ML20134H249
| ML20134H249 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/26/1985 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8508280335 | |
| Download: ML20134H249 (3) | |
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0 O PS G Company PutAc Service Electnc and Gas 80 Park Plaza, Newar k, NJ 07101/ 201430 8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation August 26, 1985 Director of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission 7920 Norfolk Avenue Bethe,da, MD 20814 Attention:
Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlemen:
ASME CODE CASE N-411 AND LATER CODE EFFECTIVE DATES HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Pursuant to the request for additional information dated July 2, 1985, letter from W.
R.
Butler, NRC, to R.
L. Mittl, PSE&G, the following information is being provided to support Public Service Electric and Gas Company's request for authorization to use ASME Code Case N-411 and selected portions of the 1981 Winter Addenda and 1983 Edition of Section III of the ASME Boiler and Pressure Vessel Coder When using Code Case N-411, the following provisions will be followed:
1.
Code Case N-411 when invoked, will only be used in conjunction with the response spectrum method of piping system analysis.
2.
Code Case N-411 when invoked, will only be used for reconciliation work.
3.
Maximum calculated displacements associated with Code Case N-411 reanalysis will be shown on the stress isometric per existing project practices.
Adequate clearances are. confirmed during stress walkdowns using the stress isometric as the source document.
OD002hh p
PDR s ()
The Energy People A
I 96 4112 (VM 4 84
Director of Nuclear Reactor Regulation 2
8/26/85 1
4.
Code Case N-411 when invoked, will not be used in conjunction with Reg. Guide 1.61.
5.
The use of Code Case N-411 will be documented in a future FSAR amendment.
When using NC 3652, 3653, 3654, and 3655 of the 1980 Edition Winter 1981 Addenda or ND 3652, 3653, 3654, and 3655 of the 1983 Edition, Summer 1984 Addenda the following provisions will be followed:
1.
When the later code is invoked, systems that are affected will be listed in the FSAR.
2.
a.
The entire anchor to anchor calculation will be analyzed to the later code requirements.
b.
Appendix I of the later code will be used to obtain related engineering data.
c.
Stress indices and stress intensification factors will be obtained from the later code.
d.
Criteria for postulating break locations shall be based on the current code equation 9 and 10 stress values pursuant to MEB 3-1, Rev.-l.
3.
The FSAR and applicable project specifica-tions are being revised to note the use of the specific later code edition / addenda and related requirements.
When using subsection NF-3000 of the 1983 Edition of ASME Section III, the following provisions will be followed:
1.
When the later code is invoked, systems that are affected will be listed in the FSAR.
2.
Since all of Section NF-3000 is being invoked, we find no other related require-ments that have major differences that relate to the use of increased design allowables.
I
i e,
Director of Nuclear.
Reactor Regulation 3
8/26/85 3.
Revisions to the FSAR and applicable project specifications to list the specific later edition / addenda are underway.
Should you have any questions in this regard, please contact i
us.
Very truly yours, lb /
C D.
H. Wagner USNRC Licensing Project Manager A. R. Blough USNRC Senior Resident Inspector MG 27 1/3
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