3F0389-19, Forwards Pipe Rupture Analysis Criteria Outside Reactor Bldg Crystal River Unit 3, for Review.Util Chose to Update Existing Pipe Rupture Criteria to Include Technical & Regulatory Advances on line-by-line Basis

Forwards Pipe Rupture Analysis Criteria Outside Reactor Bldg Crystal River Unit 3, for Review.Util Chose to Update Existing Pipe Rupture Criteria to Include Technical & Regulatory Advances on line-by-line Basis

ML20248J618
Person / Time
Site:	Crystal River
Issue date:	03/31/1989
From:	Ken Wilson FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20248J620	List: 3F0389-19, Forwards Pipe Rupture Analysis Criteria Outside Reactor Bldg Crystal River Unit 3, for Review.Util Chose to Update Existing Pipe Rupture Criteria to Include Technical & Regulatory Advances on line-by-line Basis ML20248J628
References
3F0389-19, 3F389-19, GL-87-11, NUDOCS 8904140489
Download: ML20248J618 (3)

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Power-COR PO R ATION

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March 31, 1989 3F0389-19 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington,..D.C. 20555

Subject:

Crystal River Unit 3 Docket No. 50-302 ,

Operating License No. DPR-72 High. Energy Line Break Outside Reactor Building Criteria

Dear Sir:

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Florida Power Corporation (FPC) is submitting six (6) copies of " Pipe Rupture Analysis Criteria Outside the Reactor Building Crystal River Unit 3" for NRC review. The submittal of this revised licensing and design basis criteria was discussed in FPC's letter dated December 16, 1988.

The original High Energy Line Break (HELB) Criteria outside containment for Crystal River Unit 3 (CR-3) is documented in Gilbert Associates, Incorporated's (GAI) Report Number 1811, Revision 4 as discussed in FSAR Section 5.4.4. The design methodologies and protection requirements were based on standard practice and approved criteria at that time (1973). The rules and guidelines to address the HELB issue provided in Appendix A to 10CFR50, General Design Criteria 4 (GDC-4), " Environmental ~ and Missile Design Bases" were in the developmental stage during that time frame and were therefore not included in the initial CR-3 HELB licensing position. However, the Safety Evaluation Report (SER) issued for CR-3's operating license recognized this fact and found that CR-3 met the intent of the GDC's.

Significant technical and regulatory advances in pipe rupture postulation and protection requirements have taken place since initial plant design almost 20 years ago. Most recently the emphasis has been placed on specific criteria changes that improve plant safety and reduce personnel exposure by eliminating non-stress related breaks and their associated pipe rupture restraint structures. FPC has chosen to update the existing pipe, rupture criteria for CR-3 to include thet e

' improvements, and to bring the pipe rupture criteria in line with 89041404e9 890331 { o0 fDR ADOCK 0500o302 l PDC IP POST OFFICE BOX 219

• CRYSTAL RIVER, FLORIDA 326. LO219 + (904) 563 2943 l A Florida Progress Company

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l Mhrch 31, 1989 3F0389-19 Page 2 l

current regulatory positions on a line-by-line basis as discussed with the NRC Staff at our earlier meetings. Specific criteria has been developed and plant systems been identified, for which upgrading is appropriate.

The attached report identifies and documents Florida Power Corporation's updated position on the various issues pertaining to pipe rupture requirements outside containment. The position has been established considering the technical and regulatory requirements at the time of plant design and construction, current NRC Standard Review Plan (SRP) guidance, and Generic Letter 87-11, modified as justified, to be compatible with existing design bases methods for CR-3. The purpose of this updated criteria is to provide acceptable pipe rupture l postulation and protection methods for the plant that meet the intent of current NRC requirements.

This report describes the updated methods and general criteria used to postulate and protect pipe rupture effects outside containment at CR-

3. It provides an integrated set of design bases (between existing and updated methods) for the protection of plant structures and equipment vital for public health and safety from the potentially adverse effects of pipe whip, jet impingement, compartment pressurization environmental influences of steam and water spray, and flooding associated with a postulated pipe rupture.

The criteria for postulating break locations and providing pro' cection methods for Reactor Coolant System primary piping or other piping inside the containment structure are not within the scope of this report. In addition, the methods defined for plant protection against missiles is also outside the technical scope of this report. Sections 4.2 through 6.5 of the GAI 1811 Report are also retained as Appendix A to this report, since they represent the original documented methods and results for rupture protection for CR-3.

FPC has already begun use of this document to establish break and crack locations. FPC's approach is to shield all breaks unless the installation of shields would not fit, restrict maintenance, interfere with inservice inspection, or be in an area where hardware installation wot Id create conflicts with FPC's ALARA program. FPC is reviewing schedules to determine what portion of the high energy piping can be shielded in the near term with CR-3 operating and during Refuel 7 wnich is scheduled to begin in March 1990. Therefore, an expeditious review of this criteria is requested to identify and resolve any areas of possible NRC disagreement with our planned approach. Approval by July 14, 1989 will allow FPC to maintale the accelerated schedule we have discussed with the staff.

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['>,,_. March;31,'1989.'

3F0389-19 .

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i. ' FPC ,. ' would . i suggest a- meeting in early May 1989. to discuss this l

. submittal . and answer any questions members of the Staff may have l regarding ~ur. o approach.

jl Sincerely, K. R.. Wilson, Manager l Nuclear Licensing KRW/JWT/sdr xc: ' Regional Administrator, Region II Senior Resident Inspector I

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