ML20003H945
| ML20003H945 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/01/1981 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| JPN-81-33, NUDOCS 8105080221 | |
| Download: ML20003H945 (2) | |
Text
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4 POWER AUTHORITY OF THE STATE OF NEW YORK 10 CoLUMaus CincLE NEW YoRx. N. Y.1o019 (212) 397 6200 G E0",G,E,,T;,e,a,n avc =..,
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RICH ARD W. Puf NN Oe ar i. uictoNr' May 1, 1981
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...s Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D. C.
20555 N
Attention:
Mr. Thomas A.
Ippolito, Chief
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{ g,-['Jd U g Operating Reactors Branch No. 2 Division of Licensing
&,i,tr{( 0 7 198 F ' f>
Subject:
James A. FitzPatrick Nuclear Power Plant awswo8 Docket No. 50-333
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Pipe Stress & Pipe Support Analysis co"
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Effect of Assumed Support Flexibility M r# W
Reference:
PASNY letter, P.J. Early to T.A.
Ippolito (JPN-79-47), dated August 10, 1979.
Dear Sir:
This letter provides the information to fulfill the commitment made in the referenced letter regarding the effect of pipe support stiffness on the pipe stress re-analysis performed for the Fitz-Patrick Plant.
Of the 104 Seismic Category I Pipe Stras problems analyzed for FitzPatrick, 37 problems were re-analyzed using assumed support stiff-nesses. All results were satisfactory and can be summarized as follows:
all are within allowable valuas.
Pipe Stress (37) all are within allowable values, Pipe Supports (248) loads same or decreased on all i
NOTE:
93 Supports components.
136 Supports one or more components of load increased from 0-20% or, insig-nificant load change of 100 lbs.
or less.
/f00l 19 Supports one or more compon=nts of load g
increased by more chan 20%.
/o 8105080 M
y
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Thomas A.
Ippolito Nuclear Regulatory Commission Page 2 Eacn of the 19 supports in which one or more components of load increased by more tnan 20% was reviewed.
It was concluded tnat all supports were within allowable stress levels.
Upon completion of these 37 proolems, a review was conducted to determine the actual effeat of assuming more flexible supports.
The conclusion was that the effect generally was negligible.
Tne only situation where a significant effect might occur is wnere a mode with a significant participation factor occurred at a frequency slightly higher than a peak in the applicable Amplified Response Spectra (ARS).
A review of the remaining 67 problems showed 13 in which the above criteria for potential significant jncrease existed.
Therefore, tnese 13 proolems were re-analyzed to determine the effect of changed support flexibility on seismically induced stresses.
The results in-cluding new seismic stresses are summarized as follows:
all are within allowable values.
Pipe Stress (13) all are within allowable values.
Pipe Support (156)
NOTE:
60 Supports -
loads same or decreased in all components.
61 supports -
one or more components of load increased from 0-20% or, insig-nificant load change of 100 lbs.
or less.
I 35 Supports -
one or more components of load l
increased by more than 20%.
Each of the 35 supports in which one or more components of load increased by more than 20% was revie.<ed.
The conclusion of this re-view was that all supports were within allowable stress levels.
The final conclusion of this investigation is that no design changes would be required due to the effect of changed flexibilities.
Very truly yours,
(
.Eayne[
J enior Vice President-Nuclear Generation
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