ML20134P024: Difference between revisions

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#REDIRECT [[3F0885-24, Forwards B&W Evaluation of RCS Loads & Component Support Margins Resulting from Optimized Reactor Coolant Pump Support Configuration, Per NRC Needs for Info Identified at 850805 Meeting Re Exemption Request]]
| number = ML20134P024
| issue date = 08/30/1985
| title = Forwards B&W Evaluation of RCS Loads & Component Support Margins Resulting from Optimized Reactor Coolant Pump Support Configuration, Per NRC Needs for Info Identified at 850805 Meeting Re Exemption Request
| author name = Westafer G
| author affiliation = FLORIDA POWER CORP.
| addressee name = Denton H
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000302
| license number =
| contact person =
| document report number = 3F0885-24, 3F885-24, NUDOCS 8509060110
| package number = ML20134P025
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 2
| project =
| stage = Meeting
}}
 
=Text=
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Power C O R P O R A T I O Pd August 30,1985 3F0885-24 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20535
 
==Subject:==
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 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
 
==Reference:==
: 1)            Report BAW-1847, The B&W Owners Group Leak-Before-Break Evaluation of Margins Against Full Break for RCS Piping of B&W Designed NSS, September 1984
: 2)    Florida Power Corporation (FPC) letter to NRC, Westafer to Denton, dated February 1,1985, subject Request for Exemption from a Portion of 10 'CFR 50, Appendix A, General Design Criterion 4 (GDC 4)
: 3)  Meeting on August 5,1985 among NRC, Babcock and Wilcox, and FPC representatives to present FPC plans and define NRC staff needs for information to support the FPC request of Reference 2) above.
 
==Dear Sir:==
 
The Reference I report evaluated for the B&W Owners Group the Leak-Before-Break (LBB) concept as applied to B&W designed nuclear steam supply systems.
This B&W generic report concluded that a double-ended guillotine break will not occur, and postulated flaws producing detectable leakage exhibit stable r,rowth and, thus, allow a controlled plant shutdown before any potential exists for catastrophic piping failure. The scope of the evaluation included performing structural and fracture mechanics analym s using generic bounding data (loads and materials properties) for all B&W Own.,rs Group plants.                      FPC was one of the participating B&W Owners Group members and CR-3 was one of the plants bounded by the conclusions of the B&W evaluation.
                                                                                                                              - go;i B509060110 850830                                                                                    1 PDR  ADOCK 05000302 p                  PDR GEN ERAL OFFIC'E 320* Thirty fourth Street South e P.O. Box 14042, St. Petersburg, Florida 33733 e 813-866-5151
 
1 August 30,1985 3F0885-24 Page 2 The Reference 2 letter requested an exemption from a portion of the GDC-4 requirements in order to utilize the LBB concept at CR-3 and presents a sequence of actions (tentative dates of reports) which provide additional justifications for the reduction at CR-3 in the number of large bore hydraulic snubbers restraining (unnecessarily) the reactor coolant pumps.
At the Reference 3 meeting, plans and schedules of FPC were discussed as related to procurement decisions which must be made in November 1985 to permit snubber removal during Refuel VI, now scheduled to begin in March 1987. We outlined the planned content of scheduled submissions and requested NRC input as to additional content.
The report attached to this letter is the first to be submitted in the sequence shown in Reference 2, above. The report emphasizes results of the analyses performed by B&W which are specific to CR-3. Attachments to this letter are intended to be responsive to NRC staff needs identified at the Reference 3 meeting and in subsequent discussions by FPC and NRC representatives (Wilson and Bosnak).
We stress the need for informalinput from NRC during November 1985 and request that NRC arrange a technical meeting with FPC and B&W representatives during the week of October 21, 1985 to discuss the status of the NRC review and to identify needs for additionalinformation,if required.
Sincerel ,
4
          /    /
G. R. Westafer Manager, Nuclear Operations Licensing and Fuel Management EHD/feb
 
==Enclosures:==
Report, Evaluation of RCS Loads and Component Suoport Margins Resulting from Optimized RC Pump Support Configuration with Appendices A and B.
Attachment 1.
Attachment 2.
Attachment 3.
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