3F0985-26, Forwards Safety Balance Assessment for Elimination of RCS Main Loop Pipe Break Protective Devices, Per 850201 Request for Exemption from 10CFR50,App A,Gdc 4.Meeting Arranged in Late Oct 1985 W/B&W to Discuss NRC Review
| ML20132F528 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/27/1985 |
| From: | Eric Simpson FLORIDA POWER CORP. |
| To: | Harold Denton NRC OFFICE OF ADMINISTRATION (ADM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20132F533 | List: |
| References | |
| 3F0985-26, 3F985-26, NUDOCS 8510010347 | |
| Download: ML20132F528 (2) | |
Text
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2 0.0,9 Power C ORPO R ATIO N September 27,1985 3F0985-26 Mr. Harold R. Denton Office of Nuclear Reactor Regulation Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Transmittal of Report Related to Request for Exemption from a Portion of 10 CFR 50 Appendix A, General Design Criteria 4
Reference:
- 1) Florida Power Corporation (FPC) letter to NRC, Westafer to Denton, dated February 1,1985 (3F0285-02), subject Request for Exemption from a Portion of 10 CFR 50, Appendix A, General Design Criteria 4 (GDC-4).
- 2) FPC letter to NRC, Westafer to Denton, dated August 30,1985 (3F0885-24), subject Re-evaluation of CR-3 Reactor Cooling System Loads Utilizing Leak-Before-Break Concept to Remove Reactor Coolant System Main Loop Pipe Break Protective Devices.
Dear Sir:
The reference I letter requested an exemption from a portion of the GDC-4 requirements in order to utilize the Leak-Before-Break concept at Crystal River Unit 3 (CR-3) and presented a sequence of actions (tentative dates of reports) which would provide additional justifications for the reduction at CR-3 in the !
number of large bore hydraulic snubbers restraining the reactor coolant pumps.
The reference 2 letter provided the initial report submitted in the sequence shown in reference 1. .
1 The report enclosed with this letter is the second to be submitted in the sequence l shown in reference I and is an assessment of the benefits and risks of eliminating Loss-Of-Coolant-Accident protective devices currently used in the CR-3 nuclear electric generating plant.
8510010347 850927 PDR
'ODl P
ADOCK 05000302 L l PDR I I
G EN ER AL OFFICE 3201 Thirty fourth Street South e P.O. B;x 14042, St. Petersburg, Florida 33733 e 813-366-5151
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September 27,1985 3F0985-26 l Page 2 We are arranging with NRC staff a technical meeting with FPC and B&W l representatives in late October,1985 to discuss the status of the NRC review and to identify needs for additional information, if required.- This meeting date is consistent with our continued need for informal input from NRC during November
-1985 to permit procurement by FPC of an optimized snubber arrangement to replace the current design.
Sincerely, (m P3 E. C. Simpson Director, Nuclear Operations Engineering & Licensing EHD/feb
Enclosure:
Report, B&W Document ID: 51-1159048-00, Safety Balance Assessment for Elimination of Reactor Coolant System Main Loop Break Protective Devices, Crystal River 3 Generating Plant I
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