ML20116P264
| ML20116P264 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/16/1992 |
| From: | Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, NUDOCS 9211250009 | |
| Download: ML20116P264 (13) | |
Text
't GL 92-01 i
PillLADELPillA ELECI'RIC COMPANY NUCLEAR GROUP llEADQUARTERS 95545 CHESTERBROOK BLVD.
WAYNE, PA 19087-5691 (215) MO#XX)
November 16, 1992 Docket Nos. 50-352 50-353 NUCU.d SIMICLS DUMMLNT License Nos. NPF-39 NPF-85 U.S.
Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555 SUDJFCT' Limerick Generating Station, Units 1 and 2 Supplemental Response to Generic Letter 92-01, Revision 1,
" Reactor Vessel Structural Integrity, 50.54(f)"
REFERENCE:
1)
Letter from G. J. Beck (PECo) to U.
S.
Nuclear Regulatory Commission, dated July 10, 1992
Dear Sir:
Attached is our supplemental response to the subject Generic Letter 92-01, dated March 6, 1992.
Our original response was provided in the Reference 1 1 otter.
Generic Letter 92-01 concerns licensee's compliance with requirements and commitments regarding reactor vessel integrity.
In the Reference 1 letter, Philadelphia Electric Company (PECo) committed to provide supplemental responses to requests 2b (2), 2b (5) and 3a for the Limerick Generating Station (LGS), Units 1 and 2.
The attachment provides these supplemental responses.
Also included in the attachment for LGS, Units 1 and 2 are minor correcticns to responses 1, 2b (3) and 2b (6).
If you have any questions, please contact us.
Very truly yours, e ) f~ /hkf
/
2301 d,-
I G. Jf Beck, Manager Licensing Section l
Attachment and Enclosure 1
cc:
T.
T.
Martin, Administrator, Region I, USNRC T.
J.
Kenny, USNRC Senior Resident Inspector, LGS n,
9211250009 9211ff
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PDR ADDCK 05000352 p
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COMMONWEALTli OF PENNSYLVANIA
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COUNTY OF CilESTER
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1 G.
R. Rainey, being first duly sworn, deposes and says i
That he is Vice President of Phi.1.adelphia Electric Company; that he has read the supplemental response to Generic Letter No.
92-01, and knows the contents thereof; and that the statements and t
matters set forth therein are true and correct to the best of his knowledge, information and belief.
F
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Subscribed and sworn to before me this /h day
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1992.
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LGS Units 1 and 2 Attachment Page 1
_ Request 1:
"1.
Certain addressoas are requested to provide the following information regarding Appendix H to CFR Part 50:
Addressees who do not have a surveillance program meeting ASTM E185-73, -79, or -82 and who do not have an integrated surveillance program approved by the_NRC (see Enclosure 2),
are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50.
Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC staff for review.
If the surveillance program is not to be revised to meet Appendix H to 10 CFR Part 50, addressees are requested to indicate when they plan to request an exemption from Appendix H to 10 CFR Part 50 under 10 CFR 50.60(b)."
Response
In our July 10, 1992 response to Generic Letter 92-01, we inadvertently provided the Limerick Generating Station (LGS), Unit 1 limiting end-of-life (EOL) reference temperature for nil-ductility transition (RTnuu) value of 56 F which is based on calculations performed in accordance with Regulatory Guide 1.99, Revision 1,
" Radiation Einbrittlement of Reactor Vessel Materials."
The correct RTnue value is 86*F which was calculated based on Revision 2 of itegulatory Guide 1.99.
The RTnae value provided in our July 10, 1992 response for LGS, Unit 2 (i.e., 120'F) is the correct value. to this attachment is a complete listing of the LGS, Untit 1 limiting values for the EOL RTnoe for the reactor vessel beltline plates and welds.
The LGS, Unit 1 limiting values for the EOL RTnoe were originally submitted to the NRC in our response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel-Materials and its Impact on Plant Operations," dated November 23, 1988, for only the most limiting values.
Table 5.3-5 of the LGS, Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) will be revised in the next revision to the UFSAR to reflect the Unit I limiting values calculated in accordance with Regulatory Guide 1.99, Revision 2, as contained in Enclosure 1.
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I E
LGS Units 1 and 2 Attachmsnt Paoe 2 Request 2b (2):
"2.
Certain addressees are requested tc provide the following information regarding Appendix G to 10 CFR Part 50:
b.
Addressees whose reactor vesscis were constructed to an ASME Code earlier than the Su. amer 1972 Addenda of the 1971 Edition are-requested to describe the consideration given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:
(2) the heat treatment received by all beltlin.J and surveillance materials;"
Response
In our July 10, 1! i. response to Generic Letter 92-01, Request 2b (2), we =cated that the specific heat treatment received by the beltline and surveillance materiala for LGS, Units 1 and 2 had not yet been located in existing plant documentation and that a search of vendor recorWA^J
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