ML19323C716
| ML19323C716 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/25/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Andognini G BOSTON EDISON CO. |
| References | |
| NUDOCS 8005190057 | |
| Download: ML19323C716 (3) | |
Text
800519051h $10 2
UNITED STATES NUCLEAR REGULATORY COMMISSION a
3 y
WASHINGTON, D. C. 20555
$nOV/
APRll D 5' 1980 Docket No. 50-293 4
Mr. A, Carl Andognini Boston Edison Company M/C NUCLEAR 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Andognini:
SUBJECT:
EFFECT OF A DC POWER SUPPLY FAILURE ON ECCS PERFORMANCE It has generally been recognized that the loss of a direct current (DC) power supply could disable several eme. gen (y core cooling system components and thereby could result in a limiting single failure condition for some breaks. The enclosed letter report was subm?tted to the NRC staff by the General Electric Company to provide a definitive, generic, reference analysis of the effects of DC power supply failures or, ECCS conforumce calculations.
The NRC staff is reviewing the analysis which compares the peak cladding temperatures associated with various postulated DC power supply failure (ECCS availability) cases to the peak cladding temperatures for a HPCI (small break) failure and LPCI injection valve {large break) failure cases.
Since the study was based on plant design information which may have been incomplete or out-of-date, some uncertainty exists whether the worse ECCS system availability combinations have been identified for your operating BWRs. Accordingly, in order that we may have an adequate level of assurance that the systems combinations assumed in the generic analysis are conser-vative for Pilgrim Nuclear Power Station we request that you confinn the conclusion of the reference study regarding the minimum ECCS equipment availability with a DC power supply failure.
Include in your response a list of the ECCS equipment that would be available for large and small (1) recirculation loop discharge breaks, and (2) recirculation loop suction breaks. The listing of equipment available should take into account not only DC power supply failure, but also loss of equipment due to water spillage.
Please provide a schedule for your response within 30 days of the receipt of this letter.
Page 4 Mr. G. Carl Andognini This request for generic infonmation was approved by GA0 under clearance number B-180225(579018); this clearance expires October 31, 1980.
Sincerely,
, i Thorras A. Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Enclosure:
General Electric Conpany letter (R. E. Engel) to USNRC (P. S.
Check), "DC Power Source Failure for SWR /3 and 4," dated Noventer 1, 1978.
cc w/anciosure:
See rext page l
l l
l
Mr. G. Carl Andognini Boston Edison Company.
cc:
Mr. Paul J. McGuire Pilgrim Station Acting Manager Boston Edi ton Conpany RFD fl. Rocky Hill Road Plymouth, Massachusetts 02360 Henry Herrmann, Esquire Massachusetts Wildlife Federation 151 Tremont Street Boston, Massachusetts 02111 Plymouth Public Library North Street Plymouth, Massachusetts 02360 l
e O
l e
+ -Ma -
- N
hg of ?copqp3d G, E N E R A L 4 ELECTRIC wuCtema cuenov PROJECTS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE.. SAN JOSE, CAUFORNIA 95125 MC 682, (408) 925-1153 1
November 1,1978 REE- 054-78 MrN - 410-78 U.S. Nuclear Regulatcry Commission
'0ivision of Operating Reactors Washington, D.C. 20555 Attention:
P.S. Check, Chief Reactor Safety Branch Operational Technology
Dear Mr. Check:
SUBJECT:
DC Power Source Failure For BWR/3 and 4 A recent concern expressed by members of your staff is the effect of a direct current (DC) power source failure on the currently approved 10CFR50.46 conformance calculations for operating BWR/3's and 4's.
The Ceneral Electric Company has conducted a study of this concern and has documented the results in Attachment 1 to this letter. An additional concern expressed was the lack of a peak cladding temperature (PCT) versus break area curve in the small break region which could be applied to operating BWR/3's and 4's.
This concern is also covered in Attachment 1.
The study was performed with the 1977 approved model and ircut changes using bounding assumptions to provide a generic result applicable to all operating BWR/3's and 4's.
The results of the study show that there is an increase in PCT for small breaks; however, the PCT remains less than 1950 F.
For large breaks, the PCT was not af fected. Also, the maximum average planar linear heat generscion rate (MAPLHGR) is not affected for any plant.
If you '.iave any questions or comments, please contact R. T. Hill of my staf f on (408) 925-3255.
Sincerely, h5 R. E. Engel, Manager Operating Licenses I Safety and Licensing Operation Attachment cc: - F. D. Cof fman R. H. W. Woods t
hg c/ 70 acyp23y l1 h-} $ 0 /1 '.bs{ j i.
9I P]5 ne re.
h i
i j
ATTACHMENT 1 DC P0b'ER SOURCE FAILURE FOR Bk'R 3 AND 4 I
(
l
[
DUPLICATE DOCUMENT Entire document previously entered into system under:
ANO ]
b h
No. of ages:
.