ML20129D985

Rev 2 to License Change Application 99 for License NPF-1, Updating Tech Spec Figures 3.4-2 & 3.4-3 Re Heatup & Cooldown Curves

ML20129D985
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/12/1985
From:	Withers B PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20129D962	List: ML20129D958 ML20129D985 ML20129D993
References
TAC-54218, TAC-56534, NUDOCS 8507160704
Download: ML20129D985 (3)

Text

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PORTLAND GENERAL ELECTRIC COMPANY l EUGENE WATER (. ELECTRIC BOARD AND PACIFIC POWER & LIGHT COMPANY TROJAN NUCLEAR PLANT l

Operating License NPF-1 Docket 50-344 License Change Application 99, Rev. 2 This Licenso change Application updates the heatup and cooldowr. curves (Figures 3.4-2 and 3.4-3) of the Technical Specifications, Apr endix A to Operating License NPF-1 for the Trojan Nuclear Plant, in compliance with

, Appendices G and H of 10 CFR 50.

PORTLAND CENERAL ELECTRIC COMPANY l l

l By '

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'Bart D. Withers Vice President Nuclear Subscribed and sworn to before me this 12th day of July 1985.

K uwde_ sb NotaryPublicotgregon i

My Commission Expires:

%, ,2.,j f.

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0507160704 050712 PDR ADOCK 0D000344 p PDH l

LCA 99, Rev. 2 Page 1 of 2 LICENSE CHANGE APPLICATION 99. REVISION 2 The following changes to Appendix A of Facility Operating License NPF-1 are requested (proposed replacement pages are provided as Attachment 1).

Pages 3/4 4-25 and 3/4 4-26. Changes to Section 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 referance 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. 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.

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-9469, " Analysis of Capsule U From Portland General Electric Company Trojan Reactor Vessel Radiation Survelliance Program", provided a recome. ended removal schedule for the remaining capsules based on ASTM E 185-79 requirements. The withdrawal schedule requirements of ASTM E 185-82 did not change from the 1979 version of this code, and therefore the withdrawal schedule provided 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 15 EFPY rather than 12 EFPY as stated in Paragraphs 3 and 4 on this page. A paragraph was added to discuss the bases for the change to 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 reference to ASTM E 185-73 as the version for removing and evaluating the capsule specimens is therefore being changed to ASTM E 185-82. Additionally, in Paragraph 3. Table 4.4-3 is incorrectly referenced and is being changed to read Table 4.4-5.

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 changen affected 10 CFR 50 Soc-tions 50.12, 50.55 (a), 50.60 and Appendices C and H. As a result of the change to 10 CFR 50, Appendix H. Technical Specification Table 4.4-5 must be revised to reflect the schedule changes for specimen capsulo removal, l

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LCA 99, Rev. 2 Page 2 of 2 The Bases for Technical Specification 4.4.9.1 are being revised in accordance with the changes made to Appendices C 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 C, pressure-temperature limits.

SAFETY / ENVIRONMENTAL EVALUATION Table 4.4-5 in Technical Specification 4.4.9.1 will be changed to reflect the new capsule withdrawal schedulo required by 10 CFR 50, Appendix H.

This schedule corresponds with the schedulo recommended for the Trojan Nuclear Plant in WCAP-9469, which was prepared in accordance with ASTM E 185-79. The schedule requirements in ASTM E 185-82, which is the referenced version of the standard in 10 CFR 50 Appendix H, are the same as those in ASTM E 185-79, and thorofore, the proposed schedule is in compliance with 10 CFR 50, Appendix H. This chango is administrative in nature, and does not result in an unreviewed safety question. The changes to the Technical Specification Bases are administrativo in nature, and therefore, do not decrease the margin of safety as defined therein.

Figures 3.4-2 and 3.4-3 are being revised to comply with the now require-ments in 10 CFR 50 Appendix G. For the temperature range of approxi-mately 100'F to 150*F the pressure limitation will be lower than previously required. This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in the FSAR nor does it create the possibility of an accident or malfunction of equipment of a different type than any previously evaluated. The chango does not reduce the margin of safety as defined in the Bases of the Technical Specifications.

SIGNIFICANT HAZARDS DETERMINATION This license change results in very minor changos to facility operations and has boon initiated in order to comply with NRC regulations. The change is administrativo in nature and does not involvo a significant hazards consideration. In fact, the schedulo for irradiated specimen withdrawal has boon moved up such that any potentially damaging radiation effects to the reactor vosso1 will be discovered at an earlier date than was previously scheduled, thus increasing the margin of safety.

BASIS FOR DETRRMINATION OF AMENDMENT _ CLASS It was determined that this change, as originally submitted on January 27, 1984, resulted in a Class II Amendmont as it was then defined in 10 CFR 170.22. The Class II Amendment foe of $1,200 was submitted at that time.