ML20071J477
| ML20071J477 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 07/24/1994 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9407280185 | |
| Download: ML20071J477 (4) | |
Text
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Commonwealth Edison i
1400 Opus Placa Downers Grove, Illinois 60515 July 24,1994 Mr. William Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk
Subject:
Braidwood Station Units 1 and 2 Additional Information Relatint 'o the Request to Amend Technical Specification Sections 3.4.9.1 and 3.4.9.3 NPF-72/77: NRC Docket Nos. 50-456/457 l
References:
See Attachment
Dear Mr. Russell:
Reference 5 transmitted the Commonwealth Edison Company (Comed) request to amend Sa 'on 3.4.9.1 and 3.4.9.3 of the Technical Specification for Braidwood Units 1 vA 3. Reference 3 transmitted additional information which was requested during the referenced 4 teleconference. As discussed during the reference 2 teleconference, it was brought to Com Ed's attention that Technical Specification submittal needed to be revised to account for a 60 pounds per square inch gauge (psig) pressure channel measurement uncertainty. This submittal was subsequently supplemented in the reference 1 letter. Reference 3 also transmitted information on the Braidwood 1 Capsule U LTOPS. This information was subsequentif revised to include the 60 psig pressure channel measurement uncertainty and is attached for your review.
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Mr. Russell July 24,1994 If you have any questions concerning this correspondence please contact this office.
Sincerely,
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%.Au Denise M. Saccomando Nuclear Licensing Administrator Attachment cc:
R. Assa, Braidwood Project Manager-NRR S. Dupont, Senior Resident Inspector-Braidwood B. Clayton, Branch Chief-Region III Office of Nuclear Facility Safety-IDNS l
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RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION FOR BRAIDWOOD 1 CAPSULE U LTOPS 5.37 EFPY App G LTOP PORV MI or HI Limit (psig) App G Temp. ( F) Setpsint (psig)
Overshooundigl Limiting Utilized Margin (psig) 70 490 30 hil 581 61 97 490 30 h11 628 108 127 494 30 LII 720 196 147 504 30 MI 808*
274(266)**
177 525 30 h1I 998*
443(245)**
227 610 35 HI 1566*
921(155)**
277 670 54 HI (76)**
327 725 74 HI (1)**
377 700 98 HI (2)**
A constant PORV piping structural integrity criteria of 800 psig is conservatively imposed for setpoint development when the Appendix G limit exceeds 800 psig.
Margin to PORV structural limit in parenthesis.
Pressure channel measurement uncertainty from WCAP-10529 (not accounted for in setpoint development): 60 psig NEUTRON FLUENCE ON THE INSIDE DIAMETER (WETTED SURFACE) OF THE REACTOR VESSEL AT 5.37 EFPY IS 5.0 X 10'8 N/CM2,
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REFERENCES 1)
D. Saccomando letter to W. Russell dated July 21,1994, transmitting supplement to request to amend Technical Specification Sections 3.4.9.1 and 3.4.9.3 2)
Teleconference conducted July 18,1993, between the Commonwealth Edison Company and the Nuclear Regulatory Commission concerning the proposed Technical Specification Amendment to the Heat Up/ Cool Down Curves 3)
D. Saccomando letter to W. Russell dated July 11,1994, transmitting request to amend Technical Specification Sections 3.4.9.1 and 3.4.9.3 4)
Teleconference conducted hly 7,.1993, between the Commonwealth Edison Company and the Nuclem Regulatory Commission concerning the proposed Technical Specification Amendment to the Heat Up/ Cool Down Curves 5)
D. Saccomando letter to W. Russell dated March 30,1994, transmitting request to amend Technical Specification Sections 3.4.9.1 and 3.4.9.3 b
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