ML20216H910
| ML20216H910 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/09/1997 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 6710-97-2387, NUDOCS 9709170106 | |
| Download: ML20216H910 (2) | |
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GPU Nucien inc
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Route 441 south NUCLEAR N';,'L'$*,*,*n,a Tel 717444-7621 September 09e 1997 6710 97 2387 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI 1)
Operating 1.icense No. DPR 50 Docket No. 50 289 License Amendment Request No. 269-AdditionalInformation As discussed with the NRC staff on Au;".ist 22,1997, this letter provides additional infonnation regarding TMI-l License Amendment itequest No. 269, previously submitted to the NRC on August 14,1997 (GPU Nuclear letter 6710-97 2345).1.icense Amendment Request No. 269 requested NRC review and approval of an increase in the postulated steam line break accident analysis environmental dose consequences associated with postulated accident induced steam generator tube leakage not previously analyzed. License Amendment Request No. 269 identified that the 10 CFR 50 Appendix A, GDC 19 limits for the control room were not affected by this change since the source tenu assumed for the TMI-l control room habitability analysis is based on the postulated Maximum liypothetical Accident (MilA) which is bounding. The following discussion provides additional infonnation to demonstrate that the proposed increased release during a steam line break remains bounded by the MilA release assumed for control room habitability.
The design basis loss-of colant accident (LOCA) was analyzed for the TMI l control room habitability analysis as documented in GPU Nuclear letter to the NRC, dated October 18,1985 (5211-85 2172). The postulated source tenn release for the control room habitability analysis was based on the TMI-l Updated Final Safety Analysis Report (UFSAR) Section 14.2.2.5, Maximum llypothetical Accident, assumptions which include 100 percent release of noble gases, 50 percent of the halogens (including iodine), and I percent ohhe solids. This postulated source I
term release is approximately 6 times higher than the steam line break source term release. The TMI l control room habitability analysis conservatively assumed that 100 percent of the ih~ I O radionuclides released to the reactor building were available for release to the environment using 9709170106 970909 ADOCKOSOOOg9 ljlj lj l]ll]llgljj]l PDR P
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6710 97 2387 Page 2 of 2 building leak rates consistent with Regulatory Guide 1 A, and 100 percent of the radionuclides released to the Auxiliary lluilding were conservatively assumed to infiltrate to the control building ventilation system. 'lhis analysis was approved by NRC Safety Evaluation Report (SliR) dated August 14,1986 (5211 86-3215) which concluded that the analytical approach was conservative and that the dose consequences were acceptable, The Thil 1 control room habitability dose consequences were reanalyzed to support Thil 1 Technical Specification Change Request No. 266 as a result ofincreased postulated accident recirculation system leakage. These revised control room doses remain below 10 CFR 50 Appendix A, GDC 19 limits.
The TMI l steam line break accident analysis is not the controlling accident in terms of postulated source term release for Thil 1 control room habitability. A qualitative comparison of the accident analysis assumptions for the radiation source terms, source to receptor distances, and applicable X/Q values for the postulated Thil 1 steam line break accident and hillA accident indicate that the control room habitability dose for the steam line break accident would be approximately 2.5 times smaller than the dose calculated for the hillA. This comparison provides reasonable assurance that the Thil 1 control roorn habitability analysis based on the hillA remains bounding and that a specific control room habitability study using steam line break assumptions is not required, if any additional information is needed, please contact hir. David J. Distel, Regulatory Affairs, at (201)316 7955.
Sincerely.
1%)J Wl James W. Langenbdd/
Vice President and Director, Thil DJD cc; Administrator, Region i Thil Senior Resident inspector Thil l Senior Project hianager I