ML20209B459
| ML20209B459 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 01/21/1987 |
| From: | Ainger K COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2632K, NUDOCS 8702040049 | |
| Download: ML20209B459 (2) | |
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CONHINNnveelth Edloon A t, One First Nabonal Plaza, Chicago, Hlinois j,
/ Address Reply tx Post Omco Box 767 U Chica00,lHinois 60690 0767 4
January 21, 1987 1
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation j
U.S. Nuclear Regulatory Commission Washington, DC.
20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Steam Generator Tube Rupture NRC Docket Nos. 50-454/50-455 and 50-456/50-457 2
Dear Mr. Denton:
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~ Commonwealth Edison is a member of the Westinghouse Owners Group (WOG) SGTR Licensing Issues Subgroup that is currently interacting with the NRC to address generic SGTR issues. The Subgroup has recently completed responses to various NRC questions from the Reactor Systems, Plant Systems, and Facilities Operations Branches in this regard. Although final NRC conclusions have not been reached at this time, the purpose of this letter is to address the need for Byron and Braidwood plant specific reviews concerning SGTR.
The Subgroup analytical efforts have resulted in the following publications: WCAP-10698 "SGTR Analysis Methodology to Determine the Margin i
to Steam Generator Overfill". Supplement 1 to WCAP-10698 " Evaluation of
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Offsite Radiation Doses for a Steam Generator Tube Rupture Accident", and WCAP-11002 " Evaluation of Steam Generator Overfill Due to a Steam Generator i
Tube Rupture Accident". This effort was based on a reference plant and reference site using standard design basis methodology and analysis while using conservative assumptions and initial conditions relative to margin to
-overfill. Also, this work was supplemented with the use of more realistic or best estimate assumptions and analysis. Each yielded positive results.
l It was demonstrated that the operator can perform the required SGTR recovery actions according to NRC approved emergency response guidelines and terminate primary to secondary leakage prior to steam generator overfill, thus demonstrating adequate margin to steam generator overfill. Given this configuration without overfill and assuming the worst single failure, the evaluation demonstrated offsite radiation doses to be within the allowable dose guidelines for a design basis SGTR. Because the use of licensing basis operator action times demonstrated that overfill will not occur for a design basis SGTR, it was necessary to further assume extended operator action times
_ to assure that steam generator overfill would occur prior to termination of i
primary to secondary leakage. Given this transient with a consequential l
failure of a safety valve, the evaluation found the offsite dose consequences of the forced overfill to continue to be acceptable.
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' Because of the conservative approach taken and the conclusions reached by the subgroup relative to the SGTR issue, it would appear that very little, if any, additional analysis is necessary for the NRC to generically resolve the SGTR issue for all the WOG Subgroup plants. This is currently being discussed between the NRC and the Subgroup. As we understand, the current NRC tentative schedule for issuing the SER is June 1987 and subsequent plant specific reviews will then be requested.
Upon our receipt of the NRC SER detailing the outstanding concerns, Commonwealth Edison will initiate the requisite review necessary to provide plant specific information to the NRC to resolve the SGTR issue for Byron and Braidwood Stations. At this time, we envision that the margin to steam generator overfill can be adequately demonstrated without further plant specific analysis through the relative comparison of our plant equipment and emergency operating procedures to those of the Subgroup reference plant.
Further, the Subgroup methodology or equivalent will be used to address the Byron and Braidwood offsite dose consequences. We expect that these described efforts could be completed within one year from the time of our receipt of the NRC safety evaluation. In the interim, we do not foresee any risk to plant operations because our operators are trained in plant specific emergency
' operating procedures which address the symptoms of SGTR events.
Please direct any questions regarding this matter to this office.
1 Very truly yours, 1
i K. A. Ainger Nuclear Licensing Administrator
/klj cc: Byron Resident Inspector Braidwood Resident Inspector Region III Office 2632K i
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