ML20210K905

License Change Request 44 to License DPR-72,adding Condition That Design & Licensing Basis of Reactor Coolant Pump Supports Need Not Include Consideration of Effects of Postulated Pipe Ruptures.Rept Encl

ML20210K905
Person / Time
Site:	Crystal River
Issue date:	04/24/1986
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20210K894	List: 3F0486-13, Forwards License Change Request 44 to License DPR-72,adding Condition That Design & Licensing Basis of Reactor Coolant Pump Supports Need Not Include Consideration of Effects of Postulated Pipe Ruptures ML20210K905 ML20210K922
References
NUDOCS 8604290116
Download: ML20210K905 (4)

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ATTACMENT A FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 Docket No. 50-302/ License No. DPR-72 Operating License Change Request No. 144 LICENSE DOCUMENT INVOLVED: Facility Operating License DESCRIPTION OF REQUEST:

Modify the Operating License of Crystal River Unit 3 (CR-3) to include the following new condition:

The design and licensing basis of the Reactor Coolant Pump supports for Crystal River Unit 3 (as described in Attachment C) need not include consideration of the effects of postulated pipe ruptures of the primary reactor coolant loop piping.

REASON FOR REQUEST:

This submittal is prompted by the supplementary information in the final rule which modified General Design Criterion 4 (GDC-4) of Appendix A, 10CFR50. These statements indicate that changing load path designs requires submittal of a license amendment for NRC approval.

SHOLLY EVALUATION OF REQUEST:

It has been demonstrated generically (NUREG-1061, Volume 3 dated 11/84) through the efforts of the NRC Piping Review Committee that large, thick walled piping systems will not catastrophically fail under postulated design conditions without first progressing through a substantial period of time wherein the system will leak copiously. This concept, known as Leak-Before-Break (LBB) has been demonstrated by the B&W Owners Group in Topical Report BAW-1847, to be applicable to the BAW 177FA Lowered Loop designs. FPC further demonstrated (see references in Attachment B) that it is specifically applicable to CR-3 and can be utilized in the design of CR-3 Reactor Coolant Pump Supports. The specific details of this configuration and the acceptability of the design are specified in Reference 10. Additionally, leakage detection capabilities and utilization thereof by the FPC operating staff are sufficient for detection of RCS leakage and removal of the unit from service prior to any gross piping failures.

As a result, FPC considers this request to not represent a consideration of a significant hazard since it removes a design restriction now demonstrably without merit. Further, it has been recognized oy the NRC Piping Review Committee, the ACRS, and the Commission itself, that adding flexibility to systems subjected to substantial thermal loading cycles results in an l improvement in the overall safety margin of the plant.

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Attachment A Page '

Based on the above, FPC finds that the amendment will not:

1. Involve a significant increase in the probability or consequence of an accident previously analyzed because it is recognized in the final rule modifying GDC-4 that the probability of occurrence of an accident is not increased when LBB technology is applied to heavy component supports in the reactor coolant system. The specific design for CR-3 reactor coolant pump supports is in accordance with approved guidance. Additionally, the proposed amendment affects only the design and licensing basis of RCP supports and as such  ;

does not affect the accident mitigation features of the plant.

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2. Create the possibility of a new or different kind of accident from any accident previously analyzed because the specific design for '

CR-3 reactor coolant pump supports is in accordance with approved guidance. RCP support configuration has been optimized by removing supports needed only for pipe rupture induced loads, which analysis has demonstrated can no longer occur. Additionally, the proposed change introduces no new mode of plant operation.

3. Involve a significant reduction in the margin of safety. The overall safety margin of the plant will actually be improved since >

the optimized design of the CR-3 RCP supports will result in i improved support system performance and reliability.

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W ATTACHMENT B List of References

! 1. B&W Report, BAW 1847, dated October 1984, subject the B&W Owners Group Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS.

2. FPC letter to NRC, Westaf er to ~ Denton, dated February 1, 1985 (3F0285-02), 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 l- Reference 2 above.

4. FPC letter to NRC, Westafer to Denton, dated August 30, 1985 i (3F0885-24), subject Re-evaluation of CR-3 Reactor Cooling System Loads Utilizing Le a k-B e fo re-B rea k Concept to Remove Reactor Coolant System Main Loop Pipe Break Protective Devices; transmitted B&W Report prepared for FPC, subject

, Evaluation of . Reactor Coolant System Loads and Component l Support Margins Resulting from Optimized Reactor Coolant Pump Support Configuration.

5.- FPC letter to NRC, Simpson to Denton dated September 27, 1985 (3F0985-26), subject Transmittal of Report Related to Request for Exemption from a Portion of 10 CFR 50, Appendix A, General Design Criterion 4 (GDC-4); transmitted B&W Report, Document ID 51-1159048-00, prepared for FPC, subject Safety Balance Assessment for Elimination of Reactor Coolant System Main Loop Pipe Break Protective Devices.

6. B&W Report, BAW 1847, Rev. 1, dated October 7, 1985, same subj ect 'as in Reference 1; revised report issued to address comments and questions raised by NRC staff reviewers.

7. FPC letter to NRC, Westafer to Denton, dated October 29, l

1985 (3F1085-13), subject Transmittal of Report Related to Request for Exemption from a Portion of 10 CFR 50, Appendix A, General Design Criterion 4; transmitted report prepared by FPC, subject Assessment of CR-3 RC Leak Detection System, j File: SP 83-133, dated October 25, 1985.

8. Meeting on October 31, 1985 among NRC, BaW, and FPC j representatives to discuss partial exemption from GDC-4

9. NRC summary issued by H. Silver dated November 13, 1985 of the reference (8) meeting.

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10. FPC letter to . NRC, Westafer to Denton, dated January 13, 1986 (3F0186-12), subject Additional Information Regarding Request for Partial Exemption from General Design Criterion 4.

11. FPC letter to NRC, Westafer to Denton, dated January 16, 1986 L (3F0186-18), subject Technical Speci fication ' Change Request l - No . . _14 2 ; proposes to. remove the tabular' list of snubbers from 1 the Technical Specifications in accordance with' NRC. guidance  !

provided in Generic Letter 84-13.

! 12. FPC letter to NRC, Simpson to Denton, dated January 21, 1986, subject Snubber Optimization Approval Request.

13. Meeting on February 27, 1996 among-NRC, FPC, B&W,-and G11 bert Commonwealth, Lynchburg, Va., to discuss . ten questions raised '

by NRC.

! 14. FPC letter to NRC, Westafer to Denton, dated April 2, 1986, s ubj ect Snubber Optimization Approval Request, documenting FPC's response to the ten questions of the reference (13) meeting.

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