ML20215N415
| ML20215N415 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/31/1986 |
| From: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20215N416 | List: |
| References | |
| TAC-63527, NUDOCS 8611050381 | |
| Download: ML20215N415 (6) | |
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thrt D W4hers Vce Presdent October 31, 1986 Trojan Nuclear Plant Docket 50-344 License NPF-1 l
Director of Nuclear Reactor Regulation ATTN:
Mr. Steven A. Varga Director, PWR-A Project Directorate No. 3 U.S. Nuclear Regulatory Commission Washington DC 20555
Dear Sir:
License Change Application 147 Attached are three signed originals and 40 conformed copies of License Change This Application (LCA) 147 requesting amendment of Operating License NPF-1.
LCA addresses the reactor vessel material irradiation surveillance schedule and new 10 CFR 50, Appendix G, pressure-temperature limits. An LCA fee of
$150 is attached in accordance with 10 CFR 170.
Also attached is one signed copy of a Certificate of Service for LCA 147 to the chief executive of the county in which the facility is located and the~
Director of the State of Oregon, Department of Energy.
Sincerely, Bart
. Withers Vice President Nuclear c:
Mr. Lynn Frank, Director State of Oregon Department of Energy Mr. Michael'J. Sykes chairman of County Commissioners 1
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DR ADOCK 0500 121 S W Sa mon Sveet. Pocand O'egon 97204
f UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
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I PORTLAND CENERAL ELECTRIC COMPANY,
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Docket 50-344 THE CITY OF EUGENE, OREGON, AND
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Operating License NPF-1 PACIFIC POWER & LIGHT COMPANY
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(TROJAN NUCLEAR PLANT)
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CERTIFICATE OF SERVICE I hereby certify that copies of License Change Application 147 to the Operating License for Trojan Nuclear Plant, dated October 31, 1986, have been served on the following by hand delivery or by deposit in the United States mail, first class, this 31st day of October 1986:
Mr. Lynn Frank Director-State of Oregon Department of Energy 625 Marion St NE Salem OR 97310 Mr. Michael J. Sykes Chairman of County Commissioners Columbia County Courthouse St. Helenn OR 97051 E. L. Kershul,' Acting Manager Nuclear Regulation Branch Nuclear Safety & Regulation Subscribed and sworn to before me this 31st day of October 1986.
9MyCommissionE Notary Public of Orep6n s'
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PORTLAND GENERAL ELECTRIC COMPANY EUGENE WATER & ELECTRIC BOARD AND PACIFIC POWER & LIGHT COMPANY Operating License NPF-1 Docket 50-344 License Change Application 147 This License Change Application requests modifications of the Technical Specification contained in Appendix A to Operating License NPF-1.
In order to maintain compliance with NRC regulations, changes are proposed to revise the reactor vessel material irradiation surveillance schedule and address new 10 CFR 50, Appendix C, pressure-temperature limits.
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PORTLAND GENERAL ELECTRIC COMPANY By h y=
% t D. Withers Vice President Nuclear Subscribed and sworn to before me this 31st day of October 1986.
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U Notary Public of [ gon My Commission Expires:
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LCA 147-Page 1 of 3 LICENSE CHANGE APPLICATION 147 The proposed replacement pages to Appendix A of Facility Operating License NPF-1 are provided as Attachment 1.
A description of the changes to the existing Technical Specifications are as follow:
Pages 3/4 4-25 and 3/4 4-26.
Changes to Secti'on IV.A.2 of 10 CFR 50,
. Appendix G, require that When the core is not critical, and pressure exceeds 20 percent of the preservice system hydrostatic test pressure (620 psi), that the temperature of the closure flange regions must exceed the reference temperature of the material in those regions by at least 120*F for normal operation. The heatup and cooldown rate pressure-temperature limit curves of Figures 3.4-2 and 3.4-3 have been revised to reflect this requirement.as well as the results of the analysis of radia-tion specimen Capsule X.
For the purpose of the change, the reference temperature for the head flange was determined to be 20*F from Technical Specification Table B 3/4 4-1.
Possible instrument errors of 10*F and 60 psig have been included in the curves.
Page 3/4 4-27.
The Capsule withdrawal schedule in 10 CFR 50, Appendix H was changed to require compliance with ASTM E 185-82 for Capsules with-drawn after July 26, 1983. The withdrawal schedule in ASTM E 185-82 is based on effective full power years (EFPY) rather than calendar years as is required by the current Technical Specification. WCAP-10861,
" Analysis of Capsule X From Portland General Electric Company Trojan Reactor Vessel Radiation Surveillance Program", provided a recommended removal schedule for the remaining capsules, which is in compliance with ASTM E 185-82.
Page B 3/4 4-5.
The heatup and cooldown curves shown in Figures 3.4-2 and 3.4-3 are applicable for the first 10 EFPY rather than 12 EFPY as stated in Paragraphs 3 and 4 on Page B 3/4 4-5.
A paragraph was added to discuss the bases for the " knee" in Figures 3.4-2 and 3.4-3.
Page B 3/4 4-9.
In accordance with the change to 10 CFR 50, Appendix H, ASTM ~E 185-82 is now the accepted version'of this standard. The refer-ence to ASTM E 185-73 as the version for removing and evaluating the Capsule specimens is therefore being changed to ASTM E 185-82.
Addi-tionally, in Paragraph 3, Table 4.4-3 is incorrectly referenced and is being changed to read Table 4.4-5.
In the final. paragraph on this page the applicability of the heatup and cooldown curves is being changed from the first 15 EFPY to the first 10 EFPY to reflect the Capsule I analysis.
REASON FOR CHANGE On Friday, May 27, 1983, the NRC published in the Federal Register a Final Rule amending the' fracture toughness requirements for light water nuclear power reactors and the requirements for reactor vessel material surveillance programs.
These rule changes affected 10 CFR 50, Sections 50.12, 50.55(a), 50.60 and Appendices G and H.
As a result of
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l LCA 147 Page 2 of 3 the change to 10 CFR 50, Appendix H and the analysis of radiation speci-men Capsule X. Technical Specification Table 4.4-5 should be revised to reflect the schedule changes for future specimen Capsule removal. -The.
Bases for Technical Specification 4.4.9.1 are being revised in accordance
-with the changes made to Appendices G and H and.to correct typographical errors. Finally, Figures 3.4-2 and 3.4-3 and the Technical Specification Bases for these figures are being revised to reflect new 10 CFR 50, Appendix G, pressure-temperature limits and the results from'the analysis of radiation specimen Capsule X.
SIGNIFICANT HAZARDS DETERMINATION In accordance with the requirements of 10 CFR 50.92, this application is judged to involve no significant hazards based upon the following infor-mation:
1.
Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
This change would have no effect upon the probability or con-'
sequences of.an accident previously evaluated. Operating limits are being adjusted to incorporate empirical data obtained through analysis of irradiation specimen Capsule X.
The Capsule X data demonstrates the reactor vessel is less sensitive to radiation affects than predicted by Regulatory Guide 1.99, Revision 1.
Changing these operating limits does not affect the probability or consequences of previously evaluated accidents. The Plant will operate within these new limits ensuring that the probability and consequences of.previously evaluated accidents are unchanged.
2.
Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
The possibility of a new kind of. accident is not created since the operating limits are merely being updated according to changes in federal regulations (10 CFR 50, Appendices G and H) and the Capsule X analysis report. No physical changes are being made to the Plant or operating requirements.
The revised heatup and cooldown curves and schedule for radiation specimen removal reflect empirical data obtained through analysis of radiation surveillance specimen Capsule X.
3.
Does the proposed amendment involve a significant reduction in a margin of safety?
The equipment of concern is the reactor pressure vessel. The margin of safety for the reactor pressure vessel will not be affected. The predicted affects of irradiatien on the vessel
u LCA 147 Page 3 of 3 have been determined to be more conservative than the. actual effects as determined by the analysis of surveillance Capsule X.
The change in the operating limits is based upon actual data and ensures that original assumptions, adequate conservatisms, and the margin of safety of the reactor pressure vessel remain as initially designed.
In the April'6,.1983 Federal Register, the NRC published a list of examples of amendments that are not likely to involve significant hazards considerations. Example Number 7 of that list applies to these proposed changes and states that a change similar to the following would not likely involve a significant hazards consideration:
"A change to make a license conform to changes in the regulations where the license change results in very minor changes to facility operations clearly in keeping with the regulations."
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