ML023460458
| ML023460458 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/10/2002 |
| From: | Joseph E Pollock Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:2349-03 | |
| Download: ML023460458 (30) | |
Text
Indiana Michigan Power Company 500 Circle Drive Buchanan, MI 49107 1395 INDIANA MICHIGAN POWER December 10, 2002 AEP:NRC:2349-03 10 CFR 50.90 10 CFR 50.61 10 CFR 50.60(b) 10 CFR 50.12 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001
SUBJECT:
Donald C. Cook Nuclear Plant, Unit 1 Docket No. 50-315 License Amendment Request for Unit 1 Reactor Coolant System Pressure-Temperature Curves, and Request for Exemption from Requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G
References:
- 1) Letter from J. E. Pollock, Indian Michigan Power Company (I&M), to Nuclear Regulatory Commission (NRC) Document Control Desk, "License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request," AEP:NRC:2900, dated June 28, 2002
- 2) Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Supplement to License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request (TAC No. MB5498),"
AEP:NRC:2900-02, dated October 15, 2002
Dear Sir or Madam:
Pursuant to 10 CFR 50.90, I&M, the licensee for Donald C. Cook Nuclear Plant, Unit 1, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating License DPR-58. I&M proposes to revise the Unit 1 reactor coolant system (RCS) pressure-temperature curves in TS Figures 3.4-2 and 3.4-3 and the associated TS Bases. The revised curves will reflect new limiting reactor
U. S. Nuclear Regulatory Commission AEP:NRC:2349-03 Page 2 vessel materials, will bound operation of the unit for the remainder of its operating license duration, and will bound operation with a planned increase in the licensed maximum power level for the unit. In support of theses changes, I&M is requesting, pursuant to 10 CFR 50.60(b), an exemption from requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G. I&M is also proposing format changes to the affected TS pages that improve their appearance but do not affect any requirements. to this letter provides an affirmation statement pertaining to the requested amendment. Enclosure 2 provides a description of the proposed amendment and a safety analysis, including an evaluation of significant hazards considerations pursuant to 10 CFR 50.92(c) and an environmental assessment. provides TS pages marked to show the proposed changes. provides TS pages with the proposed changes incorporated. provides the results of a revised fluence analysis of surveillance capsule U which was removed from the reactor vessel in 1989. Attachment 4 provides the new RCS pressure-temperature curves and describes their development. Attachment 5 provides an evaluation demonstrating that the Unit 1 reactor vessel beltline materials will continue to meet the pressurized thermal shock criteria of 10 CFR 50.61.
Attachments 3, 4, and 5 also provide information that is applicable beyond the 32 effective full power year (EFPY) of operation assumed for the operating license duration of 40 years. The information that is applicable beyond 32 EFPY is provided for future reference. NRC review and approval of information that is applicable beyond 32 EFPY is not requested at this time.
Attachment 6 contains a request for an exemption from requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G, in support of the new pressure-temperature curves provided in Attachment 4. Attachment 7 provides responses, applicable to Unit 1, to NRC requests for additional information pertaining to a similar amendment proposed for the Unit 2. Attachment 8 provides reactor vessel stress intensity data for Unit 1 similar to that requested by the NRC for Unit 2. Westinghouse Electric Company LLC (Westinghouse) has designated information in Attachment 8 as proprietary pursuant to 10 CFR 2.790.
Attachment 9 provides an affidavit from Westinghouse setting forth the basis on which information contained in Attachment 8 may be withheld from public disclosure. Attachment 10 contains a non-proprietary version of Attachment 8.
U. S. Nuclear Regulatory Commission AEP:NRC:2349-03 Page 3 One pending amendment request affects the TS pages submitted in this amendment request. The pending amendment request is an Appendix K Measurement Uncertainty Recapture (MUR) power uprate amendment request that was submitted by Reference 1 and supplemented by Reference 2.
Reference 2 contains the most recent version of the TS pages that are affected by this amendment request. I&M is requesting that the MUR power uprate amendment request be approved by the NRC prior to approval of the amendment proposed by this letter. Therefore, the amendment proposed by this letter is based on TS pages as modified by Reference 2. If any future submittals affect these TS pages, I&M will coordinate the changes to the pages with the NRC Project Manager to ensure proper TS page control when the associated license amendment requests are approved.
As noted above, I&M requests that the NRC approve the amendment proposed by this letter subsequent to approval of the MUR power uprate amendment.
I&M anticipates that the EFPY applicability limit of the RCS pressure-temperature curves proposed in Reference 2 will be reached in early November, 2004. I&M requests approval of this proposed amendment by September 1, 2004, to provide adequate time for personnel training and procedure changes prior to implementation. Once approved, the amendment will be implemented within 60 days.
There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Brian A. McIntyre, Manager of Regulatory Affairs, at (269) 697-5806.
Sincerely, J. E. Pollock Site Vice President JW/rdw
Enclosures:
1 Affirmation 2 License Amendment Request for Unit 1 Reactor Coolant System Pressure-Temperature Curves
U. S. Nuclear Regulatory Commission AEP:NRC:2349-03 Page 4 Attachments:
1 Technical Specification Pages Marked to Show Proposed Changes 2 Proposed Technical Specification Pages 3 WCAP-12483, Revision 1, "Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit I Reactor Vessel Radiation Surveillance Program," dated December 2002 4 WCAP-15878, Revision 0, "D. C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation for 40 Years and 60 Years," dated December 2002 5 WCAP-15879, Revision 0, "Evaluation of Pressurized Thermal Shock for D. C. Cook Unit 1 for 40 Years and 60 Years," dated December 2002 6 Request for Exemption from Requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G 7 Responses, Applicable to Unit 1, to NRC Requests for Additional Information Regarding Similar Unit 2 Amendment Request 8 Westinghouse Letter LTR-EMT-02-316, "Thermal Stress Intensity Factors for D. C. Cook Unit 1 PT Curves (Proprietary Version)," dated November 12, 2002 9 Westinghouse Letter CAW-02-1572, "Application for Withholding Proprietary Information from Public Disclosure," dated November 12, 2002 10 Westinghouse Letter LTR-EMT-02-319, "Thermal Stress Intensity Factors for D. C. Cook Unit 1 PT Curves (Non-Proprietary Version),"
dated November 12, 2002 c: K. D. Curry, Ft. Wayne AEP J. E. Dyer, NRC Region m MDEQ - DW & RPD NRC Resident Inspector J. F. Stang, Jr., NRC Washington, DC R. Whale, MPSC
U. S. Nuclear Regulatory Commission AEP:NRC:2349-03 Page 5 bc: A. C. Bakken III M. J. Finissi S. A. Greenlee D. W. Jenkins, w/o attachments J. A. Kobyra, w/o attachments B. A. McIntyre, w/o attachments J. E. Newmiller J. E. Pollock D. J. Poupard T. Satyan-Sharma/P. G. Schoepf M. K. Scarpello, w/o attachments T. K. Woods, w/o attachments
Enclosure 1 to AEP:NRC:2349-03 AFFIRMATION I, Joseph E. Pollock, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.
Indiana Michigan Power Company J. E. Pollock Site Vice President JULIE E.NEWMILLER Notary Public, Berrien County, MI My Commission Expires Aug 22, 2004 (1/ - k /
..... My Commission Expires
.'-C-, -- -. - - ,- --
to AEP:NRC:2349-03 Page I LICENSE AMENDMENT REOUEST FOR UNIT 1 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE CURVES References for this enclosure are identified below, in Section 8.
1.0 DESCRIPTION
Pursuant to 10CFR50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP), Unit 1, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating License DPR-58. I&M proposes to revise the Unit 1 reactor coolant system (RCS) pressure-temperature curves in TS Figures 3.4-2 and 3.4-3 and the associated TS Bases. The revised curves will reflect new limiting reactor vessel materials, will bound operation of the unit for the remainder of its operating license duration, and will bound operation with a planned increase in the licensed maximum power level for the unit. In support of theses changes, I&M is requesting, pursuant to 10 CFR 50.60(b), an exemption from requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G. I&M is also proposing format changes to the affected TS pages that improve their appearance but do not affect any requirements.
2.0 PROPOSED CHANGE
I&M previously submitted a proposed Appendix K Measurement Uncertainty Recapture (MUR) power uprate amendment request (Reference 1). The change proposed by this letter is based on the TS pages as modified by a supplement to the MUR power uprate amendment, submitted by Reference 2.
The curves in TS Figures 3.4-2 and 3.4-3 specify limits on RCS pressure and temperature for heatup, cooldown, criticality, and inservice leak and hydrostatic testing. The proposed amendment will revise these curves such that they reflect new limiting reactor vessel materials consisting of intermediate shell axial weld heat 13253/12008 and lower shell plate B4407-3. The proposed amendment will also revise these curves such that they bound operation of the reactor for up to 32 effective full power years (EFPY) at a power level of up to 3600 megawatts-thermal (MWt) for the current fuel cycle, Cycle 18, and beyond. The revised curves also reflect the use of a new fluence analysis methodology specified by Regulatory Guide (RG) 1.190 (Reference 3),
reflect the use of American Society of Mechanical Engineers (ASME) Code Case N-641, include boltup limits, and do not include instrument uncertainty margins. The proposed amendment will also change the titles, labels, and orientation of TS Figures 3.4-2 and 3.4-3, and will change the TS Bases for Section 3/4.4.9 to be consistent with the revised curves.
to AEP:NRC:2349-03 Page 2 I&M also proposes three types of format changes to the revised TS pages. The changes are:
"* Reformatting of the headers to include numbered first and second-tier TS section titles.
"* Reformatting of the footers to include "Page (page number)" center page, and a full-width single line to separate the footer from the page text.
"* Fully justifying the text and changing the font. to this letter provides TS pages from Reference 2 that are marked to show the proposed changes (except for the above described format changes). Attachment 2 provides TS pages with the proposed changes incorporated.
3.0 BACKGROUND
Basis for RCS Pressure-Temperature Curves All components in the reactor coolant pressure boundary are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are identified in Section 4.1.4 of the Updated Final Safety Analysis Report (UFSAR). The curves in TS Figures 3.4-2 and 3.4-3 establish operating limits that provide a margin to brittle failure for the reactor coolant pressure boundary, considering the effect of these cyclic loads. These curves limit the rate of temperature and pressure change during startup and shutdown so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. These limits apply mainly to the reactor pressure vessel since it is the component most subject to brittle failure.
The heatup limit curve provided in TS Figure 3.4-2 is a composite curve which is prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60 degrees Fahrenheit per hour. The cooldown limit curves provided in TS Figure 3.4-3 are composite curves which are prepared based upon the same type analysis, with the exception that the controlling location is always the inside wall, where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall.
The heatup and cooldown curves are based on the most limiting value of the predicted adjusted reference temperature (ART). The neutron embrittlement effect on the ART is addressed by periodically removing and evaluating one of the reactor vessel surveillance capsules and adjusting the heatup and cooldown limit curves as necessary.
The RCS pressure-temperature curves submitted in Reference 2 specify the same pressure-temperature restrictions as the previous curves, approved by the Nuclear Regulatory Commission (NRC) in 1992 (Reference 4). The curves approved in 1992 were based on reactor vessel intermediate plate B4406-3 as the limiting reactor vessel material, and specified an applicability limit of 32 EFPY, corresponding to the operating license duration of 40 years. As to AEP:NRC:2349-03 Page 3 documented in 1999 (Reference 5), I&M subsequently determined that intermediate plate B4406-3 would cease to be the limiting material for the Unit 1 reactor vessel prior to reaching the 32 EFPY applicability limit. In October 2002, I&M proposed (Reference 2) reducing the 32 EFPY applicability limit specified in TS Figures 3.4-2 and 3.4-3 such that the curves would be valid only for the period in which intermediate plate B4406-3 was the limiting reactor vessel material (18.6 EFPY). The applicability limit of 18.6 EFPY proposed in Reference 2 was determined from a fluence analysis that assumed a maximum power level of 3315 MWt, which bounded the MUR uprate power level. That fluence analysis was also submitted in Reference 2.
The change in EFPY applicability limit was the only change to TS Figures 3.4-2 and 3.4-3 proposed in Reference 2.
Reason for Requesting Amendment I&M committed, in Reference 2, to submit a proposed license amendment containing new Unit 1 RCS pressure-temperature curves that reflect the limiting reactor vessel material for operation beyond 18.6 EFPY. By reflecting the appropriate limiting material, the EFPY applicability specified in the pressure-temperature curves may be restored to the 32 EFPY value that corresponds to the operating license duration of 40 years. Also, I&M plans to request NRC approval of an additional increase in the licensed power level for Unit 1, above the power level requested for the MUR power uprate in Reference 1 and Reference 2. Therefore, I&M is proposing revised curves that are based on the appropriate reactor vessel limiting material, are applicable for 32 EFPY, and are valid for operation at a power level of up to 3600 MWt for the current fuel cycle and beyond (which will bound the planned additional increase in the licensed power level).
4.0 TECHNICAL ANALYSIS
The proposed amendment is supported by Attachments 3 through 10 to this letter. These attachments are described below. provides a copy of Westinghouse Electric Company LLC (Westinghouse)
WCAP-12483, Revision 1, "Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program." This WCAP documents the Charpy V-Notch testing, tensile testing, and associated analyses that were performed on surveillance capsule U following its removal from the reactor vessel. Capsule U was removed in 1989 after a total of 9.17 EFPY of operation. The WCAP also documents an analysis to determine the neutron radiation environment within the reactor vessel, including projections of future neutron exposure.
to AEP:NRC:2349-03 Page 4 This WCAP is a revision to the original WCAP-12483, which was submitted in support of the 1992 amendment (Reference 4) that updated the Unit 1 RCS pressure-temperature curves prior to Reference 2. Revision 1 to WCAP-12483 includes an update of the fluence analysis methodology described in Section 6 of the WCAP, "Radiation Analysis and Neutron Dosimetry."
This section was revised to reflect the fluence analysis methodology specified in RG 1.190, which was issued subsequent to the original WCAP-12483. Revision I to WCAP-12483 includes a new Sub-section 6.4, "Projections of Reactor Vessel Exposure." This sub-section includes fluence projections for 32 EFPY of operation, at a power level of up to 3600 MWt for the current fuel cycle and beyond. Revision 1 to WCAP-12483 also includes relocation of the RCS pressure-temperature curves. The RCS pressure-temperature curves were incorporated into WCAP-15878, Revision 0, which is provided as Attachment 4 to this letter.
As documented in Section 1 of the WCAP, "Summary of Results," the surveillance capsule materials exhibit an upper shelf energy level that is more than adequate for continued safe unit operation, and are expected to retain an upper shelf energy of greater than 50 ft-lb. throughout the life (32 EFPY) of the vessel as required by 10 CFR 50, Appendix G. provides a copy of Westinghouse WCAP-15878, Revision 0, "D. C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation for 40 Years and 60 Years." This WCAP provides the RCS pressure-temperature curves that are proposed as revisions to TS Figures 3.4-2 and 3.4-3, and describes how the curves were developed. These curves are based, in part, on fluence data from the revised analysis of surveillance capsule U documented in Attachment 3 to this letter. The fluence data includes projections for 32 EFPY, assuming operation at a power level of up to 3600 MWt for the current fuel cycle and beyond. These fluence projections have been used to calculate ART values in accordance with the guidance provided in RG 1.99 (Reference 6), as described in Section 8.0 of the WCAP. The ART values were used to develop RCS pressure-temperature curves using the methodology described in Section 3.0 of the WCAP. As noted in Section 3.0 of the WCAP, the methodology of Code Case N-641 has been used in development of the RCS pressure-temperature curves. I&M is requesting an exemption from requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G, to allow use of Code Case N-641. Attachment 6 to this letter documents the requested exemption and the justification.
The proposed RCS pressure-temperature curves differ from those contained in Reference 2 as follows:
The proposed curves specify new limiting materials. The pressure-temperature curves submitted in Reference 2 are based on intermediate shell plate B4406-3 as the limiting material. Although Reference 5 identified circumferential weld heat 1P3571 as a new limiting material, that determination was made without the application of ASME Code to AEP:NRC:2349-03 Page 5 Section XI Code Case N-641. As stated in Section 8.0 of the WCAP, Code Case N-641 allows less restrictive methodology when a circumferential weld has the highest ART.
Therefore, use of Code Case N-641 results in axial weld 13253/12008 and lower shell plate B4407-3 being more limiting (for 1/4T and 3/4T respectively) than circumferential weld heat 1P3571, even though the circumferential weld was determined to have a higher ART. As a result, axial weld 13253/12008 and lower shell plate B4407-3 are the limiting materials, as indicated on the proposed curves.
The proposed curves have a higher EFPY applicability limit. The EFPY applicability limit of the pressure-temperature curves submitted in Reference 2 is 18.6 EFPY. As described in Reference 2, this applicability limit was determined to be the exposure at which intermediate shell plate B4406-3 would no longer be the limiting material, and the curves would, therefore, no longer be valid. The proposed curves were developed based on 32 EFPY, corresponding to the 40 year operating license duration.
The proposed curves bound operation at a higher power level. The pressure-temperature curves submitted in Reference 2 bound operation up to 3315 MWt. The proposed curves were developed so as to bound operation up to 3600 MWt for the current fuel cycle and beyond. This higher power level will bound a planned additional increase in the licensed power level of the unit.
The pressure-temperature relationships established by the proposed curves are less restrictive.
Use of the new fluence analysis methodology specified in RG 1.190 and use of stress intensity factors determined in accordance with Code Case N-641 has resulted in less restrictive pressure-temperature relationships than are specified by the curves submitted in Reference 2.
The proposed curves do not include a margin for instrument uncertainty. This is consistent with the NRC approved methodology documented in WCAP 14040-NP-A (Reference 7).
Instrument uncertainty margins are incorporated in RCS pressure-temperature limits specified in CNP procedures.
The proposed curves include limits for bolting the RPV head. Inclusion of the boltup limits is consistent with WCAP 14040-NP-A.
Section 10.0 of the WCAP, "Enable Temperature Calculation," provides a revised low-temperature overpressure protection system enable temperature limit. I&M has determined that, since the revised RCS pressure-temperature curves are less restrictive than the curves submitted in Reference 2, the basis for the currently licensed enable temperature remains valid.
Therefore, I&M is not proposing changes to TS related to the low-temperature overpressure protection system.
to AEP:NRC:2349-03 Page 6 provides a copy of Westinghouse WCAP-15879, Revision 0, "Evaluation of Pressurized Thermal Shock for D. C. Cook Unit I for 40 Years and 60 Years." The previous pressurized thermal shock evaluation, submitted to the NRC by Reference 8, was based, in part, on the fluence values in the original WCAP-12483. Similarly, the pressurized thermal shock evaluation documented in Attachment 5 is based, in part, on the fluence projections documented in the current revision of Attachment 3 to this letter. As described in Attachment 5, these fluence projections have been used to determine that the pressurized thermal shock criteria of 10 CFR 50.61 will be satisfied for 32 EFPY of operation, at a power level of up to 3600 MWt for the current fuel cycle and beyond. Since Code Case N-641 is not applicable to pressurized thermal shock evaluations, the limiting material is circumferential weld heat 1P3571.
In support of Attachment 4, I&M is requesting exemption from certain requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G. Attachment 6 provides the exemption request.
In July 2002, I&M submitted a proposed amendment to revise the CNP Unit 2 RCS pressure-temperature curves. The NRC subsequently requested additional information regarding the proposed Unit 2 amendment. Since the amendment proposed for Unit 1 by this letter is similar to that proposed for Unit 2, I&M has provided responses, applicable to the proposed amendment for Unit 1, to the NRC requests for information regarding the amendment proposed for Unit 2. The responses applicable to Unit 1 are provided in Attachment 7.
Attachment 8 A portion of the Unit 1 information provided to address the NRC requests for information regarding the Unit 2 amendment has been designated as proprietary by Westinghouse.
Attachment 8 provides the proprietary portion of the requested information.
Attachment 9 Pursuant to 10 CFR 2.790, Westinghouse has prepared an application to withhold the proprietary information in Attachment 8 from public disclosure. Attachment 9 provides this application.
to AEP:NRC:2349-03 Page 7 0 0 provides a non-proprietary version of Attachment 8 that may be made available to the public.
Summary The technical analysis of the proposed TS changes has determined the following:
"* The revised RCS pressure-temperature curves are based on the appropriate limiting reactor vessel materials.
"* The revised RCS pressure-temperature curves will remain valid for the duration of the operating license.
"* The revised RCS pressure-temperature curves are based on a power level that bounds the planned additional increase in the licensed power level.
"* The surveillance capsule materials exhibit a Charpy upper shelf energy level that is more than adequate, and will retain an upper shelf energy of greater than 50 ft-lb. as required by 10 CFR 50, Appendix G for the duration of the operating license, including the planned additional increase in licensed power level.
"* The currently licensed low-temperature overpressure protection system enable temperature will remain valid since the revised RCS pressure-temperature curves are less restrictive than the curves submitted in Reference 2.
"* The pressurized thermal shock criteria of 10 CFR 50.61 are satisfied for the duration of the operating license, including the planned additional increase in licensed power level.
Additionally, the NRC requests for additional information regarding a similar amendment proposed for Unit 2 have been addressed for the proposed Unit 1 amendment.
5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration I&M has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?
Response: No Probability of Occurrence of an Accident Previously Evaluated to AEP:NRC:2349-03 Page 8 The proposed change will revise the RCS pressure-temperature curves to reflect new limiting reactor vessel materials, to bound operation of the reactor up to 3600 MWt for the current fuel cycle and beyond, to reflect new fluence analysis methodology, to reflect the use of ASME Code Case N-641, to include boltup limits, and to no longer include instrument uncertainty margins.
The proposed change will not result in physical changes to structures, systems, or components (SSCs), or changes to event initiators or precursors. The proposed change will not affect the ability of personnel to control RCS pressure at low temperatures and, thereby, ensure the integrity of the reactor coolant pressure boundary. Use of Code Case N-641 in developing the proposed revision to the RCS pressure-temperature curves is in accordance with methodologies accepted by the ASME. These methodologies provide assurance that the reactor vessel will withstand the effects of normal cyclic loads due to temperature and pressure changes, and provide an acceptable level of protection against brittle failure.
Additionally, the proposed changes will not impact the design or operation of plant systems such that previously analyzed SSCs will be more likely to fail. The initiating conditions and assumptions for accidents described in the UFSAR will remain as previously analyzed. Therefore, the proposed changes will not involve a significant increase in the probability of an accident previously evaluated.
Consequences of an Accident Previously Evaluated The proposed change does not reduce the ability of any SSC to limit the radiological consequences of accidents described in the UFSAR. The proposed change will not alter any assumptions made in the analysis of radiological consequences of previously evaluated accidents, nor does it affect the ability to mitigate these consequences. No new or different radiological source terms will be generated as a result of the proposed change. Therefore, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.
The format changes will improve the appearance of the affected pages but will not affect any requirements. In summary, the probability of occurrence and the consequences of an accident previously evaluated will not be significantly increased.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No to AEP:NRC:2349-03 Page 9 The proposed change will not result in physical changes to SSCs. The proposed change will not involve the addition or modification of plant equipment (no new or different type of equipment will be installed) nor will it alter the design of any plant systems. The proposed change solely involves RCS pressure-temperature limits. The types of potential accidents associated with these limits have been previously identified and evaluated. No new accident scenarios, accident or transient initiators or precursors, failure mechanisms, or single failures will be introduced as a result of the proposed changes. No new or different modes of failure will be created. The format changes will improve the appearance of the affected pages but will not affect any requirements. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No The proposed RCS pressure-temperature curves will continue to provide adequate margins of protection for the reactor coolant pressure boundary. The proposed changes have been determined, through supporting analyses, to be in accordance with the methodologies and criteria set forth in the applicable regulations, or in accordance with technically adequate alternatives. Compliance with these methodologies provides adequate margins of safety and ensures that the reactor coolant pressure boundary will withstand the effects of normal cyclic loads due to temperature and pressure changes as well as the loads associated with postulated faulted events as described in the UFSAR. The format changes will improve the appearance of the affected pages but will not affect any requirements. Therefore, the proposed change will not significantly reduce the margin of safety.
In summary, based upon the above evaluation, I&M has concluded that the proposed changes involve no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements/Criteria The proposed amendment will revise TS Figures 3.4-2 and 3.4-3, and their associated Bases. The curves in these figures specify limits on RCS pressure and temperature during heatup, cooldown, criticality and inservice leak and hydrostatic testing.
to AEP:NRC:2349-03 Page 10 The proposed curves were developed based on an updated fluence analysis. The proposed amendment is also supported by an evaluation demonstrating compliance with the pressurized thermal shock screening criteria. Except as noted, the capsule analysis, curve development, and pressurized thermal shock evaluation have been conducted in accordance with 10 CFR 50.60(a), 10 CFR 50.61, 10 CFR 50, Appendix G, RG 1.190, and RG 1.99. Approval of the noted exception to 10 CFR 50.60(a) and 10 CFR 50, Appendix G, has been requested and justified in Attachment 6 to this letter in accordance with 10 CFR 50.60(b) and 10 CFR 50.12. The technical analysis supporting the proposed amendment demonstrates that it does not involve significant hazards considerations as described in 10 CFR 50.92.
Compliance with other regulations or TS will not be affected by the proposed amendment.
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 NRC 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.
6.0 Environmental Considerations I&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. I&M has determined that the proposed amendment will 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 will 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 needs to be prepared concerning the proposed amendment.
7.0 Precedent Licensing Actions The NRC has approved, by Reference 4, a previous revision to the Unit 1 RCS pressure-temperature curves based on an earlier revision of the analysis provided in Attachment 3 to this letter.
to AEP:NRC:2349-03 Page 11 8.0 References
- 1. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request,"
AEP:NRC:2900, dated June 28, 2002
- 2. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Supplement to License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request (TAC No. MB5498)," AEP:NRC:2900-02, dated October 15, 2002
- 3. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2002
- 4. Letter from J. F Stang, NRC, to E. E. Fitzpatrick, I&M, "Donald C. Cook Nuclear Plant, Unit No 1 - Amendment No. 167 to Facility Operating License No. DPR-58 (TAC No. M71480 and M75260)," dated October 26, 1992
- 5. Letter from M. W. Rencheck, I&M, to NRC Document Control Desk, "Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity at Donald C. Cook Nuclear Plant, Unit 1, TAC No. MA0539," C0699-01, dated June 28, 1999
- 6. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2, dated May 1988
- 7. WCAP 14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," dated January 1996
- 8. Letter from M. P. Alexich, I&M, to NRC Document Control Desk, "Updated Reference Temperature, Pressurized Thermal Shock Analysis," AEP:NRC:0561D, dated August 7, 1990
Attachment 1 to AEP:NRC:2349-03 TECHNICAL SPECIFICATIONS PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 1 3/4 4-27 3/4 4-28 B3/4 4-6 B3/4 4-7
REACTOR COOLANT SYSTEM HEATUP LIMITATIONS D
APPLICABLE FOR FIRST18.6EFFECTIVE FULL POWER 2400 0 EARS(MARGINS OF S0 PSIG AND 10 F ARE INCLUDED
'-I I-! FOD OSSIBLE INSTRUMENT ERROR.)
2200 2000 ,
LEAK TEST LIMIT--".-
--4 1800 UNACC PADLE ACCEPTABLE OPERATIDON, OPERATION 1400 1200 PRESSURE-TEMPERATURE - ,______
1000 C-CRITICALITY LIMIT FOR HEATUP RATES UP TO 6OF/HnR---"-- LIMIT 800 600 MATER PROPERTY BASIS INTERMED E PLATE. D4406-3 S,> --Cu - .15 . * .49 X 400 INITIAL RT NOT 40OF 200 ___- 18.6 EFPY RTNoT 1/4 -171:F C
Eli z
D
.... NOT 3/4
, T_ f*.
CO)..
f*
n
_.ýZ3 i00 200 250U ,.UU JoU -1UU LI
-g 0 *VER8E -REAeTGR eBIANT SYSTEM TEHFER*ATUREF rF FIGURE 3.4-2 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS VERSUS d)F/HR RATE
,5 .... O.TAT. TEST bLIII I CRITICALITY LIMIT
L Insert 1 Reactor Coolant System Heatup Limitations Without Margins for Instrumentation Error Applicable for 32 EFPY of Operation Limiting Material: Axial Weld 13253112008, Cu = 0.21%, Ni = 0.873%, and Lower Shell Plate B4407-3, Cu = 0.14%, Ni = 0.50%
Initial ART: 58 Deg. F, Limiting ART Values at 32 EFPY: 114T = 199 Deg. F, 3/4T = 143 Deg. F 2500 2250 2000 (3 1750 o3 a,
[L F
1500 (n
a.,
1..
E 1250 (n
0 1000 0
C.,
750 500 250 0
200 - 250 300 350 400 450 500 550 0 50 100 150 Average Reactor Coolant System Temperature (Deg. F)
00 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS C APPLICABLE FOR FIRST18.6EFFECTIVE FULL POWER S YEARS (MARGIHS OF 60 PSIG AND lO°F ARE INCLUDED CP F POSSIBLE INSTRUMENT EnnoR.) ._ _/
-' 2200 I 2000 DO0 *UHNACCEPTABLE 0
ORATION
-- , OPERATION
- ,1400 t
- 16000 1A2oo 12,
_______________OPRTO LIMT ITS . ,,
BOO _ ---- .. . // "
600 COOLDOWN RATE "F/Han MATER'L PROPERTY BASIS 400-..- 0 .- * -tcd - 1i , lPLATE..49 INTERMED2TE B4406-3 0F 200 --,__0__-- INITIAL RTNDT I40 18.6 EFPY RT I -171 0 F 200__00___
ZIT 0 100: 150 '200 250 300 350 400 44
' A.ERA.E sEACTem COOLA!T ......
,. TEMPERATUHE- f FIGURE 3.4-3 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS.VERSUS COOLDODWM,nATES c3Z
I Insert 2 Reactor Coolant System Cooldown Limitations Without Margins for Instrumentation Error Applicable for 32 EFPY of Operation Limiting Material: Axial Weld 13253112008, Cu = 0.21%, Ni = 0.873%, and Lower Shell Plate B4407-3, Cu = 0.14%, Ni = 0.50%
Initial ART: 58 Deg.'F, Limiting ART Values at 32 EFPY: 1/4T = 199 Deg. F, 314T = 143 Deg. F 2500 2250 Un-lpeainacceptable-- _tAcpal 2000 Oprto 200 Acpal OpOperation 1750 0
In U)
In 1500 a.
E 1250 U) 0 0 1000 0
ca Cooldown Rate 750 - Deg. FIHr.
40 500 60 100
- 250 Temp.
0
0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (Deg. F)
3/4 BASES 3/4.3 REACTOR COOLANT SYSTEM 3/4.49 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case. There are several factors which influence the postulated location The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall. During cooldown the bending stress profile is reversed. In addition, the material toughness is dependent upon irradiation rate and temperature and therefore the fluence profile through the reactor vessel wall, the rate of heatup and also the of cooldown influence the postulated flaw location.
The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60°F per hour. The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown ctirves were prepared based upon the most limiting value of the predicted adjusted reference temperature c _
it*AT at the end of EFPY.
2-8-6-v Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, and the copper and nickel content of the material must be predicted. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include the adjusted RTNDT at the end of 4--6 L2 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.
AMENDMENT 88.467 1 Page B 3/4 4-6 AMENDMENT 88,4-6-7 COOK COOK NUCLEAR PLANT-UNIT I Page B 3/4 4-6
314 BASES 3/4.3 REACTOR COOLANT SYSTEM The 4-&.6 12 EFPY heatup and cooldown curves were developed based on the following:
I. The intermediatecel lae B16 3, being thp l11---_ -at-.4,1 _4h+ ope and nickel eentent ef15 and .19%, rweeId 13253/28 2lower 7ith'ope nt 0.-8Iýk 73%0respe and _ andnkei con*et*f0A4%° and 0.50iespec -lOiveleO
- 2. The fluence values doeumented in I&N letter- AEP.NRC; 2900 02co-ardi~aý1ýZ14' eact r ss1.madiatiori SurveL]anc qo i" dated Decemberi20021
- 3. Figure 1, NRC Regulatory Guide 1.99, Revision 2 The shift in RTNDT of the reactor vessel material has been established by removing and evaluating the material surveillance capsules installed near the inside wall of the reactor vessel in accordance with the removal schedule in Table 4.4-5. Per this schedule, Capsule U is the last capsule to be removed until Capsule S is to be removed after 32 EFPY (EOL). Capsule V, W, and Z will remain in the reactor vessel, and will be removed to address industry reactor embrittlement concerns, if required.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two PORVs, or of one PORV and the RHR safety valve, ensures that the RCS will be protected from pressure transients which could exceed the lirmts of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 1520F. Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50°F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS. Therefore, any one of the three blocked open PORVs constituted an acceptable RCS vent to preclude APPLICABILITY of Specification 3.4.9 3.
COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 4-7 AMENDMENT 88, 16-7, 4-76 r
Attachment 2 to AEP:NRC:2349-03 PROPOSED TECHNICAL SPECIFICATIONS PAGES REVISED PAGES UNIT 1 3/4 4-27 3/4 4-28 B3/4 4-6 B3/4 4-7
3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM Reactor Coolant System Heatup Limitations Without Margins for Instrumentation Error Applicable for 32 EFPY of Operation Limiting Material: Axial Weld 13253112008, Cu = 0.21%, Ni = 0.873%, and Lower Shell Plate B4407-3, Cu = 0.14%, Ni = 0.50%
Initial ART: 58 Deg. F, Limiting ART Values at 32 EFPY: 1/4T= 199 Deg. F, 314T = 143 Deg. F 2500 2250 2000 1750
' 1500 E
S1250 0.
-* 1000 0
0 u 750 Ca W,
500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (Deg. F)
FIGURE 3.4-2 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS VERSUS 60 0 F/HR RATE, CRITICALITY LIMIT, BOLTUP LIMIT, AND LEAK TEST LIMIT COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-27 AMENDMENT 98,-6-7,
3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM I Reactor Coolant System Cooldown Limitations Without Margins for Instrumentation Error Applicable for 32 EFPY of Operation Limiting Material: Axial Weld 13253/12008, Cu = 0.21%, Ni = 0.873%, and Lower Shell Plate B4407-3, Cu = 0.14%, Ni = 0.50%
Initial ART: 58 Deg.*F, Limiting ART Values at 32 EFPY: 114T = 199 Deg. F, 314T = 143 Deg. F 2500 2250 2Unacceptable 2000 Operation Opeption S1750 09 I
0 1500 W..
E S1250 0) oo 1000 C.,
o Cooldown Rate
- 0) 750 Deg. FIHr. - -_____
500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (Deg. F)
FIGURE 3.4-3 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS VERSUS COOLDOWN RATES COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-28 AMENDMENT 88,4-6-7,
3/4 BASES 3/4.3 REACTOR COOLANT SYSTEM 3/4.49 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case. There are several factors which influence the postulated location. The thermal induced bending stress During during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall.
cooldown the bending stress profile is reversed. In addition, the matnrial toughness is dependent upon irradiation rate and temperatuire and therefore the fluence profile through the reactor vessel wall, the rate of heatup and also the of cooldown influence the postulated flaw location.
The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative limit case, with either the inside or outside wall controlling, for any heatup rate up to 60°F per hour. The cooldown prepared based upon the same type analysis with the curves of Figure 3.4-3 are composite curves which were inside wall where the cooldown thermal gradients tend to exception that the controlling loc'ation is always the stresses at the outside wall. The heatup and cooldown curves produce tensile stresses while producing compressive were prepared based tipon the most limiting value of the predicted adjusted reference temperature for the controlling materials at the end of 32 EFPY.
Reactor operation and resultant fast neutron (E > I Mev) irradiation will cause an increase in the RTNDT. Therefore, be an adjusted reference temperature, based upon the fluence, and the copper and nickel content of the material must limit curves of Figures 3.4-2 and 3.4-3 include the adjusted RTNDT at the end of predicted. The heatup and cooldown 32 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.
L.
COOK NUCLEAR PLANT-UNIT 1 Page B 3!4 4-6 AMENDMENT 88,1-6-7,
3/4 BASES 3/4.3 REACTOR COOLANT SYSTEM The 32 EFPY heatup and cooldown curves were developed based on the following:
- 1. The limiting materials being axial weld 13253/12008, with a copper and nickel content of 0.21% and 0.873%, respectively, and lower shell plate B4407-3, with a copper and nickel content of 0.14% and 0.50%, respectively.
- 2. The fluence values contained in Table 6-14 of WCAP-12483, Revision 1, "Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program,"
dated December 2002.
- 3. Figure 1, NRC Regulatory Guide 1.99, Revision 2 The shift in RTNDT of the reactor vessel material has been established by removing and evaluating the material in surveillance capsules installed near the inside wall of the reactor vessel in accordance with the removal schedule Capsule S is to be removed after 32 Table 4.4-5. Per this schedule, Capsule U is the last capsule to be removed until EFPY (EOL). Capsule V, W, and Z will remain in the reactor vessel, and will be removed to address industry reactor embrittlement concerns, if required.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
be The OPERABILITY of two PORVs, or of one PORV and the RHR safety valve, ensures that the RCS will protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 1520F. Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle leg RCP with the secondary water temperature of the steam generator less than or equal to 50°F above the RCS cold temperatures or (2) the start of a charging pump and its injection into a water solid RCS. Therefore, any one of the three blocked open PORVs constituted an acceptable RCS vent to preclude APPLICABILITY of Specification 3.4 9.3.
AMENDMENT 88,467,4-76, COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 4-7
Attachment 3 to AEP:NRC:2349-03 WCAP-12483, REVISION 1, "ANALYSIS OF CAPSULE U FROM THE AMERICAN ELECTRIC POWER COMPANY D. C. COOK UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM,"
DATED DECEMBER 2002