ML18085A325
| ML18085A325 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/01/1980 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8012030630 | |
| Download: ML18085A325 (3) | |
Text
0 PSEG
- Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 December 1, 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. Frank J. Miraglia, Chief Licensing Branch 3 Division of Licensing Gentlemen:
CONTAINMENT BOUNDARY FRACTURE TOUGHNESS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311
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PSE&G hereby submits the following information in response to your letter dated November 17, 1980, requesting fracture toughness data for the Al06 Grade C piping used in the Steam Generator Feedwater System from the connection to the pro-cess pipe of f eedwater penetrations to the stop check isola-tion valves.
In accordance with original design requirements, fracture toughness data was not required for the subject portion of piping.
The requirements for fracture toughness of contain-ment boundary materials specified by GDC-51 were addressed as part of the plant design within the context of boundary definitions applicable at that time.
The inclusion of the subject piping within the containment boundary was not spec-ified by GDC-51 nor was it defined as such by governing de-sign codes specified in the FSAR.
Previously requested fracture toughness test data for previously requested con-tainment boundary materials has been submitted to the satis-faction of the NRC staff.
Pursuant to a telephone conversation between your Mr. J.
Halapatz and our Messrs. R. Kirk and G. Schnabel, we have investigated the spool pieces and the Al06 Grade C mater-ial.
Fracture toughness testing was not specified as a re-quirement in our design specification and therefore is not available in our documentation records.
We have inquired through our pipe fabricator whether such fracture toughness data exists in their records or in the records of the The Energy People 8 01~eao &3e; 95-0942
I Director of Nuclear Reactor Regulation 12/1/80 original mill suppliers using the same heats.
While we have not completed our research into this matter, initial indica-tions are that specific fracture toughness data, even if originally obtained, will not be available for this piping at this time due to time limits for record retention by the producing mills.
It is our position that the feedwater piping in question is not a safety concern.
Additional information related,to de-sign features of the subject piping is provided below.
We believe this provides equivalent confidence in lieu of frac-ture toughness test data.
- 1)
The piping in question, while officially designated Nuclear II, was upgraded to the more rigorous cri-teria of Nuclear I, including materials, fabrica-tion, inspection, and quality control.
This re-sulted in additional NDE on welds (MT plus RT) and pipe base material (UT).
This provides added con-fidence that initiating faults for pipe fracture at lower temperatures would be minimal.
- 2)
PSE&G imposed an additional conservative require-ment that one end of each pipe length be etched to show the material to be free from injurious lamina-tions, cracks, and similar defects.
- 3)
Stress levels were intentionally kept low, to 25%
of allowables of combined Sa + Sh (Sa = Stress Al-lowable, Sh = Stress at the hot temperature) to ac-comodate the high energy line break analysis.
This resulted in our specifying schedule 120 piping in-stead of schedule 80 (required for normal service conditions).
- 4)
The system, as designed, does not violate contain-ment integrity requirements because the path of leakage would necessitate a break inside as well as a break outside the containment to provide the path for containment atmosphere to be released externally.
Taken in combination, the high quality level of fabrication which reduces the incidence of crack propagating flaws, plus reduced design stress levels which reduce the tendency for crack propagation, it is our view that the piping affords the required resistance to propagating fracture within the intent of GDC-51.
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Director of Nuclear Reactor Regulation 12/1/80 As indicated above, we have not yet completed our search for fracture toughness test data on the heats in question and will advise you as soon as information is available.
Based on the above, we believe that the plant design is acceptable for full power operation.
Should you have any questions in this regard, do not hesi-tate to contact us.
EP3 1/3 m;r/~Y, 195_:s,
"'F cf l'tr, ;(jl(r R. L. Mittl General Manager -
Licensing and Environment