued) NMP2 3.4.11-2 Amendment -0tI
RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY Verify RCS pressure and, RCS temperature are Once within SR 3.4.11.2 15 minutes within the in specified ap licables cri
.. 1-t, .. prior to control rod I withdrawal for the purpose of achieving criticality SR 3.4.11.3 -------------------NOTE----------------
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup. Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressqressel (RPV) coolant temperature prior to each startup of a recirculation pump I-SR 3.4.11.4 ---------------- NOTE---------------- Only required to be met in MODES 1, 2, 3, 1 and 4 during recirculation pump startup. Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be starteLakd the RPV coolant prior to each temperature is ___6" startup of a recircul ation pump 7_~4)-~_~-~_?~l-s jITL / (continued) NMP2 3.4.11-3 Amendment -
RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY
.1.
SR 3.4.11.5 ------------------ NOTE------------------ Only required to be met in single loop operation with THERMAL POWER < 30% RTP or the operating jet pump loop flow < 50% rated jet pump loop flow. Verify the difference between the bottom Once within 15 head coolant temperature dothe RPV minutes prior coolant temperature i .WT to an increase
< -]. _ in THERMAL POWER or an increase in jet pump loop flow SR 3.4.11.6 ------------------ NOTE------------------
Only required to be met in single loop operation when the idle recirculation loop is not isolated from the RPV, and with THERMAL POWER < 30% RTP or the operating jet pump loop flow < 50% rated jet pump loop flow. Verify the difference between the reactor Once within 15 coolant temperature in the recirculation minutes prior loop not in operation and the RPV coolant to an increase temperature is F in THERMAL POWER or an increase in jet pump loop flow SR 3.4.11.7 ------------------- NOTE----------------- Only required to be performed when tensioning the reactor vessel head boltinc studs. Verify reactor vessel flange and head 30 minutes flange temperatures are head (continued) NMP2 3.4.11-4 Amendment Si-
RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.4.11.8 -------------------- NOTE----------------- Not required to be performed until 30 minutes after RCS temperature ! 80*F in MODE 4. Verify reactor vessel flige.-apd head flange temperatures 30 minutes I SR 3.4.11.9 ------------------ NOTE---------------- Not required to be performed until 12 hours after RCS temperature < 90°F in MODE 4. Verify reactor vessel flan.-eada head 12 hours flange temperatures are I PT 1'ýý NMP2 3.4.11-5 Amendment-91
noc RnT Limft-s7
&4frc~ ŽCL~3Q 234411 1100 Cn 1000 E
0 0 900 E 800 700 600 ca W. cc 500 400 300 200 100 0 0 50 100 150 200 250 Downcomer Water Temperature (OF) Figure 3.4.11-1 (Page 1 of 1) Non-Nuclear System Leakage and Hydrostatic Testing Curve A ,,4 ,,4 Ima"~ii O.A. Ai4. t...4-4fl-.. 'A~,~:J
REC RPf ntt g~ cje~ e.4.1-1 1100 1000 a) E 900 800 700 0 0 600 cc 500 400 300 200 100 0 50 100 150 200 250 300 Downcomer Water Temperature (OF) Figure 3.4.11-2 (Page 1 of 1) Non-Nuclear Heatup Curve 3.4.11-:7 Amen~dment -1OE)
-~- ~%~- IROs PfT Liis 4-0 ~ ck~je~cQ 1100 1000 0
0 900 E (a 800 a) 4-700 U) 0)
- a. 600 (J@
500 0 400 a) 300 200 100 0 100 150 200 250 350 Downcomer Water Temperature (OF) Figure 3.4.11-3 (Page 1 of 1) Non-Nuclear Cooldown Curve
flOs~ -- ts ErT C)- 6 -- o9di-Q HMr2 3.4.11 9 A.R'49RPImqQF# CW n
RGS P4T Linits V J 1100 6 1000 E 0 ogo 900 E 800 00 0 700 a.
'~o 600 (3
cc 300 na" 50 100 150 200 250 300 Downcomer Water Temperature (OF) Figure 3.4.11-5 (Page 1 of 1) Nuclear Cooldown Curve A IMIV] 2 -1. .4 1 I i,- A ,- ,-,,-,t 91.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. I NMP2 5.6-4 Amendment ýJ, 9t
INSERT 2 (for TS page 5.6-4; New Section 5.6.7) 5.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
- a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and system 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 3.4.11, "RCS Pressure and Temperature (P/T) Limits."
- 2. Surveillance Requirement 3.4.11, "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 documents:
- 1. NEDC-33178P-A, Revision 1, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," dated June 2009.
- 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 2 CHANGES TO TECHNICAL SPECIFICATION BASES (MARK-UP) The current versions of the following Technical Specifications Bases pages have been marked-up by hand to reflect the proposed changes. These Bases pages are provided for information only. B 3.4.11-1 B 3.4.11-2 B 3.4.11-3 B 3.4.11-4 B 3.4.11-6 B 3.4.11-7 B 3.4.11-9 B 3.4.11-10 Nine Mile Point Nuclear Station, LLC November 21, 2012
RCS P/T Limits B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.11 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, _ within the design assumptions and the stress limits for cyclic operation. "'I1
- jMSS&(f_ ote -Speeifie-*T contains P/T limit curves for heatup, cooldown, n.
"T-..,r* )d-,*
.,*,* *r* system leakage and hydrostatic testing, and criticality, and also limits the maximum rate of change of reactor coolant temperature.
Z-Sis ssemlaaead yrsai esig'n riiaiytn also imits olimiting a .juztedrcfecrcnce tecmperaturfe for the 22 effective ftll jpuwe,
*TL~as as dfincd ,,n Table 4.1 ,,f Refe.. e-....7 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 LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel. 10 CFR 50, Appendix G (Ref. 1), requires the establishment of PIT limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section III, Appendix G (Ref. 2). The actual shift in the RTNDT of the vessel material will be established periodically by evaluating the irradiated reactor vessel material data provided as part of the Boiling Water Reactor Vessel and ternals Project (BWRVIP) Integrated Surveillance Program W . - 1 2)4 .- in accordance with 10 CFR 50, Appendix H (Ref. 4).1TFle oper=-tan P/T limit curves will be adjusted, (continued) NMP2 B 3.4.11 -1 Revision 0, !0, 14, (A4 4Q) j
RCS P/T Limits B 3.4.11 BASES BACKGROUND as necessary, based on the evaluation findings and the (continued) recommendations of Reference 5. The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions. The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. The P/T criticality limits include the Reference 1 requirement that they be at least 40'F above the non-critical heatup curve or the non-critical cooldown curve and not lower than the minimum permissible temperature for the system leakage and hydrostatic testing. Reference I also allows boiling water reactors to operate with the core critical below the minimum permissible temperature allowed for the inservice hydrostatic pressure test (i.e., system leakage and hydrostatic testing) when the water level is within the normal range for power operation and the pressure is less than 20% of the preservice system hydrostatic test pressure (for NMP2, this pressure is 312 psig). Under these conditions, the minimum temperature is 60'F above the RT.DT of the closure flange regions which are stressed by the bolt preload (for NMP2, this temperature is 70'F). regio This P/T dclineated on the P/T eurves (Figures 3.4'.1"4 ad
-3.4.11 5) using Gross hatching'.
I. The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading, to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits. (continued) NMP2 B 3.4.11-2 Revision-&,
RCS P/T Limits B 3.4.11 BASES (continued) APPLICABLE The PIT limits are not derived from Design Basis Accident (DBA) SAFETY ANALYSES analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.
*References 7 and 8 provide the basis for the curves and limits required by this Specification. Since the PIT limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of Reference 9. LCO The elements of this LCO are:
- a. I in r I ke9't"TLl - heatup RCS ff) I
+ke- r-n-l",
The temperature difference between the reactor vessel bottom he coolant and the reactor pressure vessel (RPV) coolant is dduring recirculation pump startup, and during increases in THERMAL POWER or jet pump loop flow while in single loop I operation at low THERMAL POWER or jet pump loop flow; The temperature difference between the reactor coolant in the I respective recirculation loop and in the reactor vessel i'*(* during recirculation pump startup, and during increases in THERMAL POWER or jet pump loop flow while in single loop operation at low THERMAL POWER or jet pump loop flow;
- d. RCS pressurn
" " " -: irJ
. emp~ratir.
. w. A -44 a withi W. Q,'. Aq4 C the criticality limits specified in ,ues 0.4.11-4 a,, 3.4. 15} prior to achieving criticality; andu (continued)
NMP2 B 3.4.11-3 Revision 0,-'--
RCS P/T Limits B 3-4.11 BASES LCO e. The reactor vessel flane and the head flange (continued) temperatures are'%-8 vessel head bolting s uds. ben tensioning the reactor I These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure. The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and system leakage and hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curve Violation of the lim fIplaces the reactor vessel outside of the bounds of the stress analyses and can increase stresses I in other RCS components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existence, size, and orientation of flaws in the vessel material.
APPLICABILITY The potential for violating a P/T limit exists at all times. For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core. ACTIONS A.1 and A.2 Operation outside the P/T limits while in MODE 1, 2, or 3 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. (continued)% NMP2 B 3.4.11-4 Revision-@.*"/
RCS P/T Limits B 3.4.11 BASES ACTIONS B.1 and B.2 (continued) Pressure and temperature are reduced by bringing the plant to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. r~C.1 and C.2 S,, L Operation outside the in other than MODES 1, 2, and 3 (including defueled con 1 ions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without'delay and continued until the limits are restored. Besides restoring'the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200°F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline. Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. SURVEILLANCE SR 3.4.11.1 ) , ,W#e TLI REQUIREMENTS Verification that operation is within Llimits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly (continued) NMP2 B 3.4.11-6 Revision-{)-
RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.1 (continued) REQUIREMENTS d/7 _7 increments, 30 minutes
~~~minor deviations. The limitsfpermits * .. 12' of 1-,nd correction
'*" .- 3. A 4 .1,an-d --- j.3,3-."41- 6 re met when opera ion is to the right o the applicable limit curve. Fv, Fignties 3.4.11 4landE 3.1.11 F5,the GrOGS hotched porti8n o~f the cuvei3oly applieable when FeaeteF wae leve is inthe 1 OFFMal, pawetr operation range (1:78.3 iflehe3 to '167.3 inoheos, norrofw rangge mineatoin).
Surveillance for heatup, cooldown, or system leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied. This SR has been modified by a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and system leakage and hydrostatic testing. SR 3.4.11.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before
-----------'--- withdrawing control rods that will make the reactor
-q i+TL- critical. The " -are met when operation is to the right of the applicable limit curve. T.... h ch*, poion o-f tho cu..11"i only rc, s ......
applieeblo whoen roaetor water level is ;in the normafll, peweir opefrtinl' ron* ig. (1,78.3 inIohos to 1837.3 inelges, nmarIow r.nge. mieatiaftý Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. SR 3.4.11.3 and SR 3.4.11.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of (continued) NMP2 B 3.4.11-7 Revision I
RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.5 and SR 3.4.11.6 (continued) REQUIREMENTS rate > 50% of rated jet pump loop flow. Therefore, SR 3.4.11.5 and SR 3.4.11.6 have been modified by a Note that requires the Surveillance to be met only under these conditions. The Note for SR 3.4.11.6 further limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is isolated from the RPV since the, water in the loop cannot be introduced into the remainder of the Reactor Coolant System. SR 3.4.11.7. SR 3.4.11.8. and SR 3.4.11.9
*'Limits on the reactor vessel flancle and head fla I
* *temperatures are generally bounded by et he , Timit-*
+k PL>r during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
The flange temperatures must be verified to be above the limits within 30 minutes before and every 30 minutes thereafter while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature 5 800F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature 5 90'F, monitoring of the flange temperature is required every 12 hours to ensure the temperatures are within the specified limits. The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour Frequency is reasonable based on the rate of temperature change possible at these temperatures. SR 3.4.11.7 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.11.8 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature < 80'F in MODE 4. SR 3.4.11.9 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature
< 90°F in MODE 4.
(continued) NMP2 B 3.4.11-9 Revisione-,
RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.7, SR 3.4.11.8, and SR 3.4.11.9 (continued) REQUIREMENTS The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the specified limits. REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.
- 3. (Deleted)
- 4. 10 CFR 50, Appendix H.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.
- 7. 'Repcrt Ne. MPM 502840, "Prczsurc Tempefraturo Operating etirvesfoy NII Ht 2," July 2003.
I1 Mile Point UJ,
- 8. ASME Co~de Case N-640, "Alternate Referenee Fraetture Tou9, i-ess Fou DevelopmIent of P-T, _,,,,t C.u.ves Section XI, DivsioV 1**J."
- 9. 10 CFR 50.36(c)(2)(ii).
- 10. USAR, Section 15.4.4.
. BWR*F'-8, "WR Vsseia~nd rternals PoeiB¢ I / GR"BWR
,-1-2. Vessel and Internals Project, Updated BWR Inte radSurveillance Program (ISP) Implementation Plan,"
e! ~e{-Pv~rZO T- V,-\ C ~ J~&~Z~J 6~~A NL6c -r33qi'fP ,ŽcuýO"S"Of'K NMP2 B 3.4.11-10 Revision 0, 4-0, 44--4
ATTACHMENT 3 NINE MILE POINT UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (DRAFT) Nine Mile Point Nuclear Station, LLC November 21, 2012
NMP2 Pressure and Temperature Limits Report Table of Contents Section Page 1.0 Purpose 1 2.0 Applicability 1 3.0 Methodology 1 4.0 Operating Limits 3 5.0 Discussion 3 6.0 References 6 Figure 1 Bottom Head and Composite P-T Curves for Pressure 7 Test [Curve A] Figure 2 Bottom Head and Composite P-T Curves for Core Not 8 Critical [Curve B] Figure 3 Composite P-T Curves for Core Critical [Curve C] 9 Table 1 Data Table for Composite P-T Curve 10 Appendix A Reactor Vessel Material Surveillance Program 15 Appendix B Supporting Information 16 Appendix C Checklist 25 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report 1.0 Purpose The purpose of the Nine Mile Point Nuclear Station Unit 2 (NMP2) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:
- 1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing;
- 2. RCS Heatup and Cooldown rates;
- 3. Reactor Pressure Vessel (RPV) to RCS coolant AT requirements during Recirculation Pump startups;
- 4. RPV bottom head coolant temperature to RPV coolant temperature AT requirements during Recirculation Pump startups;
- 5. RPV head flange bolt-up temperature limits.
This report has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.7, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)." 2.0 Applicability This report is applicable to the NMP2 RPV for up to the peak fluence reported in Appendix B and Reference 6.1. The peak fluence (peak ID fluence = 9.60E+17 n/cm 2 is estimated at the projected 32 EFPY) assumes an equilibrium core design, using GE14 fuel, and operating at the 3988 MWth rating. The NMP2 Regulatory Guide (RG) 1.190 fluence monitoring program monitors actual fluence and projects the validity of the PT curves. EFPY is used only as an estimate of the projected fluence. The following TS are affected by the information contained in this report: TS 3.4.11 RCS Pressure and Temperature (P/T) Limits; 3.0 Methodology The limits in this report were derived from the NRC-approved methods listed in TS 5.6.7, using the specific revisions listed below:
- 1. The Pressure-Temperature curves in this report were obtained from the Reference 6.1, which is based on conservative RG 1.190 fluence projections assuming an equilibrium core, GE14 fuel, and a total accumulated fluence assuming 32 EFPY (32 EFPY projections apply 3988 MWth rated power). NMP2 maintains a Reg.
Page 1 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Guide 1.190 fluence monitoring program and reviews the actual fluence on a routine basis, as discussed in USAR Section 4.1.4.5 and Appendix C, Section C.2.1.2. The detailed discussion of the NMP2 Reg. Guide 1.190 methods is documented in Reference 6.2. NRC approval of the Unit 2 neutron fluence calculational methodology is documented in Reference 6.3. The fluence projections in Reference 6.1 have been confirmed to be conservative using the NMP2 fluence methods, as documented in Reference 6.4.
- 2. The pressure and temperature limits were calculated per GEH P-T Curve Licensing Topical Report, NEDC-33178P-A, Revision 1, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," Report for BWR Owners' Group, Sunol, California, (GEH Proprietary),
June 2009, approved in Reference 6.5.
- 3. This revision of the pressure and temperature limits is to incorporate the following changes:
This is the initial issue of the PTLR for NMP2, which also incorporates updated Pressure-Temperature curves up to peak vessel internal diameter (ID) fluence of 9.60E+17 n/cm 2 (approximately 32 EFPY). Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the Reactor Pressure Vessel (RPV), or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance, in accordance with TS Section 5.6.7. Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR. Since NMP2 surveillance capsule program is the ISP program, new surveillance data through the Integrated Surveillance Program (ISP) does not represent new data that requires NRC review prior to incorporation into the PTLR. Page 2 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report 4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region. Complete P-T curves were developed for the estimated fluence based on 32 EFPY assuming 3988 MWth rating. The P-T curves are provided in Figures 1 through 3 (for a peak ID fluence of 9.60E+17 n/cm 2, approximately 32 EFPY). A tabulation of the P-T curves is also included in Table 1. The previous evaluation for Pressure-Temperature limits for up to 22 EFPY is contained in Reference 6.6. The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown (core not critical), referred to as Curve B; and (c) core critical operation, referred to as Curve C. Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing (Figure 1): Curve A): < 20 OF/hour. Normal Operating Heatup and Cooldown rate limit (Figures 2 and 3): Curve B - Non-Nuclear Heating and Curve C - Nuclear Heating): < 100 OF/hour. 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: < 50 OF. RPV flange and adjacent shell temperature limit: > 70 OF. 5.0 Discussion No new computer codes have been used in the development of Pressure-Temperature curves. The method for determining the initial RTNDT for all vessel materials is that defined in Section 4.1.2 of NEDC-33178P-A, Revision 1, approved for use in Reference 6.5. Initial RTNDT values for all vessel materials considered are presented in tables in this PTLR. Surveillance capsule material data to represent the NMP2 vessel is available from the Integrated Surveillance Program (ISP). Since NMP2 is not a host plant in the ISP program, the ISP program representative materials are evaluated to confirm applicability Page 3 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report of RG 1.99 position 1.1. The ISP data review is shown in Appendix B. The NMP2 representative plate is not the same as the target plate material and is therefore provided for information only and is not considered in development of the PT curves. The NMP2 representative weld is not the same as the target weld material. However, as the same heat is used in beltline axial welds, the ISP weld material is considered in the development of the PT curves. This material is not the limiting material with respect to the PT curves; the NMP2 plant-specific plate heat C3147-1, has the limiting ART. For NMP2, there are three thickness discontinuities in the vessel. One discontinuity is between the bottom head torus and dollar plate. One discontinuity is between the bottom head torus and Shell #1. One discontinuity is between the transition in the upper Shell (Shell #3 to Shell #4). The thickness discontinuities do not cause a change in the RTNDT. The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG 1.99) provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. The vessel beltline copper and nickel values (except for the N6 and N12 nozzles) were obtained from Reference 6.6 PT curve limits report. Chemistries for the surveillance materials evaluated in Appendix B of this report were obtained from the ISP. The LPCI (N6) Nozzle and the Water Level Instrumentation (N12) Nozzle are now included in the extended beltline region. For the N6 nozzle forging, CMTR's were located which contained Cu (0.07%) and Ni (0.86%) values. For the N6 nozzle welds, CMTRs were located that contained Cu and Ni values as shown in the ART table. For the Water Level Instrumentation (N12) Nozzle, the CMTR's do not contain the Cu and Ni content. Since plant-specific information regarding the Cu and Ni content for this material is not available, the evaluation was performed based on a bounding estimate for forgings fabricated from SA508 Class 1 material. This was defined based on a search of available BWR vessel purchase records for SA508 Class 1 materials. Representative values for Cu and Ni content were developed using Mean + 1 sigma. The P-T curves for the non-beltline region were conservatively developed for a Boiling-Water Reactor Product Line 6 (BWR/6) with nominal inside diameter of 251 inches. The analysis is considered appropriate for NMP2, since the plant specific geometric values are bounded by the generic analysis for a large BWR/6. The generic value was adapted Page 4 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report to the conditions at NMP2 using plant-specific RTNDT values for the reactor pressure vessel. The value of R/t 112 for NMP2 = 126.6875 / (7.19)1/2 = 47.3 inch1 /2. The peak RPV ID fluence used in the P-T curve evaluation for 32 EFPY is calculated using methods that comply with the guidelines of RG 1.190, (References 6.1, 6.2, 6.3 and 6.4). Appendix B lists the peak fluence levels used for the PT curves. The P-T curves forthe heatup and cooldown operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and 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, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits. For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of < 100°F/hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of _<20°F/hr must be maintained. The P/T limits and corresponding heatup/cooldown rates of either Curve A or B may be applied while achieving or recovering from test conditions. Curve A applies during pressure testing and when the limits of Curve B cannot be maintained. For NMP2, the plate heat C3147-1 is the limiting material for the beltline region. Peak fluence values used in the development of the Pressure-Temperature curves are identified in Appendix B. The limiting ART for the beltline LPCI N6 and Water Level Instrumentation N12 nozzle forgings and welds are also considered in the development of the beltline PT curves. In order to ensure that the limiting vessel discontinuity has been considered in the development of the P-T curves, the methods in Sections 4.3.2.1 and 4.3.2.2 of NEDC-Page 5 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report 33178P-A, Revision 1, approved for use in Reference 6.5, for the non-beltline and beltline regions, respectively, are applied. 6.0 References 6.1 GEH Report NEDC-33414P, Revision 1, "Pressure-Temperature Curves for Constellation Generation Group Nine Mile Point Nuclear Station Unit 2",'October 2012 6.2 Benchmarking of Nine Mile Point Unit 1 and Unit 2 Neutron Transport Calculations, MPM-402781 Revision 1, September 2003. 6.3 NRC Letter to 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)." 6.4 Neutron Transport Analysis for Nine Mile Point Unit 2 Uprate, MPM-1 108779, Rev. 1, December 2008 6.5 Letter from T. R. Bount, USNRC to D. Coleman, BWROG, "Final Safety Evaluation for Boiling Water Reactors Owners" Group Licensing Topical Report NEDC-33178P, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves (TAC No. MD2693)," April 27, 2009 6.6 Pressure-Temperature Operating Curves for Nine Mile Point Unit 2, MPM-502840, dated July 2003. 6.7 "BWRVIP-135, Revision 2: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations", 1020231, EPRI, Palo Alto, CA, October 2009 (EPRI Proprietary Information) 6.8 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 MC1759)" 6.9 Nine Mile Point Unit 2 3-Degree Surveillance Capsule Report, MPM-1200676, December, 2000. Page 6 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Figure 1 - Bottom Head and Composite P-T Curves for Pressure Test [Curve A] 1400 1300 INITIAL RTndt VALUES ARE 1200 0°F FOR BELTLINE, 5°F FOR UPPER VESSEL, AND 1100 10'F FOR BOTTOM HEAD ,S1000 a w BELTLINE CURVES = 900 ADJUSTED AS SHOWN: 0 EFPY SHIFT ('F) I- 32 51 L 800 w W 700 0 HEATUP/COOLDOWN RATE OF COOLANT w 600 < 2 0 °F/HR Z 500 w 400 Lu w(L 300 200 100 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) Page 7 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Figure 2 - Bottom Head and Composite P-T Curves for Core Not Critical [Curve B] 1400 1300 1200 1100 U) "B1000 z 0 900 w 800 0 oe 700 Lu600 I- HEATUP/COOLDOW N RATE OF COOLANT
< 100°F/HR U)
Z Lu 500 400 300 200 100 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) Page 8 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Figure 3 - Composite P-T Curves for Core Critical [Curve C] 1400 INITIAL RTndt VALUES 1300 ARE 0°F FOR BELTLINE, BREAK AT 1250 PSIG 5'F FOR UPPER VESSEL, 1200 AND 24.6°F FOR BOTTOM HEAD 1100 1000 BELTLINE CURVE ADJUSTED AS SHOWN:
- a. EFPY SHIFT ('F)
- 0. 900 32 51 800 0 HEATUP/COOLDOWN RATE OF COOLANT I- 700 < 100°F/HR ac Z I.
600 U) U) 500 Lu w I- 400 300 200
-BELTLINE AND NON-100 BELTLINE LIMITS Minimum Criticality Temperature 70*F 0
0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F) Page 9 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Table 1 - NMP2 Tabulation of Composite P-T Curves - Required Metal Temperature with Required Coolant Temperature Rate at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 1, 2 & 3 BOTTOM UPPER RPV & BOTTOM UPPER RPV & HEAD BELTLINE HEAD BELTLINE LIMITING PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) (*F) (OF) (*F) (*F) 0 68.0 70.0 68.0 70.0 70.0 10 68.0 70.0 68.0 70.0 70.0 20 68.0 70.0 68.0 70.0 70.0 30 68.0 70.0 68.0 70.0 70.0 40 68.0 70.0 68.0 70.0 70.0 50 68.0 70.0 68.0 70.0 70.0 60 68.0 70.0 68.0 70.0 70.0 70 68.0 70.0 68.0 70.0 70.0 80 68.0 70.0 68.0 70.0 70.0 90 68.0 70.0 68.0 70.0 70.0 100 68.0 70.0 68.0 70.0 70.0 110 68.0 70.0 68.0 70.0 71.9 120 68.0 70.0 68.0 70.0 75.7 130 68.0 70.0. 68.0 70.0 79.2 140 68.0 70.0 68.0 70.0 82.4 150 68.0 70.0 68.0 70.0 85.2 160 68.0 70.0 68.0 70.0 87.9 170 68.0 70.0 68.0 70.0 90.5 180 68.0 70.0 68.0 70.0 92.9 190 68.0 70.0 68.0 70.0 95.2 200 68.0 70.0 68.0 70.0 97.3 210 68.0 70.0 68.0 70.0 99.3 220 68.0 70.0 68.0 70.0 101.3 230 68.0 70.0 68.0 70.0 103.1 240 68.0 70.0 68.0 70.0 104.9 250 68.0 70.0 68.0 70.0 106.6 260 68.0 70.0 68.0 70.0 108.2 270 68.0 70.0 68.0 70.0 109.8 280 68.0 70.0 68.0 71.3 111.3 290 68.0 70.0 68.0 72.8 112.8 Page 10 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Table 1 - NMP2 Tabulation of Composite P-T Curves - , Required Metal Temperature with Required Coolant Temperature Rate at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 1, 2 & 3 BOTTOM UPPER RPV & BOTTOM UPPER RPV & HEAD BELTLINE HEAD BELTLINE LIMITING PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (WF) ('F) (°F) (*F) (*F) 300 68.0 70.0 68.0 74.2 114.2 310 68.0 70.0 68.0 75.5 115.5 312.5 68.0 70.0 68.0 75.9 115.9 312.5 68.0 100.0 68.0 130.0 170.0 320 68.0 100.0 68.0 130.0 170.0 330 68.0 100.0 68.0 130.0 170.0 340 68.0 100.0 68.0 130.0 170.0 350 68.0 100.0 68.0 130.0 170.0 360 68.0 100.0 68.0 130.0 170.0 370 68.0 100.0 68.0 130.0 170.0 380 68.0 100.0 68.0 130.0 170.0 390 68.0 100.0 68.0 130.0 170.0 400 68.0 100.0 68.0 130.0 170.0 410 68.0 100.0 68.0 130.0 170.0 420 68.0 100.0 68.0 130.0 170.0 430 68.0 100.0 68.0 130.0 170.0 440 68.0 100.0 68.0 130.0 170.0 450 68.0 100.0 68.0 130.0 170.0 460 68.0 100.0 68.0 130.0 170.0 470 68.0 100.0 68.0 130.0 170.0 480 68.0 100.0 68.0 130.0 170.0 490 68.0 100.0 68.0 130.0 170.0 500 68.0 100.0 68.0 130.0 170.0 510 68.0 100.0 68.0 130.0 170.0 520 68.0 100.0 68.0 130.0 170.0 530 68.0 100.0 68.0 130.0 170.0 540 68.0 100.0 68.0 130.0 170.0 550 68.0 100.0 68.0 130.0 170.0 560 68.0 100.0 68.0 130.0 170.0 570 68.0 100.0 68.0 130.0 170.0 Page 11 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Table 1 - NMP2 Tabulation of Composite P-T Curves - Required Metal Temperature with Required Coolant Temperature Rate at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 1, 2 & 3 BOTTOM UPPER RPV & BOTTOM UPPER RPV & HEAD BELTLINE HEAD BELTLINE LIMITING PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) ('F) (*F) (°F) (*F) 580 68.0 100.0 68.0 130.0 170.0 590 68.0 100.0 68.0 130.0 170.0 600 68.0 100.0 68.0 130.0 170.0 610 68.0 100.0 68.0 130.0 170.0 620 68.0 100.0 68.0 130.0 170.0 630 68.0 100.0 68.0 130.0 170.0 640 68.0 100.0 68.0 130.0 170.0 650 68.0 100.0 68.0 130.0 170.0 660 68.0 100.0 68.0 130.0 170.0 670 68.0 100.0 68.0 130.0 170.0 680 68.0 100.0 68.7 130.0 170.0 690 68.0 100.0 69.9 130.0 170.0 700 68.0 100.0 71.0 130.0 170.0 710 68.0 100.0 72.2 130.0 170.0 720 68.0 100.0 73.3 130.0 170.0 730 68.0 100.0 74.4 130.0 170.0 740 68.0 100.0 75.5 130.0 170.0 750 68.0 100.0 76.6 130.0 170.0 760 68.0 100.0 77.6 130.0 170.0 770 68.0 100.0 78.6 130.0 170.0 780 68.0 100.0 79.6 130.0 170.0 790 68.0 100.0 80.6 130.0 170.0 800 68.0 100.0 81.5 130.0 170.0 810 68.0 100.0 82.5 130.0 170.0 820 68.0 100.0 83.4 130.0 170.0 830 68.0 100.0 84.3 130.0 170.0 840 68.0 100.0 85.2 130.0 170.0 850 68.0 100.0 86.0 130.0 170.0 860 68.0 100.0 86.9 130.0 170.0 870 68.0 100.0 87.7 130.0 170.0 Page 12 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Table 1 - NMP2 Tabulation of Composite P-T Curves - Required Metal Temperature with Required Coolant Temperature Rate at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 1, 2 & 3 BOTTOM UPPER RPV & BOTTOM UPPER RPV & HEAD BELTLINE HEAD BELTLINE LIMITING PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) (*F) (°F) (*F) (*F) 880 68.0 100.0 88.6 130.0 170.0 890 68.0 100.0 89.4 130.0 170.0 900 68.0 100.0 90.2 130.0 170.0 910 68.0 100.0 91.0 130.0 170.0 920 68.0 100.0 91.7 130.0 170.0 930 68.0 100.0 92.5 130.0 170.0 940 68.0 100.0 93.3 130.0 170.0 950 68.0 100.0 94.0 130.0 170.0 960 68.0 100.0 94.7 130.0 170.0 970 68.0 100.0 95.5 130.0 170.0 980 68.0 100.0 96.2 130.0 170.0 990 68.0 100.0 96.9 130.0 170.0 1000 68.0 100.0 97.6 130.0 170.0 1010 68.0 100.0 98.2 130.0 170.0 1020 68.0 100.0 98.9 130.0 170.0 1030 68.0 100.0 99.6 130.0 170.0 1035 68.0 100.0 99.9 130.0 170.0 1040 68.0 100.0 100.2 130.0 170.0 1050 68.0 100.0 100.9 130.0 170.0 1055 68.0 100.0 101.2 130.0 170.0 1060 68.0 100.0 101.5 130.0 170.0 1070 68.0 100.0 102.1 130.0 170.0 1080 68.0 100.0 102.8 130.0 170.0 1090 68.6 100.0 103.4 130.0 170.0 1100 69.2 100.0 104.0 130.0 170.0 1105 69.6 100.0 104.3 130.0 170.0 1110 69.9 100.0 104.6 130.0 170.0 1120 70.6 100.0 105.2 130.0 170.0 1130 71.2 100.0 105.8 130.0 170.0 1140 71.9 100.0 106.3 130.0 170.0 Page 13 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Table 1 - NMP2 Tabulation of Composite P-T Curves - Required Metal Temperature with Required Coolant Temperature Rate at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 1, 2 & 3 BOTTOM UPPER RPV & BOTTOM UPPER RPV & HEAD BELTLINE HEAD BELTLINE LIMITING PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) (°F) (°F) (*F) (oF) 1150 72.5 100.0 106.9 130.0 170.0 1160 73.1 100.0 107.5 130.0 170.0 1170 73.8 100.0 108.0 130.0 170.0 1180 74.4 100.0 108.6 130.0 170.0 1190 75.0 100.0 -109.1 130.0 170.0 1200 75.6 100.4 109.7 130.0 170.0 1210 76.2 101.0 110.2 130.0 170.0 1220 76.8 101.7 110.8 130.0 170.0 1230 77.3 102.3 111.3 130.0 170.0 1240 77.9 102.9 111.8 130.0 170.0 1250 78.5 103.5 112.3 130.0 170.0 1260 79.0 104.1 112.8 130.3 170.3 1270 79.6 104.7 113.3 130.6 170.6 1280 80.1 105.2 113.8 130.8 170.8 1290 80.7 105.8 114.3 131.1 171.1 1300 81.2 106.4 114.8 131.4 171.4 1310 81.7 106.9 115.3 131.8 171.8 1320 82.3 107.5 115.8 132.2 172.2 1330 82.8 108.0 116.2 132.7 172.7 1340 83.3 108.6 116.7 133.1 173.1 1350 83.8 109.1 117.2 133.5 173.5 1360 84.3 109.7 117.6 134.0 174.0 1370 84.8 110.2 118.1 134.4 174.4 1380 85.3 110.7 118.5 134.8 174.8 1390 85.8 111.2 119.0 135.2 175.2 1400 86.3 111.7 119.4 135.7 175.7 Page 14 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix A Reactor Vessel Material Surveillance Program In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, the first surveillance capsule was removed from the NMP2 reactor vessel after Cycle 7, in March 2000, and tested. 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 materials within the core beltline region. The methods and results are presented in Reference 6.9, as required by 10 CFR 50, Appendices G and H. As described in the NMP2 Updated Safety Analysis Report (USAR) Section 5.3.1.6, Material Surveillance, the BWRVIP Integrated Surveillance Program (ISP) will determine the removal schedule for the remaining two (2) NMP2 surveillance capsules. Per the BWRVIP ISP, NMP2 is not a host plant; all remaining surveillance capsules are classified as "Standby." The NMP2 material surveillance program is administered in accordance with the BWRVIP ISP. The ISP combines the US BWR surveillance programs into a single integrated program. This program uses similar heats of materials in the surveillance programs of BWRs to represent the limiting materials in other vessels. It also adds data from the BWR Supplemental Surveillance Program (SSP). NMP2 maintains a Reg. Guide 1.190 fluence monitoring program and reviews the actual fluence on a routine basis, as discussed in USAR Section 4.1.4.5 and Appendix C, Section C.2.1.2. Page 15 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Reactor Pressure Vessel P-T Curve Supporting Plant-Specific Information Figure of NMP2 Reactor Pressure Vessel TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL #4 SHELL #3 LPCI SHELL #2 TOP OF ACTIVE FUEL (TAF)366.31 er Level rumentation Nozzle BOTTOM OF ACTIVE FUEL (BAF)216.31 Page 16 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Initial RTNDT Values for RPV Materials Plate and Flange Materials Drop Test 1 Component Heat or Heat / Flux I Lot Temp Charpy Energy ' (Ts0o-60) Weight RTNDT (CF) (ft-lb) (*F) NDT (*F) Top Head & Flange. Shell Flange Mark 27-1 48D 1072-1-1 /4813121-1-1 40 67 131 110 10 -20 10 Top Head Flange Mark 32-1 49B 168-1-1 /49D 161-1-1 30 94 98 77 0 -30 0 Top Head Dollar Mark 36-2 A0678-1 40 53 62 53 -20 -20 -20 Top Head Torus Plates _________ ____ _ _ __w_ Marks 36-1-1. 36-1-2, 36-1-3 C2325-2 60 55 55 52 0 -20 0 Marks 36-1-4. 36-1-5, 36-1-6 C2480-2 30 62 50 66 -30 -40 -30 Shell Courses _Upper Shell Plates rvMrk 24-1-1 C3193-1 60 59 56 58 0 -10 0 Mark 24-1-2 C3192-2 70 61 54 63 10 -10 10 Mark 24-1-3 C3192-1 70 58 51 72 10 -10 10 Upper Intermediate Plates Mark 23-1-1 C3121-1 70 52 50 50 10 -20 10 Mark 23-1-2 03138-1 50 56 56 50 -10 -50 -10 lMrk 23-1-3 B6991-1 20 70 64 56 -40 -40 -40 Lower-Intermediate Plates Mark 22-1-1 C3065-1 50 70 50 50 -10 -30 -10 lMbrk 22-1-2 C3121-2 60 50 53 50 0 -30 0 Mark 22-1-3 C3147-1 60 50 50 52 0 -20 0 Lower Shell Plates Mark 21-1-1 C3147-2 60 52 50 50 0 -20 0 Mark 21-1-2 C3066-2 40 55 52 51 -20 -30 -20 Mark 21-1-3 C3065-2 70 51 53 51 10 -10 10 Bottom Head Bottom Head Dollar Mark 13-1 B6803-2 50 55 58 52 -10 -10 -10 Mark 13-2 C2944-2 40 62 58 60 -20 -20 -20 Mark 13-3 B6803-1 70 68 56 54 10 -30 10 Bottom Head Torus Plates Marks 13-4, 13-5, 13-6 C3073-1 60 51 54 53 0 -20 0 Marks 13-7, 13-8, 13-9 C3073-2 70 51 57 55 10 -10 10 Skirt Knuckle Mark 9-1-1, 9-1-2, 9-1-3, 9-1-4 C3957-3 30 55 54 60 -30 -20 -20 (1) Minimum Charpy values are used in these calculations. Page 17 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Initial RTNDT Values for RPV Materials, Continued Nozzle Materials Test Drop Component Heat or Heat / Flux / Lot Temp Charpy Energy" (T05or60) Weight RTNDT (CF) (ft-lb) (*F) NDT (*F) (*F) N1 Recirculation Outlet Nozzle Ibrk 49-1-1 Q2QL1W / 63A-1 40 102 109 91 -20 -20 -20 Mark 49-1-2 Q2QL1W / 63A-2 40 95 106 82 -20 -20 -20 N2 Recirculation Inlet Nozzle Mark 52-1-1 Q2QL2W / 62A-1 30 93 97 94 -30 -20 -20 Mark 52-1-2 Q2Q67W / 62A-2 40 95 103 105 -20 -20 -20 Mark 52-1-3 Q2Q67W / 62A-3 40 82 81 101 -20 -20 -20 Mark 52-1-4 Q2Q67W / 62A-4 40 67 64 77 -20 -20 -20 Mark 52-1-5 Q2Q67W / 62A-5 40 62 67 109 -20 -20 -20 Mark 52-1-6 Q2QL2W / 62A-6 40 102 102 109 -20 -20 -20 Mark 52-1-7 Q2QL2W / 62A-7 40 79 83 87 -20 -20 -20 Mark 52-1-8 Q2Q65W / 62A-8 40 105 120 128 -20 -20 -20 Mark 52-1-9 Q2Q65W / 62A-9 40 105 103 124 -20 -20 -20 Mark 52-1-10 Q2QL2W / 62A-10 40 94 90 66 -20 -20 -20 N3 Steam Outlet Nozzle Mabrk 56-1-1 Q2Q68W / 182B-1 40 78 99 52 -20 -20 -20 Mark 56-1-2 Q2Q68W / 182B-2 40 76 87 75 -20 -20 -20 Mark 56-1-3 Q2Q68W / 182B-3 40 69 107 97 -20 -20 -20 Mark 56-1-4 Q2Q68W / 182B-4 40 121 120 116 -20 -20 -20 N4 Feedwater Nozzle NMrk 59-1-1 Q2QL2W / 315A-1 40 103 103 87 -20 -20 -20 Mark 59-1-2 Q2QL2W / 315A-2 30 77 96 100 -30 -20 -20 NMrk 59-1-3 Q2QL2W / 315A-3 30 77 56 54 -30 -20 -20 Mark 59-1-4 Q2QL2W / 315A-4 30 114 101 92 -30 -20 -20 Mbrk 59-1-5 Q2QL2W / 315A-5 40 95 72 116 -20 -20 -20 Mark 59-1-6 Q2QL2W / 315A-6 40 101 79 73 -20 -20 -20 N5 Core Spray Nozzle (Low Pressure) Mark 63-1-1 Q2QL3W / 867B-1 20 82 81 66 -40 -20 -20 N6 RHR-LPCI Nozzle Mark 67-1-1 Q2QL3W / 854A-1 40 53 101 84 -20 -20 -20 NMrk 67-1-2 Q2QL3W / 854A-2 40 64 74 70 -20 -20 -20 M/brk 67-1-3 Q2QL3W / 854A-3 40 65 70 60 -20 -20 -20 N7 Top Head Spray Nozzle Mabrk71-1-1 Q2Q60QT / 569F-1 30 98 114 131 -30 -20 -20 Mark 72-1 Weld Neck Flange 720230 / 10618D / 3445-1 20 90 122 179 -40 -40 -40 Mark 73-2-1 Blind Flange C4889-3 30 54 70 78 -30 -30 -30 N8 Top Head Vent Nozzle Mark 74-1 211290 / L3156 40 50 54 51 -20 -20 -20 Mark 75-1 Weld Neck Flange 720230 / 10618D / 3445-1 20 90 122 179 -40 -40 -40 Mark 76-1A Blind Flange D1295-3 20 52 58 72 -40 -40 -40 N9 Jet Pump Instrumentation Nozzle Mark 77-1-1 Q2QL4W / R64A-1 40 66 62 54 -20 -20 -20 Mark 77-1-2 Q2QL4W / R64A-2 40 57 62 56 -20 -20 -20 N10 CRD Hydraulic System Return Mark 80-1 Q2QL4W / 868B 40 67 63 65 -20 -20 -20 Page 18 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Initial RTNDT Values for RPV Materials, Continued Nozzle Materials Drop Test Charpy Energyil] (Tsor-60) Weight RTNDT Component Heat or Heat / Flux / Lot Temp (ft-lb) (*F) NDT (*F) _ _ _ _ (°F) N11 Core AP & Liquid Control Mark 84-1 NXS001 / 1765 (Alloy 600) (Note 2) Mark 89-1 Weld Coupling 67993 / 1550 (Stainless) (Note 2) Mark 85-1-1 54318 / 1095 (Alloy 600) (Note 2) N12 Instrument Nozzle Mark 88-1-1 thru 88-1-4 717456 / L9295 / FE-1 40 64 58 57 -20 -20 -20 N13 Instrument Nozzle Mark 88-1-5 thru 88-1-6 717456 / L9295 / FE-1 40 64 58 57 -20 -20 -20 N14 Instrument Nozzle Mark 91-1 -1 thru 91-1-4 717456 / L9295 / FE-1 40 64 58 57 -20 -20 -20 N15 Drain Nozzle Mark 93-1 B19W (213099) / R853A 40 85 86 93 -20 -20 -20 N16 Core Spray (High Pressure) Mark 63-1-2 Q2QL3W / 867B-2 40 89 89 71 -20 -20 -20 N17 Seal Leak Detector Nozzle Mark 104-1 NX4745 / 481 (Alloy 600) (Note 2) Mark 30-1 615146 / 3444 (Note 3) Mark 142-1 615146 / 3444 (Note 3) NI 8 Top Head Spare Nozzle Mark 71-1-2 Q2Q60QT / 569F-2 30 51 66 93 -30 -20 -20 Mark 99-1 Weld Neck Flange 720230 / 10618D / 3445-1 20 90 122 179 -40 -40 -40 Mark 73-2-2 Blind Flange B7492-2 40 62 50 55 -40 -40 -40 CRO Stub Tube Mark 14 NX4994 /M1 753 (Alloy 600) (Note 2) Mark 14 NX5190 /M1769 (Alloy 600) (Note 2) Mark 14 NX6104 /M4728 (Alloy 600) (Note 2) Mark 14 NX7249G (Alloy 600) (Note 2) Mark 14 1NX5151 / M1 758 (Alloy 600) (Note 2) 4 Mar( 1 2NX5291/M1789 (Alloy 600) (Note 2)
!varK 14 4NX5190 / M 770 (Alloy 600) (Note 2)
Mark 14 NX5159 (Alloy 600) (Note 2) Mark 14 NX4902 (Alloy 600) (Note 2) Nozzle Welds N6 LPCI Nozzle 629865/A421A27AD -10 69 70 88 -70 -90 -70 N6 LPCI Nozzle 04T931/A423B27AG 0 65 69 72 -60 -90 -60 N6 LPCI Nozzle 402P3162/H426B27AE -10 60 54 68 -70 -70 -70 N6 LPCI Nozzle 5P6771/Linde 124/0342 (Single) 30 78 53 68 -30 -30 -30 N6 LPCI Nozzle 5P6771/Linde 124/0342 (Tandem) 40 77 81 83 -20 -20 -20 N6 LPCI Nozzle 492L4871/A421B27AF 10 56 58 61 -50 -80 -50 N6 LPCI Nozzle 05T776/L314A27AH -10 69 72 81 -70 -70 -70 (1) Minimum Charpy values are used in these calculations. (2) Alloy 600 and Stainless Steel components do not require fracture toughness evaluation; see Appendix A of Reference 6.1 for additional information. (3) Small diameter piping does not require fracture toughness evaluation; see Appendix A of Reference 6.1 for additional information. Page 19 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Initial RTNDT Values for RPV Materials, Continued Appurtenance Materials Drop Test t Component Heat or Heat/ Flux/ Lot Temnp Charpy Energyp (Tsor-60) Weight RTNDT ( °F ) (ft-lb ) ( °F
) NDT ( °F)
(°F) Support Skirt M ark 10-1-1 C2762-1F
. 30 60 62 59 -30 -50 -30 Mark 10-1-2 A0797-3 20 54 62 51 -40 -10 -10 Support Skirt Extension _____ _ _0 _i___ 10.
Mark 10-2-1 thru 10-2-6 R0530-1 70 112 110 90 10 10 10 Shroud Support Marks 20-1-1 thru 20-1-4 (Horizontal Support to BH) 3NX6945-1 / PC-9435 (Alloy 600) _(Note 2) Marks 20-1-1 thru 20-1-4 (Horizontal Suppo rt to BH) 2-NX7083-1 / PC -9729 (Alloy600) . . . . .. (Note 2) Marks 20-2-1,20-2-2 (Horizontal Support to BOH) 2NX6945-1 / PC-9437 (Alloy 600) _ (Note 2) Marks 20-4-1 thru 20-4-6 (Vertical __ Supports from COP) 1NX6897-1 PC-9401
/ (Alloy 600) _ (Note 2)
Marks 20-4-7 thru 20-4-14 (Vertical Sup ports from CP) 1NX6897-7 I PC-9402 (Alloy 600)_ (Note 2) Marks 17-1-1 thru 17-1-14 (Vertical Suppors to BH) 636478-4X (Alloy 600) (Note 2) Stabilizer Bracket Mark 101-1 A1322-2B 50 67 50 51 -10 -20 -10 Refueling Bellows Bar ..... Mark 29-1-1 thru 29-1-6 A2457-7 60 50 50 52 0 -20 0 Mark 45-1-1 thru 45-1-6 R0503-1 60 53 46 60 0 0 0 Guide Rod Bracket Mark 106-1-1 & 106-1-2 120867 / 333 (Stainless) (Note 2) Steam Dryer Suppo)rt Bracket Mark 108-1-1 thru 108-1-4 120867 / 333 (Stainless) (Note 2) Core Spray Bracket Mark 116-1-1 thru 116-1-8 120867 / 333 (Stainless) _ (Note 2) r 'eedw ater Sparger B racket _ Mark 112-1,1 thru 112-1-12 633345 / 333 (Stainless) (Note 2) Top Head Lifting Lug E E :: _ Mark 43-1-5 thru 43-1-8 C3916-2B 60 58 53 59 0 -40 0 Thermocouple Pads and Clamps Mark 47-1 and 47-2 C3427-13A 30 50 53 50 -30 -30 -30 Dryer Hold Down Bracket _ _ Marks 1,10-1-1 thru 110-1-4 C2325-2 60 55 55 52 0 -20 0 Marks 110-1-1 thru 110-1-4 C2480-2 20 50 53 50 -40 -40 1 40 (1) Minimum Charpy values are used in these calculations. (2) Alloy 600 and Stainless Steel components do not require fracture toughness evaluation; see Appendix A of reference 6.1 for additional information. CP = Core Plate; BH = Bottom Head Page 20 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Initial RTNDT Values for RPV Materials, Continued Bolting Materials Test Charpy Energy Min Lat LST Component Heat Temp (ft-lb) Exp (OF) (OF) (mills) STUDS Closure Mark 38-1-1 thu 38-1-13 11949 10 54 56 56 32 10 Mark 38-1-14 thu 38-1-26 11054 10 46 47 46 26 10 Mark 38-1-27 thu 38-1-44 11949 10 54 56 56 32 10 Mark 38-1-45 thu 38-1-56 84176 10 49 49 48 26 10 Mark 38-1-57 thu 38-1-76 11949 10 54 56 56 32 10 N7 & N18 Nozzles Mark 73-4 82116 10 63 64 63 25 10 N8 Nozzle Mark 76-3 82116 10 63 64 63 25 10 NUTS Closure Mark 39-5-1 thu 39-5-48 43320 10 48 50 48 29 10 Mark 39-5-49 thu 39-5-54 83706 10 50 51 54 26 10 Mark 39-5-55 thu 39-5-76 84751 10 48 48 47 25 10 N7 & N18 Nozzles Mark 73-5 14886 10 69 72 76 30 10 N8 Nozzle Mark 76-4 14886 10 69 72 76 30 10 CLOSURE BUSHINGS Mark 152-1-1 thru 152-1-76 52504 10 48 48 50 28 10 CLOSURE WASHERS Mark 39-6-1 thru 39-6-76 83706 10 50 51 54 26 10 Page 21 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 Adjusted Reference Temperatures [32 EFPY (Peak ID Fluence: 9.60E+17 n/cm 2 )] Lower Shell, Shell I to Shell 2 Girth Weld and Lower Shell Axial Welds 2 Thickness in inches = 6.1875 54 EFPY Peak I.D. fiuence = 1.58E+18 n/cm 2 32 EFPY Peak I.D. fluence = 9.36E+17 n/cm 2 32 EFPY Peak 1/4 T fiuence = 6.46E+17 n/cm Lower-intermediate Shell and Axial Welds 2 Thickness in inches= 6.1875 54 EFPY Peak I.D. fluence = 1.62E+18 n/cm 32 EFPY Peak I.D. fluence = 9.60E+17 2 32 EFPY Peak 1/4 T fluence = 6.62E+17 n/cm N6 Nozzle 2 Thickness in inches= 6.1875 54 EFPY Peak I.D. fluence = 5.34E+17 n/cm 32 EFPY Peak I.D. fluence = 3.16E+17 2 32 EFPY Peak 1/4 T fluence = 2.18E+17 n/cm N12 Nozzle 2 Thickness in inches= 6.1875 54 EFPY Peak I.D. fluence = 3.65E+17 n/cm 32 EFPY Peak I.D. fluence = 2.16E+17 2 32 EFPY Peak 1/4 T fluence = 1.49E+17 n/cm Adjusted hintial 1/4 T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT %Cu %N1 CF CF RTNmT Fluence A RTNDT Oa o, Margin Shift ART 2
°F n/cm °F (Note 8) °F °F °F PLATES:
Lower-Intermedlate Shell C3065-1 0.06 0.63 37 -10 6.62E+17 12.6 0 6 12.6 25.1 15 C3121-2 0.09 0.65 58 0 6.62E+17 19.7 0 10 19.7 39.4 39 C3147-1 0.11 0.63 74.5 0 6.62E+17 25.3 0 13 25.3 50.6 51 Lower Shell C3065-2 0.06 0.63 37 10 6.46E+17 12.4 0 6 12.4 24.8 35 C3066-2 0.07 0.64 44 -20 6.46E+17 14.7 0 7 14.7 29.5 9 C3147-2 0.11 0.63 74.5 0 6.46E+17 25.0 0 12 25.0 49.9 50 WELDS: Circumferential 4P7216(S)/0751 0.045 0.80 61 -50 6.46E+17 20.4 0 10 20.4 40.9 -9 4P7216(T)/0751 0.035 0.82 47.5 -80 6.46E+17 15.9 0 8 15.9 31.8 -48 4P7465(S)/0751 0.02 0.82 27 -60 6.46E+17 9.0 0 5 9.0 18.1 -41 4P7465(T)/0751 0.02 0.80 27 -60 6.46E+17 9.0 0 5 9.0 18.1 -41 Axial Lower Shell 5P6214B(S)/0331 0.02 0.82 27 -50 6.46E+17 9.0 0 5 9.0 18.1 -31 Lower Shell 5P6214B(T)/0331 0.014 0.70 22.8 -40 6.46E+17 7.6 0 4 7.6 15.3 -25 Lower-ihtermediate Shell 5P5657(S)/0931 0.07 0.71 95 -60 6.62E+17 32.3 0 16 32.3 64.5 5 Lower-hItermediate Shell 5P5657(T)/0931 0.04 0.89 54 -60 6.62E+17 18.3 0 9 18.3 36.7 -23 NOflLES: Forgings 12 N6 LPCI ' C2QL3W 0.07 0.86 44 -20 2.18E+17 8.0 0 4 8.0 16.0 -4 1 23 N12 Water Level hstrumentation 717456 0.272 0.214 136 0 1.49E+17 19.4 0 10 19.4 38.9 39 1 Weldsn ) 12 N6 LPC1 ) 629865/A421A27AD 0.05 1.10 68 - -70 2.18E+17 12.3 0 6 12.3 24.7 -45 1 ti LPC (21 04T931/A423B27AG 0.03 1.00 41 -60 2.18E+17 7.4 0 4 7.4 14.9 -45 22 ti LPCI ' 402P3162/H-426B27AE 0.03 0.83 41 -70 2.18E+17 7.4 0 4 7.4 14.9 -55 1 NWLPCI (2) 5P6771(S)/0342 0.03 0.88 41 -30 2.18E+17 7.4 0 4 7.4 14.9 -15 2 W6LPCI ) 5P6771(T)/0342 0.04 0.95 54 -20 2.18E+17 9.8 0 5 9.8 19.6 0 1 ti LPCI (2) 492L4871/A421B27AF 0.03 0.98 41 -50 2.18E+17 7.4 0 4 7.4 14.9 -35 1 N6 LPC 12) 05T776/L314A27AH 0.06 0.92 82 -70 2.18E+17 14,9 0 7 14.9 29.8 -40 N12 Water Level instrumentation w Inconel INTEGRATED SURVEILLANCE 4 PROGRANM " Plate "' C2761-2 0.10 0.54 65 10 6.62E+17 22 0 11 22.1 44.1 54 Weld (1,71 5P6214B 0.027 0.94 36.8 53.7 -40 6.62E+17 18 0 9 18.2 36.5 -4 Notes: (1) This evaluation includes both Single (S) and Tandem (T) wire materials. (2) Chemistries and InitialRTNDTfor the nozzle forgings were obtained from BWR fleet CMTR data. Copper content was based upon 14 data points and nickel was based upon 35 data points. (3) The N12 nozzle is classified as a partial penetration in Shell Ring #2. (4) Representative materials as defined by the Integrated Surveillance Program (ISP). (5) The ISP plate material is not the same heat number as the target plate as defined in the ISP. Therefore, the CF from RG1.99, Position 1.1 is used to determine the ART. This information is riot required per BWRVIP-135, to dictate the ART used for the NMP2 PT curves, and is provided for information only. Plate C3147-1 remains the limiting beltline material for the purpose of developing the PT curves. (6) The ISP weld material is not the same heat number as the target weld (5P5657) as defined in the ISP. However, the ISP weld material is the same heat as another beltline weld. Therefore, the surveillance data is considered and the CF is adjusted as defined in RG1.99 Position 2.1. (7) The Adjusted CF was conservatively calculated using the limiting parameters for this heat because there are two (2) sets of data provided by the ISP (from Perry and SSP materials that originated from Grand Gulf). CFAdjusted = (27/20) 39.75 = 53.7°F. (8) Previous submittals included a conservative , = 14.5°F. BWROG NRC-approved methodology has been applied, allowing reduction such that o, = 0°F. (9) Non-ferritic materials do not require evaluation for fracture toughness. Page 22 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B NMP2 RPV Beltline P-T Curve Input Values Adjusted RTNDT = Initial RTNDT + Shift A = 0 + 51 = 51°F (Based on ART Table) Vessel Height H = 869.75 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to base metal) R = 126.7 inches Minimum Vessel Thickness (without clad) t = 6.1875 inches Page 23 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix B 1 NMP2 Definition of RPV Beltline Regionli Elevation Component (inches from RPV "0") Shell # 2 - Top of Active Fuel (TAF) 366.3" Shell # 1 - Bottom of Active Fuel (BAF) 216.3" Centerline of Recirculation Outlet Nozzle in Shell # 1 172.5" Top of Recirculation Outlet Nozzle Ni in Shell # 1 197.9" Centerline of Recirculation Inlet Nozzle N2 in Shell # 1 181.0" Top of Recirculation Inlet Nozzle N2 in Shell # 1 198.7" Centerline of LPCI Nozzle N6 in Shell #2 372.5" Bottom of LPCI Nozzle N6 in Shell #2 337.1" Centerline of Water Level Instrumentation Nozzle N12 in Shell # 2 366.0" Bottom of Water Level Instrumentation Nozzle N12 in Shell #2 364.4" [1] The beltline region is defined as any location where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm 2. [2] The dimensions identified above are specified as the distance (elevation) above vessel "0" The review of the axial fluence profile indicated that the RPV fluence projected to 54 EFPY drops to less than 1.0e17 n/cm 2 at - 12" below the BAF and at - 12" above TAF. The beltline region considered in the development of the P-T curves has been conservatively adjusted to include the region from 204.6" to 378.4" above reactor vessel "0". Based on the above, it is concluded that none of the NMP2 reactor vessel plates, nozzles, or welds, other than those included in the Adjusted Reference Temperature Table, are in the beltline region.
.Page 24 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Appendix C NMP2 Reactor Pressure Vessel P-T Curve Checklist Parameter Completed Comments/Resolutions/Clarifications Initial RTNDT Initial RTNDT has been determined The beltline LPCI N6 and Water Level for NMP2 for all vessel materials Instrumentation N12 nozzle forgings including plates, flanges, forgings, and welds have been considered in the studs, nuts, bolts, welds. development of the beltline PT curves. Include explanation (including methods/sources) of any exceptions, resolution of discrepant data (e.g., deviation from originally reported values). Appendix B contains tables of all Initial RTNDT values for NMP2 Has any non-NMP2 initial RTNDT The review performed indicated that the information (e.g., ISP, comparison NMP2 plant-specific weld material ART to other plant) been used? values bound those determined for the ISP representative weld material. If deviation from the LTR process occurred, sufficient supporting No deviations from the LTR process information has been included (e.g., Charpy V-Notch data used to determine an Initial RTNDT). All previously published Initial RTNDT RVID was reviewed. All initial RTNDT values from sources such as the values agree. No further review was GL88-01, RVID, FSAR, etc., have required. been reviewed. Adjusted Reference Temperature (ART) Sigma I (standard deviation for NMP2 has previously calculated the Initial RTNDT) is 0°F unless the ART using a conservative a( of 14.5 0 F RTNDT was obtained from a source for all materials. However, since the other than CMTRs. If ai is not equal GE/BWROG method of estimating to 0, reference/basis has been RTNDT per NEDC-32399-P operates on provided, the lowest Charpy energy value and provides a conservative adjustment to the 50 ft-lb level, the value of ca is taken to be 0°F for the vessel materials in the evaluation. Sigma A (standard deviation for ARTNDT) is determined per RG 1.99, Rev. 2 Page 25 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report Parameter Completed Comments/Resolutions/Clarifications Chemistry has been determined for The vessel chemistries are consistent all vessel beltline materials including with previously reported information. plates, forgings (if applicable), and welds for NMP2. The LPCI (N6) Nozzle and the Water Include explanation (including Level Instrumentation (N12) Nozzle are methods/sources) of any now included in the extended beltline exceptions, resolution of discrepant region. For the N6 nozzle forging and data (e.g., deviation from originally welds, CMTR's were located which reported values), contained Cu and Ni values as shown in the ART table. Non-NMP2 chemistry information An adjusted CF was required for ISP (e.g., ISP, comparison to other Heat 5P6214B, and was determined plant) used has been adequately per Paragraph 2.1 of RG 1.99. defined and described. For any deviation from the LTR No deviations from the LTR process process, sufficient information has been included. All previously published chemistry values from sources such as the GL88-01, RVID, FSAR, etc., have been reviewed. The fluence used for determination of ART and any extended beltline region was obtained using an NRC-approved methodology. The fluence calculation provides an axial distribution to allow determination of the vessel elevations that experience fluence of 1.OE17 n/cm 2 both above and below active fuel. The fluence calculation provides an axial distribution to allow determination of the fluence for intermediate locations such as the beltline girth weld (if applicable) or for any nozzles within the beltline region. All materials within the elevation range where the vessel experiences a fluence -1.0EI 7 n/cm 2 have been included in the ART calculation. All initial RTNDT and chemistry information is available or Page 26 of 27 PTLR-2 Revision 0
NMP2 Pressure and Temperature Limits Report SParameter Completed Comments/Resolutions/Clarifications explained. Discontinuities The discontinuity comparison has been performed as described in No deviations Section 4.3.2.1 of the LTR. Any deviations have been explained. Discontinuities requiring additional components (such as nozzles) to be considered part of the beltline have been adequately described. It is clear which curve is used to bound each discontinuity. Appendix G of the LTR describes the process for considering a The thickness discontinuity evaluation thickness discontinuity, both beltline demonstrated that no additional and non-beltline. If there is a adjustment is required; the curves discontinuity in the NMP2 vessel bound the discontinuity stresses. that requires such an evaluation, the evaluation was performed. The affected curve was adjusted to bound the discontinuity, if required. Appendix H of the LTR defines the An evaluation was performed to assure basis for the CRD Penetration curve that the CRD discontinuity bounds the discontinuity and the appropriate other discontinuities that are protected transient application. The NMP2 by the CRD curve with respect to evaluation bounds the requirements pressure stresses. For heatup/ of Appendix H. cooldown conditions, the CRD penetration provides bounding limits. Appendix J of the LTR defines the basis for the Water Level Instrumentation Nozzle curve discontinuity and the appropriate transient application. The NMP2 evaluation bounds the requirements of Appendix J. Page 27 of 27 PTLR-2 Revision 0
ATTACHMENT 4 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION - NON-PROPRIETARY VERSION Nine Mile Point Nuclear Station, LLC November 21, 2012
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version The following Requests.for Additional Information (RAI) were received by Grand Gulf Nuclear Station (GGNS) in email correspondence dated January 31, 2011 (Accession Number ML110310390). The RAls were in response to their submittal of a license amendment request (LAR) for an Extended Power Uprate (EPU). The original RAI included questions for vessel internals and the Pressure-Temperature Limits Report (PTLR) implementing the methodology of NEDC-33178P-A, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperate Curves." This attachment only includes the GGNS RAI questions that apply to the PTLR. The NMP2 response was prepared by General Electric Hitachi (GEH) to address the applicability of the GGNS RAIs to Nine Mile Point Unit 2 (NMP2). The GGNS RAI question is in italics, followed by the response as it applies to NMP2. NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of attachment 6 which has the proprietary information removed. Portions of the document, that have been removed are indicated by an open and closed double square bracket as shown here (( )). Page 1 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAI-3 Confirm that the proposed PTLR will take effect prior to or concurrent with the proposed EPU, replacingthe P-T limits currently in the GGNS Technical Specifications (TS). If the previous statement is correct the staff will not review the P-T limits in the GGNS TS, as only the PTLR is applicable to the EPU. NMP2 Response to GGNS RAI-3 This RAI is not applicable to NMP2. EPU has already been implemented at NMP2. The P-T limit curves currently contained in the NMP2 Technical Specifications are valid for the peak vessel fluence corresponding to 22 Effective Full Power Years (EFPY), based on a 3467 MWth rating. Conservative projections for the 1st EPU cycle indicate that NMP2 will reach the peak vessel fluence corresponding to the current PT curves in December 2013. Page 2 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAJ-4 Do the P-T limit curvesprovided include a hydrostaticpressure adjustmentfor the column of water in a full RPV? If so, provide the pressure head used in the P-T limit curve analysis. NMP2 Response to GGNS RAI-4 Yes, the PT limit curves include a hydrostatic pressure adjustment for the column of water in a full RPV. The pressure head for the beltline hydrostatic test curve (Curve A) for NMP2 is 23.6 psig. This is determined using the height of the vessel and the elevation of bottom of active fuel. The full vessel pressure head is 31.4 psig. This pressure is indirectly used in the PT curve analysis. It is considered in the determination of KI for the bottom head curve as discussed in the PT curve licensing topical report, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," NEDC-33178P-A, Revision 1. Sections 4.3.2.1.1 and 4.3.2.2.2 of NEDC-33178P-A, Revision 1, include additional discussion regarding the use of the pressure head. Page 3 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAI-5 Address inconsistencies between the statement that "the P-T curves are beltline (A.1224-1 plate) limited above 1330 psig for Curve A for 35 EFPY... " and the staff determination that the P-T curves are beltline (A1224-1 plate) limited above -1360psigfrom data in Table I of GNRO-2010/00056. NMP2 Response to GGNS RAI-5 This RAI is not applicable to NMP2 as it addresses a plant-specific GGNS statement. Page 4 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAI-6 Provide the surveillance data and the analysis of the surveillance data used to determine ART from reference 6.3 (BWRVIP-135, Revision 1 "BWR Vessel and Internals ProjectIntegrated Surveillance Program(ISP)Data Source Book and PlantEvaluations'),as required by NEDC-33178P-A. NMP2 Response to GGNS RAI-6 BWRVIP- 135, Revision 2 provides the surveillance data considered in determining the chemistry and any adjusted Chemistry Factors (CF) for the beltline materials. Excerpt from BWRVIP-135, Revision 2 (( For NMP2, the Integrated Surveillance Program (ISP) representative weld, heat (( )), is not the target vessel material. However, the ISP weld heat is the same heat as another beltline weld. Therefore, the surveillance data is considered and the CF is adjusted as defined in Regulatory Guide 1.99 (RGI.99), Position 2.1. Plate Material The ISP representative plate material, heat (( )), is not the target vessel material and is therefore provided for information only. The NMP2 plant-specific plate heat C3147-1 is the limiting material for the vessel, and is used in development of the PT curves. Note that the Adjusted Reference Temperature (ART) table provided in the PTLR includes chemistries based on both the plant-specific information and using the BWRVIP-135 best estimates. For the plate material, heat (( )) was contained in one (1) of the (( capsules that has been tested and analyzed. The resultant chemistry is E[ )) Cu and (( )) Ni. The CF from Regulatory Guide 1.99, Revision 2 (RG1.99) is (( fl; a fitted CF will be determined once a second capsule is tested. Therefore, the CF used for the ISP evaluation for the plate material is [E I]. Page 5 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version Excerpt from BWRVIP-135, Revision 2 ((I Weld Material For the weld material, heat (( was included in (( capsules and (( )) capsule that have been tested and analyzed. The resultant chemistry is provided as )) Cu and (( )) Ni for the (( )) capsule material, and as (( )) Cu and (( )) Ni for the (( )) capsule materials; the CF for this chemistry is (( )). The mean surveillance chemistry is defined as (( )) Cu and (( )) Ni. The resulting RG1.99 CF for the mean chemistry is (( )), and the fitted CF is (( )). The maximum scatter in the fitted data is within the 1-sigma value of 28°F from RG1.99. BWRVIP-135 also provides best estimate chemistries that are used in the ART evaluation. Best estimate weld heat (( )) information is provided, defining the chemistry as [E )) Cu and (( )) Ni. The CF from RG1.99 for the best estimate chemistry of (( )) Cu and (( )) Ni is (( )). The chemistry for this weld heat that occurs in the NMP2 vessel beltline is (( )) Cu and (( )) Ni; the resulting CF is ((i )). For the weld material, the CF is determined using the equation: Adjusted Surv CF- (Table CF,,.,. *CFFri+/-*D-Table C. Therefore, the adjusted CF (( )), which is used for the ISP weld material evaluation. It is noted that the (( )) weld material from the ISP has a CF of (( )) and the EE )) weld material from the ISP has a CF of l)). For the equation above, use of the (( )) CF has been conservatively considered. Page 6 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version Excerpts from BWRVIP-135, Revision 2 [I Page 7 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version I'l
))
Page 8 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAJ- 7 Provideadditionaldetailfor the non-beltline analysis conducted in the following areas in orderfor the staff to complete independent verification of the proposedP-T limits:
- 1. Identify limiting materialsfor the Reference Temperature for Nil Ductility Transition (RTNDT) values used to shift the generic Bottom Head and Upper Vessel P-T curves when applying NEDC-33178P-A.
- 2. The staff identified a limiting RTNDT of 1OF for the Bottom Head Torus Plates, while GGNS assumed a RTNDT of 24.60F for Bottom Head Curve B. Support all RTNDT values reported by providing details of any plant-specific analysis conducted.
- 3. Explain minor differences in assumedRTNDT valuesfor the Bottom Head.Specifically Curves A and C assume a limiting RTNDT of 19 'F, while Curve B assumes a limiting RTNDT of 24. 6 °F.
- 4. Which region of the RPVis limitingfor Curve C < 312 psig?
NMP2 Response to GGNS RAI-7 In order to determine how much to shift the Pressure-Temperature (PT) curves, an evaluation is performed -using Tables 4-4b and 4-5b from NEDC-33178P-A. These tables define the required Temperature minus Reference Temperature of Nil Ductility Transition (T-RTNDT) that is used to develop the non-shifted curves. Each component listed in these tables is evaluated using the plant-specific initial RTNDT for each component. The required temperature is then determined by adding the T-RTNDT to the plant-specific RTNDT, thereby resulting in the required T for the curve. As the upper vessel curve is initially based on the non-shifted feedwater (FW) nozzle T-RTNDT, all resulting T values are compared to the FW nozzle T. The difference between the maximum T and the FW nozzle T-RTNDT is used to shift the upper vessel curve. The same method is applied for the Control Rod Drive (CRD) curve. In this manner, it is assured that each curve bounds the maximum discontinuity that is represented. For the NMP2 upper vessel curve, the maximum T value from the method described above is [ )). The initial required T-RTNDT for the (( )); this is then adjusted by the NMP2-specific maximum (( )), resulting in (( )). Comparing this to the FW nozzle values, the required T-RTNDT is (( )), which is added to the E[
]. It is seen that the resulting T required for the FW nozzle is (( )). As (( )) is E[
)) the baseline non-shifted FW nozzle curve [f )), which is based on the NMP2 upper vessel curve is based on an RTNDT of (( )). As noted above, this calculation was performed for each component shown in Table 4-4a; only the limiting case is presented here. For the NMP2 bottom head or CRD [] respectively), the maximum T values from the method described above is (( )). The required T-RTNDT for the )); this is adjusted by the NMP2-specific fl maximum (( )), resulting in (( )). Comparing this to the CRD values, the required T-RTNDT is 161'F, which is added to the (( )). It is seen that the resulting T required for the bottom head is (( )). As (( )) is Er )) the baseline non-shifted CRD curve (( )), the NMP2 bottom Page 9 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version head (CRD) curve is based on an (( )). As noted above, this calculation was performed for each component shown in Table 4-5a; only the limiting case is presented here. Appendix H ofNEDC-33178P-A contains the details of an analysis performed to determine the baseline requirement (non-shifted) for the (( It can be seen in Section H.5 of Appendix H that the stresses developed in this finite element analysis demonstrated that the (( )), resulting in a baseline non-shifted required T-RTNDT of (( )). Therefore, considering the determination of the required shift from the paragraph above for (( calculations for all components listed in Table 4-5a were compared to the CRD T, which is (( )) (where (( )) materials). Therefore, the shift for the bottom head ((. For Curve C, the upper vessel is bounding at pressures between 110 and 312.5 psig. For pressures above 1250 psig, the beltline is bounding. For the remaining pressure ranges from 0 to 110 psig and between 312.5 and 1250 psig, the I0CFR50 Appendix G requirement are bounding. Page 10 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAI-8 identifies nozzle N12 as a beltline water level instrument nozzle and notes that an evaluation was conducted using the limiting materialpropertiesfor the adjoining shell ring, which appears to be appropriateas nozzle N12 is identified as austenitic.Provide details of this evaluation which demonstratesthat the drill hole for the beltline water level instrument nozzle is not limiting. NMP2 Response to GGNS RAI-8 Appendix J of NEDC-33178P-A provides detailed results of an analysis performed for the water level instrumentation nozzle that provides the required stresses for the drill hole in the shell plate. These stresses were used to generate a specific curve applicable for the water level instrumentation nozzle to assure that this location is bounded in the PT curves. For NMP2, the water level instrumentation nozzle is (( The LPCI (N6) nozzle is included in the extended beltline region in accordance with 10 CFR 50 Appendix G. For the N6 nozzle, CMTRs were located that contained Cu (0.07%) and Ni (0.86%) values. An assessment of the N6 nozzle is contained in Appendix B of the PTLR. The Water Level Instrumentation (N12) nozzle is included in the extended beltline region in accordance with 10 CFR 50 Appendix G. For this nozzle, the CMTRs do not contain the Cu and Ni content. Since plant-specific information regarding the Cu and Ni content for this material is not available, the evaluation was performed based on a (( )) fabricated from SA508 Class 1 material. This was defined based on a search of (( )) for SA508 Class 1 materials. The Cu and Ni values (( respectively. The weld connecting the forging to the vessel shell is fabricated from Inconel, and does not require a fracture toughness evaluation. Page 11 of 12
Attachment 4 Responses to Requests for Additional Information - Non-Proprietary Version GGNS RAI-9 Provide detailson any plant-specificfeedwaternozzle evaluation conducted in support of the proposed P-T limits or explain why plant-specific evaluation was not required. NMP2 Response to GGNS RAI-9 An evaluation was performed for the feedwater nozzle as described in Section 4.3.2.1.3 of NEDC-33178P-A. This evaluation confirmed that the feedwater discontinuity bounds the other discontinuities defined in Table 4-4a of NEDC-33178P-A. The first part of the evaluation is as described in the response to RAI 7, where it is assured that the limiting component that is represented by the upper vessel nozzle curve is bounded. A second evaluation was performed using the NMP2-specific feedwater nozzle dimensions; this evaluation is shown below to demonstrate that the ((
)) curve is applicable to NMP2.
Vessel radius to base metal, Rv R Vessel thickness, tv Vessel pressure, Pv Pressure stress = PR/t = (( Dead Weight + Thermal Restricted Free End stress Total Stress = (( )) 1 The factor F (a/rn) from Figure A5-1 of "PVRC Recommendations on Toughness Requirements for Ferritic Materials", Welding Research Council Bulletin 175, August 1972 (WRC-175) is determined where: a = / (tn2 + tV2) Y2 tn = thickness of nozzle tv = thickness of vessel rn = apparent radius of nozzle = n + 0.29*rc n = actual inner radius of nozzle rc = nozzle radius (nozzle corner radius) ] Therefore, a/rn = )). The value F (a/rn), taken from Figure A5-1 of WRC-175 for an1/2 a/rn
)). Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 CF(r1a) of ((
F(a/r, 1 ): Nominal Ki = 1.5
- A detailed upper vessel example calculation for core not critical conditions is provided in Section 4.3.2.1.4 ofNEDC-33178P-A. Section 4.3.2.1.3 ofNEDC-33178P-A, presents the (( ))
FW nozzle evaluation upon which the baseline non-shifted upper vessel PT curve is based. It can be seen that the nominal K, from this evaluation is (( )). Therefore, it has been shown that the nominal K, for the NMP2-specific FW nozzle is bounded by the E[ )) K, demonstrating applicability of the FW nozzle curve for NMP2. Page 12 of 12
ATTACHMENT 5 AFFIDAVITS JUSTIFYING WITHHOLDING OF PROPRIETARY INFORMATION Nine Mile Point Nuclear Station, LLC November 21, 2012
I ELECTRIC POWER RESEARCH INSTITUTE Christine King, EPRI Director, Nuclear Fuels & Chemistry November 5, 2012 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Request for Withholding of the Proprietary Information included in the following Document: Attachment A- Pressure Temperature Curve Request (RAI Responses)- Proprietary Version Attachment B- Pressure Temperature Curve Request (RAI Responses)- Non-Proprietary Version To Whom It May Concern: This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure the information identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified above (the "Report"). Proprietary and non-proprietary versions of the Report and the Affidavit in support of this request are enclosed. EPRI desires to disclose the Report in confidence to assist the NRC review of the enclosed submittal to the NRC by Constellation Energy.The Proprietary information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Propreitary information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed. Ifyou have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (650)-855-2164. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440. Sincerely, ,_. Attachment(s) c: Sheldon Stuchell, NRC (sheldon.stuchell@nrc.gov) Together . . . Shaping the Future of Electricity 1300 West W.T. Harris Boulevard, Charlotte, NC 28262-8550 USA
- 704.595.2732
- Mobile 704.490.2653 . nwilmshurst@epri.com
- 1 ELECTRIC POWER RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Proprietary Information included in the following Document:
Attachment A- Pressure Temperature Curve Request (RAI Responses)- Proprietary Version Attachment B- Pressure Temperature Curve Request (RAI Responses)- Non-Proprietary Version I, Christine King, being duly sworn, depose and state as follows: I am the Director of Nuclear Fuels and Chemistry at Electric Power Research Institute, Inc. whose principal office is located at 3420 Hillview Avenue, Palo Alto, California ("EPRI") and I have been specifically delegated responsibility for the above-listed Report that is sought under this Affidavit to be withheld (the "Report"). I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Report on behalf of EPRI. EPRI requests that the Report be withheld from the public on the following bases: Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information:
- a. The Report is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Report do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Report. The Report is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof.
- b. EPRI considers the Report and the proprietary information contained therein (the "ProprietarV Information") to constitute trade secrets of EPRI. As such, EPRI holds the Report in confidence and disclosure thereof is strictly limited to individuals and entities who have agreed, in writing, to maintain the confidentiality of the Report. EPRI made a substantial economic investment to develop the Report, and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Report. If the Report and the Proprietary Information were publicly available to consultants and/or other businesses providing services in the electric-and/or nuclear power industry, they would be able to use the Report for their own commercial benefit and profit and without expending the substantial economic resources required of EPRI to develop the Report.
- c. EPRI's classification of the Report and the Proprietary Information as trade secrets is justified by the Uniform Trade Secrets Act which California adopted in 1984 and a version of which has been adopted by over forty states. The California Uniform Trade Secrets Act, California Civil Code §§3426 -3426.11, defines a "trade secret" as follows:
"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:
(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."
- d. The Report and the Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Report only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Report. EPRI was required to devote these resources and effort to derive the Proprietary Information and the Report. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Report is highly valuable to EPRI.
- e. A public disclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRI's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally. The Proprietary Information and Report can only be acquired and/or duplicated by others using an equivalent investment of time and effort.
I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and under the laws of the State of California. Executed at 3420 Hillview Avenue, Palo Alto, CA. 94304 being the premises and place of business of Electric Power Research Institute, Inc. Date: ' j /zo_. Christine King (State of California) Subs ribed and swo to (or affirmed) before me on this:ý day of 20,_-.by proved to me on the basis of satisfactory evidence to be the pe(rlon(s) who a)ppearedr*ý mje. Signature (Seal) ,*-,*! :,_ My Commission Expirea> day of A* .z , 20-L' I
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GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Edward D. Schrull, PE, state as follows: (1) I am the Vice President, Regulatory Affairs, Services Licensing, of GE-Hitachi Nuclear Energy Americas LLC ("GEH"), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in Attachment A of GEH letter, 315497-11, "Pressure Temperature Curve Request (RAI Responses)," dated November 6, 2012. The GEH proprietary information in Attachment A, which is entitled "Pressure Temperature Curve Request (RAI Responses)," is identified by a dotted underline inside double square brackets. ((This sentence is an example..'31 )) In each case, the superscript notation 13) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983). (4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;
- b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;
- d. Information that discloses trade secret and/or potentially patentable subject matter for which it may be desirable to obtain patent protection.
(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my Affidavit for Attachment A of 315497-11 Page I of 3
GE-Hitachi Nuclear Energy Americas LLC knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary and/or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7). (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a "need to know" basis. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, comipetitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements. (8) The information identified in paragraph (2), above, is classified as proprietary because it contains the detailed GEH methodology for pressure-temperature curve analysis for the GEH Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the pressure-temperature curves were achieved at a significant cost to GEH. The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods. The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their Affidavit for Attachment A of 315497-11 Page 2 of 3
GE-Hitachi Nuclear Energy Americas LLC own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools. I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief. Executed on this 6 th day of November 2012. Edward D. Schrull, PE Vice President, Regulatory Affairs Services Licensing GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd. Wilmington, NC 28401 Affidavit for Attachment A of 315497-11 Page 3 of 3}}