ML17328A751

Application for Amend to License DPR-58,revising Tech Specs to Reflect Results of Reactor Vessel Capsule U Analysis, Consisting of Revised Heatup & Cooldown Curves for First 32 EFPYs

ML17328A751
Person / Time
Site:	Cook
Issue date:	10/29/1990
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17328A752	List: ML17328A751 ML17328A753
References
AEP:NRC:08940, AEP:NRC:8940, NUDOCS 9011020041
Download: ML17328A751 (9)

Text

ACCELERATED D. UTION DEMONS TION SYSTEM REGULAT Y INFORMATION DISTRIBUTIO SYSTEM (RIDS)

ACCESSION NBR:9011020041 DOC.DATE: 90/10/29 NOTARIZED: NO DOCKET FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana a 05000315 AUTH. NAME AUTHOR AFFILIATION ALEXICH,M.P. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION MURLEYZT.E. Document Control Branch (Document Control Desk)

SUBJECT:

Application for amend to License DPR-58,revising Tech Specs to reflect results of reactor vessel Capsule U analysis.

DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR ( ENCL g SIZE: 7+ IC TITLE: OR Submittal: General Distribution NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 COLBURN,T. 2 2 INTERNAL: NRR/DET/ECMB 9H 1 1 NRR/DOEA/OTSB11 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 1 0 OGC/HDS1 1 0 REG FILE 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WAS'ONTACI'HE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 17 ENCL 15

Indiana Michigan Power Company

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P.O. Box 16631 Columbus, OH 43216 AEP:NRC:08940 Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315 License No. DPR-58 TECHNICAL SPECIFICATION CHANGE REQUEST REVISED HEATUP AND COOLDOWN, AND LTOP SETPOINT FOR THE FIRST 32 EFFECTIVE FULL POWER YEARS U. ST Nuclear Regulatory Commission Attn: Document Control Desk Washington, DE C. 20555 Attn: T. E. Murley October 29, 1990

Dear Mr. Murley:

This letter and its attachments constitute an application for a technical specification (T/S) change for Donald C. Cook Nuclear Plant Unit 1. Specifically, we request that the heatup and cooldown curves, and the low temperature overpressurization protection (LTOP) setpoint be revised to reflect the results of Unit 1 reactor vessel Capsule U analysis, Westinghouse report No. WCAP-12483 (submitted in our letter AEP:NRC:0894M dated June 22, 1990). The heatup and cooldown curves have been developed for the most limiting material up to the first 32 effected full power years (EFPY).

Our evaluation concerning significant hazards considerations are provided in Attachment 1. The proposed revised T/S pages are included in Attachment 2.

We believe that the proposed change will not result in (1) a significant change in the types of effluents or a significant increase in the amounts of any effluent that may be released offsite, or (2) a significant increase in individual or cumulative occupational radiation exposure.

The proposed changes have been reviewed by the Plant Nuclear Safety Review Committee. The Nuclear Safety and Design Review Committee will review, these\

changes at their next regularly scheduled meeting.

90ii02004i 90i029 PDR ADOCK 05000315 P PNU pl

Dr. T. E. Murley AEP:NRC:08940 In compliance with the requirements of 10 CFR 50.91(b)(1), copies of this letter and its attachments have been transmitted to Mr. J. R.

Padgett of the Michigan Public Service Commission and to the Michigan Department of Public Health.

This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature of the undersigned.

Sincerely, M. P. Ale ich Vice President MPA/eh Attachments cc: D. H. Villiams, Jr.

A. A. Blind - Bridgman J. R. Padgett G. Charnoff NFEM Section Chief A. B. Davis - Region III NRC Resident Inspector - Bridgman

gT 4

ATTACHMENT 1 TO AEP:NRC:08940 REASONS AND 10 CFR 50 '2 ANALYSIS FOR CHANGES TO THE DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATIONS

ATTACHMENT 1 TO AEP:NRC:08940 Page 1 Introduction The revised T/S heatup and cooldown curves, and LTOP setpoint were developed based upon the results of the reactor vessel Capsule U analysis reported in the Westinghouse report No. WCAP-12483, as per the requirements of Reg Guide 1.99, Rev. 2 (submitted in our letter AEP:NRC:0894M dated June 22, 1990). That analysis determined that the limiting reactor vessel material is no longer the weld metal but the base metal (intermediate shell plate B4406-3), with a copper and nickel content of .15X and .49X, respectively. Also, the current T/S heatup and cooldown curves were prepared considering the most limiting value of the predicted adjusted reference temperature at the end of 12 EFPY. This revision to the T/S heatup and cooldown curves has considered 32 EFPY.

As noted above, the reactor vessel Capsule U analysis determined that the base metal is now more limiting than the weld metal.

Therefore, we are requesting that the following pages be revised to reflect the new limiting reactor vessel material, and to reflect the recalculated LTOP temperature setpoint based on 32 EFPY: 3/4 1-11, 3/4 l-lla, 3/4 4-3, 3/4 4-31, 3/4 5-7, 3/4 5-8, B 3/4 1-2, B 3/4 4-1, B 3/4 4-6, B 3/4 4-7, and B 3/4 5-2.

T/S pages 3/4 4-27 and 3/4 4-28 contain T/S Figures 3.4-2 and 3.4-3, the heatup and cooldown curves, which have been revised to reflect the results of the Capsule U analysis contained in WCAP-12483.

We are also requesting an editorial change to obtain consistency throughout the technical specifications. Specifically, we request that the maximum heatup rate be changed from 100 F/hr. to 60 F/hr.

in T/S 3.4.9.l.a on T/S page 3/4 4-25. This revision will make the subject specification consistent with the heatup rate currently specified by our heatup curve contained on T/S page 3/4 4-27, and will remain consistent with the above described revision to that curve.

Justification for Re uest and Si nificant Hazards Considerations We believe that operating with the revised heatup and cooldown curves and revised LTOP setpoint will not adversely impact public health and safety.

ATTACHMENT 1 TO AEP:NRC:08940 Page 2 10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

1) involve a significant increase in the probability or consequences of an accident previously analyzed,

2) create the possibility of a new or different kind of accident from an accident previously analyzed or evaluated, or

3) involve a significant reduction in a margin of safety.

Our evaluation of the proposed change with respect to these criteria is provided below.

Criterion 1 Heatup and cooldown limit curves are calculated using the most limiting value of the reference nil-ductility temperature (RTNDT) for the reactor vessel. Previously RT T was dependent on the phosphorous and copper content. With Re issuance of Regulatory Guide 1.99, Rev. 2, the phosphorous content no longer needs to be considered; instead, the nickel content must be considered when calculating the RT T. Our reactor vessel Capsule U analysis (WCAP-12483) submAPed in our letter AEP:NRC:0894M dated June 22, 1990, provided revised calculated RT values for both the base metal and the weld metal. Table 5-8 ok WCAP-12483 shows, for the various surveillance materials, a comparison of the transition temperature (ART D ) increases seen during the tests to the increases predic(ed using the methods of NRC Regulatory Guide 1.99, Rev. 2. This comparison shows that, for the plate B-4406-3 material (longitudinally), the transitiyy temperature increase resulting from irradiation to 1.88 x 10 n/cm is 9 F greater than that predicted by the guide, which includes a 2 sigma allowance for the shift prediction of 34 F. For the weld metal, the transition temperature increase was 37 0 F less than that predicted by the guide. The actual value of the change in the transition temperature is, for the base metal, slightly above the value calculated using the regulatory guide and slightly below for the weld.

ATTACHMENT 1 TO AEP:NRC:08940 Page 3 The LTOP system provides protection against exceeding the vessel ductility limits, as expressed by the pressure-temperature limits in 10 CFR 50, Appendix G, during cold shutdown, heatup and cooldown operations. As per Generic Letter 88-11, the implementation of Reg. Guide 1.99, Rev. 2 requires that the LTOP setpoints be re-evaluated. We have performed this re-evaluation and calculated the revised LTOP setpoint based on the plant-specific Westinghouse Owners Group methodology and the Unit 1 Capsule U analysis, WCAP-12483. The recalculated LTOP setpoint was determined to0 be 435 psig for the PORVs, with an enable temperature of 152 F.

Therefore, we have concluded that the above changes represent the application of a small refinement to a previously used calculation model or design method, and should not result in a significant increase in the probability or consequences of an accident previously analyzed.

We have made one editorial change in this submittal.

Specifically, on T/S page 3/4 4-25 we are requesting that the maximum heatup rate be changed to be consistent with the heatup rate specified on T/S page 3/4 4-27. This constitutes a purely administrative change to achieve consistency throughout the technical specifications, and therefore should not result in a significant increase in the probability or consequences of an accident previously analyzed.

Criterion 2 The proposed T/S changes concerning the heatup and cooldown curves and LTOP setpoint do not involve any physical modifications to the plant. The changes will involve changes to the plant's operating procedures; however, as noted in Criterion 1, the changes are based on the results of the reactor vessel Capsule U analysis, which followed the latest NRC guidance, Reg. Guide 1.99, Rev. 2.

Therefore, we have concluded that the above changes represent the application of a small refinement to a previously used calculation model or design method to conform to a revised NRC regulation and does not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 The heatup and cooldown curves and the revised LTOP setpoint were developed based on the results of the reactor vessel Capsule U analysis (WCAP-12483) which was submitted in our letter AEP:NRC:0894M dated June 22, 1990. The new heatup and cooldown curves were also developed using the criterion noted in Reg. Guide 1.99, Rev. 2.

ATTACHMENT 1 TO AEP:NRC:08940 Page 4 As per the requirements of Generic Letter 88-11, the LTOP setpoints have been re-evaluated and revised as part of implementation of Reg. Guide 1.99 Rev. 2. The LTOP setpoints were revised based on the plant-specific Westinghouse Owners Group methodology and Capsule U analysis.

As noted in Criterion 1, there was a slight change in the NDT temperature of the base metal and the weld metal. However, both materials exhibited an average charpy upper shelf egergy greater than 50 ft-lb at 32 EFPY at a fluence of 1.88 x 10 n/cm Capsule U received a fluence of 1.88 x 10 19 n/cm 2 . The calyylated cumu/ative fluence at the vessel inner surface is 1.41 x 10 n/cm at the end of 32 EFPY.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards considerations. The sixth of these examples refers to relief granted for changes resulting from the application of a small refinement of a previously used calculation model or design method. The changes proposed in this submittal are a result of our reactor vessel Capsule U analysis and the implementation of the requirements of Generic letter 88-11 and Regulatory Guide 1.99, Rev. 2 and, therefore, does not involve a significant reduction in the margin of safety. For this reason, we believe the example cited is relevant and conclude that the changes do not involve significant hazards considerations, which is consistent with previous NRC actions on applications of this type.