ML20033B381
| ML20033B381 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/23/1981 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8112010341 | |
| Download: ML20033B381 (4) | |
Text
.
M.nlling Addrses -
Alabama Power Company 600 North 18th Street Post Othce Box 2641 Birmingham. Alabama 35291
=
Telephone 205 783-6081 F. L. Claytori, Jr.
Ef,'n"! %*"'
Alabama Power o
tMsouthern ekxtrc systern November 23, 1981 Docket Nos. 50-348 A
A 50-354 g
g Director, Nuclear Reactor Regulation b
U. S. Nuclear Regulatory Commission p NOV3 0 ;ggg Washington, D. C.
20555 u.a.
-p Attention: Mr. S. A. Varga
}
Gentlemen-JOSEFH M. FARLEY NUCLEAR PLANT - UNITS 1 AND 2 ECCS Analysis /Fo Technical Specification Change 4
Alabama Power Company's Nuclear Operations Review Board (NORB) has 4
reviewed the proposed changes to the Fg Technical Specification Section 3.2.2, previously submitted by Alabama Power Company letter. dated November 16, 1981 and has determined that the change does not involve an unreviewed safety question. The NORB revised the last paragraph of the Safety Evaluation, which was submitted in Alabama Power Company's November 16, 1981 letter, to state that even with an increase in core average temperature of 1.6 F, the core average temperature will remain below the design basis value for non-LOCA transients. The revised Safety Evaluation page is enclosed for your review.
The proposed changes to Technical Specification 3.2.2 have been made to allow plugging the first row tubes in the Unit 1 steam generators.
The changes will allow the option of plugging the Unit 2 steam generator tubes if it becomes necessary in the future. The Technical Specification change is needed by December 7, 1981 to allow tube plugging on Unit 1 during the current refueling outage.
If there are any questions, please advise.
Yours very truly, dWfC/b (
F. L. Clayto, Jr.
FLCJr/RLG:Jc Enclosure cc: See Page 2 he \\
8112010341 811123-PDR ADOCK 05000348 p
Director, Nuclear Reactor Regulation November 23, 1981 U. S. Nuclear Regulatory Commission Page 2 Attention: Mr. S. A. Varga cc: Mr. R. A. Thomas (w/ enclosure)
Mr. G. F. Trowbridge (w/ enclosure)
Mr. J. P. O'Reilly (w/ enclosure)
Mr. E. A. Reeves (w/ enclosure)
Mr. W. H. Bradford (w/ enclosure)
SWORN TO AND SUBSCRIBED BEFORE ME
.2 3 M DAY OF [ M M,
THIS
(
1981.
V N6tary Pub' ic My Commission Expires:
/0-l?-f5 2
t 4
f l
1
~
1 ENCLOSURE Revised Page 3 to Safety Evaluation i
1 TNe*'*W"'W--wm--p m wm.,
,p g,,.
s Safety Evaluation Page 3 age model. Thus, a change to Technical Specification 3.2.2 will be required to revise the Fg Limit from 2.32 to 2.31.
Farley Unit 1 is currently under a forced outage due to a turbine generator repair and will be ready for Cycle 4 criticality by mid-February 1982. Farley Unit 2 is approaching one fourth of the first cycle fuel burnup level. The slight reduction in the Fg limit from 2.32 to 2.31 will have no operational impact on Unit 1 Cycle 4 or Unit 2 Cycle 1.which were initially analyzed using a heat flux hot channel factor, Fg, of 2.32.
Both Farley Units 1 and 2 are operated well below the Fg limit. The expected maximum total peaking factor F during nonnal operations of the reactor including ' load following maneuvers for beginning, middle, and end of cycle conditions are below the 2.31 limit.
The surveillance requirements necessary to assure that this limit is not exceeded are provided in plant Technical Specification 4.2.2.
Plugging of the steam generator tubes to the 5% level is calculated to result in an increase in the core inlet temperature of 1.6* F with an accompanying increase in the core average temperature of approximately the same amount.
Even with this increase, the core average temperature will remain below the design basis value for non-LOCA transients, which is a core average temperature of 577.2* F.
The effect on core flow due to the steam generator tube plugging is insignificant. Therefore, plugging the steam generators tubes to a level of 5% does not invalidate the current non-LOCA transient analyses.
==
Conclusion:==
The proposed change to Technical Specification 3.2.2 does not involve an unre-viewed safety question as defined by 10CFR50.59.