3F0186-12, Forwards Addl Info Re Request for Exemption from Portion of GDC 4,consisting of Responses to NRC Concerns Raised During 851031 Meeting,Rept on Reactor Coolant Pump Support Configuration Loadings & Benchmarking Info

Forwards Addl Info Re Request for Exemption from Portion of GDC 4,consisting of Responses to NRC Concerns Raised During 851031 Meeting,Rept on Reactor Coolant Pump Support Configuration Loadings & Benchmarking Info

ML20137E694
Person / Time
Site:	Crystal River
Issue date:	01/13/1986
From:	Westafer G FLORIDA POWER CORP.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20137E698	List: 3F0186-12, Forwards Addl Info Re Request for Exemption from Portion of GDC 4,consisting of Responses to NRC Concerns Raised During 851031 Meeting,Rept on Reactor Coolant Pump Support Configuration Loadings & Benchmarking Info ML20137E719
References
3F0186-12, 3F186-12, NUDOCS 8601170300
Download: ML20137E694 (6)

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..e Power CGR POR ATION January 13, 1986 3F0186-12 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Crystal River Unit 3 Docket No. 50-302 Operati ng Li cense No. DPR-72 Additional Information Regarding Request for Partial Exemption from General Design Criterion 4

Reference:

(1) 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)

(2) Meeting on October 31, 1985 among NRC, Babcock and Wilcox (B&W), and FPC Representatives to Discuss Partial Exemption f rom GDC-4 (3) NRC Summary Issued by H. Silver dated November 13, 1985 of the Reference (2) Meeting

Dear Sir:

The Reference 1 letter requested an exemption from a portion of the GDC-4 requirements in order to utilize the LBB concept at Crystal River Unit 3 (CR-3) to provide justification for reducing the number of large bore hydraulic s nu bbe rs rest rai ni ng the reactor coolant pumps.

At the Reference 2 meeting, plans and schedules of FPC were discussed and related to the schedule for procurement decisions to permit snubber removal during Refuel VI, scheduled to begin in g March 1987. We requested an identification of NRC concerns so that additional documentation could be provided, i f necessary. * '

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PDR ADOCK 05000302 *J P PDR gt.% g.% imp A)CN GENERAL OFFICE 3201 Thirty fourth Street South e P.O. Box 14042, St. Petersburg, Florida 33733 e 813-866-5151

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January 13,f1986 23F0186 PageL2 The Reference 3. report s'ummarized the meeting and identified the

-; remaining - information . desired by NRC .to complete its review of

- PPC 's exemption ' request. During the meeting, FPC. indicated that

?32 L snubbers . ( 8 per- pump) were. currently utilized and that an optimized design would utilize 8 snubbers _and 4 rigid. struts.

Following' the ' October 31, 1985 meeting, .B&W continued its optimization studies and concluded that an improved configuration could 7 utilize '4 snubbers and 4 struts. .Therefore, _a new report has been prepared by B&W utilizing this configuration.

'To f acilitate the NRC ' review of our. exemption request, several attachments to this letter are provided. Attachment A lists the NRC concerns identified at- the Octo ber. 31, 1985 meeting and g provides FPC responses. Attachment B is-the B&W report on the new snubber configuration.

- Atitachment- A provides overview responses to the NRC concerns and identifies the portions of Attachments B, C, D, and E where more

detailed information appears.

We regret . that .our response. has been delayed by several weeks

-because 'of the need for careful evaluation by B&W and FPC of tihe latest configuration.' We hope that the information transmitted

.previously, when coupled with the information transmitted by this letter, - will be suf ficient to permit completion of the NRC review by the end of January 11986.

Y As ' discussed with the NRC Project Manager for CR-3, some of the large . bore-'RCP snubbers are now being removed to permit' L . inspection and repair of the pump which failed on January 1,

1986. .By separate correspondence, we will be requesting that the

, removed . snubbers -(possibly including snubbers removed from other y _

pumps) . be replaced by the new pump support configuration.

Sincerely,

. /

G. - R. Westafer ~

Manager, Nuclear Operations .

tLicensing 'and Fuel Management

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Attachments: 'A,B, C, D, & E (identified on page 3)

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e. 10-January 13, 1986 3F0186-12

-Page 3' LIST OF ATTACHMENTS A: CONCERNS RAISED - BY NRC AT MEETING OF OCTOBER 31, 1985 AND

. RESPONSES.BY FPC (2 PAGES)

B: EVALUATION. OF REACTOR COOLANT SYSTEM COMPONENTS AND COMPONENT- SUPPORTS FOR OPTIMIZED REACTOR COOLANT PUMP SUPPORT CONFIGURATION LOADINGS (60 PAGES AND 6 DRAWINGS)

C:. INFORMATION ON BENCHMARKING AND NRC AUDITING OF B&W COMPUTER

-CODES (2 PAGES)

D: RELIABILITY AND SUPPORT CONFIGURATION (6 PAGES)

E: BASIS AND COSTS ASSOCIATED WITH VALUE IMPACT ANALYSIS (2 PAGES) ,

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-CONCERNS RAISED BY NRC AT MEETING

. OF OCTOBER- 31,-1985 AND RESPONSES BY FLORIDA POWER CORPORATION i

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-Concerns (C) Raised By NRC At Meetina O(

October 31. 1985 And Responses (R) By'FPC

-C1. FPC should also provide stress information.on component nozzles (including stress margins) and should state why SSE loads were not governing.

R1. Stress results for the optimized configuration and comparisons ~ to the original stress report and to Code allowable- stresses are shown in pages 19-48 of

-Attachment B. Results 'of the analysis indicate that

= all new stresses fall under code allowable stresses.

In most cases, SSE loads were not governing because Code allowable stresses for SSE are much higher than OBE allowable stresses.

-C2. Are there any hydraulic transients that should be included in the analysis?

-R2. An- evaluation was performed on the effects of guillotine breaks of branch piping, pump seal failure loadings, and other RCS piping hydraulic loads. All results were within acceptable design parameters of the

- piping, components,~ and remaining supports. This

' evaluation is discussed in pages 55-58 of Attachment B.

C3. Have STALUM and RESPECT been benchmarked?

R3. The codes have been certified to B&W QA procedures.

STALUM has been benchmarked internally against several structural test problems, calculations and/or other industry computer codes. Problems 1 and 2 of NUREG CR-1677 have been performed with the STALUM code and the results compare favorably, in the judgement of B&W. In addition, the NRC has audited both of these codes and specific lists of NRC Inspection Reports, and computer codes included in these audits are shown in Attachment C.

C4, A comparison of new margins with FSAR allowable stresses should be provided.

R4. The . comparison indicated that revised stresses were within Code allowable stresses. Comparisons of the optimized stresses with those shown in the original stress report and with Code allowable values are shown in several portions of Attachment B:

RV-support skirt - pages 29-31 OTSG support skirt - pages 37-38 A-1

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A comparison of SSE loads vs. design loads for the optimized RC pump ' supports s(snubbers / struts) is shown on page 48. '

The' revised seismic capacity factors for component supports 'as compared to those- calculated by LLNL is shown on pages 53 and 60.

C5. Document and docket the . reliability of the optimized

. support . configuration as discussed in the meeting of October 31, 1985.

RS. A description of the revised support configuration and a discussion of component and system reliability is shown in Attachment D. The reliability is enhanced because of fewer snubbers, improved design, and increased flexibility to accommodate thermal expansion and contraction.

, C6. Provide the basis for manhours in the value impact analysis and approximate cost savings in utilizing the optimized support configuration.

R6. Breakdowns for- manhours and costs are shown in Attachment E. Note that these costs are based on the configuration assumed at the October 31 meeting.

L Similar information was also provided to LLNL on L November 7, 1985 for their use in preparing (under l' contract to the NRC) value impact estimates for all B&W

• - plants. We anticipate (not based on detailed analysis) that greater benefits and reduced costs and exposure will result from the final optimized configuration. l C7. The licensee should consider independent verification of the revised design and ' installation. The design verification need not include a separate analysis, but should include a technical assessment of the process, inputs, and interfaces.

R7. FPC is planning to have a third party verification performed on the B&W analysis by an independent engineering organization. We consider that the internal reviews performed at B&W under the safety-related procedures give a high level of

_ confidence'that the design results are accurate and are based on valid processes, inputs and interfaces.

Nevertheless, we are proceeding with the third party

~ verification ' as a means of adding to the confidence level of both the NRC and the FpC organizations.

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