ML19282B157
| ML19282B157 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/05/1979 |
| From: | Whitmer C GEORGIA POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19282B158 | List: |
| References | |
| TAC-10710, NUDOCS 7903090170 | |
| Download: ML19282B157 (9) | |
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Georgia Power Chas. E Whdmri v4e n neet Enybem o March 5, 1979 tr a"ourta" e
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,1! v" Dir,cetor of Nuclear Reactor Regulation U. S. Nuclear Regulatory C)mmission Washington, D. C.
20555 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 10CFR50, APPENDIX J Centlemen:
Pursuant to our letter. dated November 16, 1977, Georgia Power Company hereby submits a reevaluation of the Plant Hatch Unit 1 containment leak rate test program. The program has been prepared to conform with the most recent positions on leakage rate testing which were established with the NRC during the Plant Hatch Unit 2 review. Also included are proposed changes to the Technical Specifications which reflect the provisions of the updated program.
Included as part of the reevaluation are proposed plant modifications needed for several penetrations to facilitate a method of leak rate testing that is in compliance with the requirements of 10CFR50, Appendix J.
Your approval of these modifications is requested prior to initiation of any activities concerning equipment procurement and engineering design changes.
Implementation of the full containment leak test program will be dependent upon the successful installation of J ase proposed nodifications whose completion is limited by component availability and plant scheduling.
As indicated in our letter dated August 3, 1978, concerning the in-service inspection program, justification has been provided in this con-tainment leak rate test program for the reverse Icakage testing of several types of containment isolation valves.
This information also serves as a basis for our requests for ASME Section XI code relief as stipulated in the in-service inspection program submitted that date. Also, the plant modifi-cations needed for proper leak rate testing, which were referred to in our August 3,1978, letter, have been identified in the containment leak rate test program as discussed above.
With regard to the proposed changes to the Technical Specifications, you will find that Tables 3.7-2, 3.7-3, and 3.7-4 have been deleted.
Each table contained a list of primary containment penetrations for double 0-ring seals, 790-3090170 O
P-
e Geo@iapower A Director of Nuclear Reactor Regulation U.' S. Nuclear Regulatory Commission March 5, 1979 Page 2 expansion bellows, and isolation valves respectively. With respect to the updated program, these tables are inaccurate and incomplete.
Rather than include all of the penetrations in these tables, we determined that it would be more prudent to incorporate statements in the Surveillance Re-quirements which require the program for Type A, B, and C tests to be in
'n'ecordance with the requirements of 10CFR50, Appendix J.
The intent, theref o.re, is more comprehensive than trying to list each penetration in the Tech'nical Specifications similar to the listing in the containment leak rat'e test program. Additionally, we believe such a detailed amount of
.information is not proper for inclusion in the Technical Specifications.
This approach is consistent with the Unit 2 Standard Technical Specifica-tions concerning this subject.
The Plant Review Board and Safety Review Board have reviewed and approved the proposed changes to the Technical Specifications and have determined that they do not constitute an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety are not increased, nor is the possibility of an accident or malfunction of a different type than those previously analyzed created since the proposed changes to the Technical Specifications and proposed plant modifications are based on compliance with the require-ments of 10CFR50, Appendix J.
Therefore, margins of plant safety are not
' reduced.
Yours very truly, A
W:/ '
y h
L y
y Ches. F. Whitmer WEB /TMM/mb Enclosures
'l. Proposed Determination of Amendment Class 2.
Proposed Changes to Technical Specifications
'3.
Proposed Containment Leak Rate Test Program
' Swo to and subscribe,d before me this 5th day of March, 1979.
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My Commission Expires Sept. 29,19dl xc:
Mr. Ruble A. Thomas George F. Troubridge, Esquire
ATTAC1IMENT 1 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. IIATCil NUCLEAR PLANT UNIT 1 PROPOSED CIIANGES TO TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 170.12 (c), Georgia Power Company has evaluated the attached proposed amendment to Operating License DPR-57and have determined that:
a)
The proposed ~ amendment does not require the evaluation of a new Safety Analysis Report or rewrite of the facility license;
~
b)
The proposed amendment does not contain several complex issues, does not involve ACRS review, or does not require an environmental
. impact statement; c)
The proposed amendment does not involve a complex issue, an environ-mental issue or more than one safety issue; d)
The proposed amendment does involve a single issue; namely, the up-dating of the containment leak rate test program to meet current requirements; e)
The proposed amendment is therefore a Class III amendment.
e
ATTACllMENT 2 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. IIATCll NUCLEAR PLANT UNIT 1 PROPOSED CIIANGES TO TECIINICAL SPECIFICATIONS The proposed changes to the Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:
Remove Page Insert Page 3.7-5 3.7-5 3.7-6 3.7-6 3.7-6a 3.7-6a 3.7-21 thru 3.7-27 3.7-21 thru 3.7-27 o
LIMITI!?G CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.2.b(2)
(c) The acceptance criteria for subsequent peak pressure tests shall require the new Lam not to exceed La*
(d) The allowable operational leak rete, Lao, which shall be met prior to resumption of power operation following a test (either as measured or following repairs and retest) shall not exceed 0.75 L -
a c.
Corrective Action for Type A Tests If leak repairs are necessary to meet tho allowable operational leak rate, the integrated leak rate test need not be repeated provided local leakage measure-ments are conducted and the Icak rate differences prior to and after repairs, when corrected to the test pressure and deducted from the integrated leak rate measurements, yield a leak rate value not in excess of the allowable operational leak rate, d.
Frequency for Type A Tests After the initial preoperational leak rate test, two integrated leak rate tests shall be per-formed at approximately equal intervals between the major shutdowns for inservice inspection conducted at ten-year intervals.
In addition, an integrated leak rate test shall be performed at the end of the ten-year interval, which may coincide with the inservice inspection shutdown
- period, e.
Type B Test - Leak Tests of Pene-trations with Seals and Bellows Type B tests shall be performed under the program established in Appendix J of 10CFR Part 50 (Reference 1),
3.7-5 s
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4 2 7. A. 2. e. Type B Tes t - Leak Tes ts of Pene-trations with Seals and Bellows (Continued)
(1) Primary containment components which seal or penetrate the pressure containing boundary of the containment shall be tested at a pressure not less than P.
These components a
shall be tested at each major refueling shutdown or at intervals not to exceed two years.
(2) (a)
The personnel air lock shall be tested at in-tervals not to exceed six months at P by Pres-a surizing the compartment between the two air lock doors.
During intervals of door use when containment integrity is required, the door seals shall be tested at 10 psig after each opening.
(b)
Personnel air lock leakage shall not exceed 0.05 L -
a f.
Type C Tests-Local Leak Tests of Containment Isolation Valves Type C tests shall be performed under the program established in Appendix J of 10CFR Part 50 (Reference 1).
Containment isolation valves (except for main steam line iso-lation valves) shall be tested at a pressure not less than Pa-Type C tests shall be performed at each major refueling shutdown or at intervals not to exceed two years.
3.7-6
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS g.
Acceptance Criteria for Type B and Type C Tests The combined leakage rate of components subject to Type B and C tests shall be determined under the program established in Appendix J of 10CFR Part 50 (Reference 1) and shall net axceed 0.6 L.
3 h.
Main Steam Line Isolation Valves The main steam line isols<
tion valves shall be tested at a pressure of 1/2 Pa for leakage at least once per operating cycle.
If a total leak rate of 11.5 scf per hour for any one main steam line isolation valve is exceeded, repairs and retest shall be performed to correct this condition.
3.7-6a
LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 4.7.E.
References 1.
" Reactor Containment Leakage Testing for Water Cooled Power Reactors", Appendix J to 10CFR50.54 (o) February 14, 1973 as corrected and amended through April 19, 1976.
2.
" Testing Criteria for Integrated Leak Rate Testing of Primary Containment Structures for Nuclear Power Plants", Topical Report BN-TOP-1, Revision 1, Bechtel Corp.
Issued November 1, 1972.
e 3.7-15
I These pages have been left blank.
l 1
3.7-21 thru 3.7-27