ML20114B184
| ML20114B184 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/11/1985 |
| From: | Edelman M CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| REF-PT21-85-116-000 (RDC-123(85)), PT21-85-116, PT21-85-116-000, NUDOCS 8501280706 | |
| Download: ML20114B184 (2) | |
Text
TAlb Ik!E bb[Y Eb: ii O E. b (( C I O ! O b b b [n ! 5 n I! iib bOntbObY P.O. Box 5000 - CLEVELAND. OHIO 44101 - TELEPHONE (216) 622-9800 - ILLUMINATING BLDO. - 55 PUBLIC SQUARE Serving The Best Location in the Nation MURRAY R. EDELMAN
"" * #I VICE PRESIDENT NUCLEAR Mr. James G. Keppler Regional Administrator, Region III Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 RE: Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Overpressurization of Isolated Volumes in Drywell Post-LOCA
[RDC 123(85)]
Dear Mr. Keppler:
This letter serves as the interim report pursuant to 10CFR50.55(e) on the potential design deficiencies associated with Drywell overpressurization which could affect certain components in a post-LOCA (Loss of Coolant Accident) environment.
Mr. J. McCormick-Barger of your office was notified on December 13, 1984, by Mr. T. A. Boss of The Cleveland Electric Illuminating Company (CEI) that this problem was being evaluated per our Deviation Analysis Report 218.
This report contalas a description of the deficiency, actions planned to complete our i. valuation, and the planned date for our next report.
Description of Deficiency The drywell area in the reactor building contains four isolated piping volumes which are unprotected by pressure relief valves. The B21 system and the dryvell personnel airlock door seals and associated piping are primarily affected.
It has been determined that the original design for the piping and components related to these four volumes did not consider worst case pressures due to post-LOCA temperature excursions. This resulted in underestimation of the upper design ptessure limit occurring in a post-LOCA environment.
Completion of Evaluation Gilbert / Commonwealth, Inc., our Architect / Engineer, performed an analysis of the design with the correct temperature excursions and resultant pressures.
The analysis indicates that the piping and components in the drywell area, except for piping associated with the drywell airloc.k will not be overstressed by the pressures resulting from post-LOCA conditions. The piping and inner 8501280706 050111 PDR ADOCK 05000440 g
PDR h it ng5 }I 10 act
.Mr.uJames G. K:ppist Jcnuary 11, 1985 door inflatable seals on the drywell personnel airlock require further testing and evaluation to determine if they can be requalified to the anticipated post-LOCA pressures.
Safety Relief Valves and Main Steam Isolation Valves also are being evaluated to ensure that they will operate at the higher post-LOCA drywell pressures.
Newport News Inc./W. J. Wooley Co. and General Electric are presently testing and evaluating their respective components to determine the effect of the elevated drywell pressure.
We anticipate the next report on the subject to be submitted by April 12, 1985.
Please call if there are any additional questions.
Sincerely, 4
Murray R. Edelma Vice President Nuclear Group MRE:pab cc:
Mr. J. A. Grobe USNRC, Site Office - SBB50 1
Mr. D. E. Keating USNRC, Site Office - SBB50 Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555 U.S. Nuclear Regulatory Commission c/o Document Management Branch 1
d Washington, D.C.
20555 Records Center, SEE-IN Institute of Nuclear Power Operations 1100 circle 75 Parkway, Suite 1500 Atlanta, Georgia 30339 e
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