RC-12-0179, Relief Request RR-III-09 Request for Information Fatigue Crack Growth and Transients

Relief Request RR-III-09 Request for Information Fatigue Crack Growth and Transients

ML12325A056
Person / Time
Site:	Summer
Issue date:	11/16/2012
From:	Gatlin T South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RC-12-0179
Download: ML12325A056 (6)

Text

Thomas D. Gatlin Vice President, Nuclear Operations 803.345.4342 November 16, 2012 A SCANA COMPANY RC-12-0179 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 RELIEF REQUEST RR-II1-09 REQUEST FOR INFORMATION FATIGUE CRACK GROWTH AND TRANSIENTS

Reference:

1. Letter from T. D. Gatlin (VCSNS) to Document Control Desk (NRC),

"Reactor Vessel Head Penetration Weld Repair Under WCAP-1 5987,"

dated October 22, 2012 [ML12306A530]

2. Letter from T. D. Gatlin (ICSNS) to Document Control Desk (NRC),

"Relief Request RR-III-09 Alternative Weld Repair For Reactor Vessel Head Penetration," dated October 30, 2012

3. Letter from T. D. Gatlin (VCSNS) to Document Control Desk (NRC),

"Relief Request RR-III-09 Supplemental Information," dated November 5, 2012

4. Letter from T. D. Gatlin (VCSNS) to Document Control Desk (NRC),

"Relief Request RR-III-09 Request For Information," dated November 14, 2012 South Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for South Carolina Public Service Authority, hereby submits a response to the Request for Additional Information (RAI).

This letter contains no commitments.

Virgil C. Summer Station

• Post Office Box 88

• Jenkinsville, SC. 29065

• F (803) 345-5209

Document Control Desk CR-12-04775 RC-12-0179 Page 2 of 2 Should you have any questions, please call Bruce L. Thompson at 803-931-5042.

Very truly yours, Thomas D. Gatlin JGITDG/ts

Enclosure:

Relief Request RR-III-09, Fatigue Crack Growth RAIs :

Response to RAIs c:

K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams J. H. Hamilton R. J. White W. M. Cherry V. M. McCree R. E. Martin NRC Resident Inspector K. M. Sutton NSRC RTS (CR-12-04775, LTD 1331)

File (810.19-2)

PRSF (RC-12-0179)

Document Control Desk Enclosure CR-12-04775 RC-12-0179 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 REQUEST FOR ADDITIONAL INFORMATION Relief Request RR-II-09 Alternative Weld Repair For Reactor Vessel Head Penetration Fatigue Crack Growth By E-mail on November 15, 2012, additional information was requested to support the review of the subject relief request:

1. Please provide a sketch of the head cross section / J-groove weld!

nozzle tube / weld overlay showing the dimensions of the flaw assumed in the present analysis (LTR-PAFM-12-137-NP Rev. 2).

2. Which two or three of the reactor coolant system transients given in Table 2-2 of the analysis are the most significant drivers of fatigue crack growth? What are the stress and delta K associated with each of these transients?

3. Please explain why the analysis for the fatigue life of the Byron head with the assumed 2.54 inch J-groove weld flaw predicted "at least 10 years of service life time" (WCAP-16401) and the present analysis for the fatigue life of VC Summer head with the slightly smaller 2.32 inch assumed flaw size showed that the "fatigue crack growth for 20 years is insignificant."

[VCSNS Response]

The responses to these topics are provided in Attachment 1.

Document Control Desk CR-12-04775 RC-12-0179 Page 1 of 3 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 Response to RAIs

Document Control Desk CR-12-04775 / RC-12-0179 Page 2 of 3 Please provide a sketch of the head cross section / J-groove weld / nozzle tube I weld overlay showing the dimensions of the flaw assumed in the present analysis (LTR-PAFM-12-137-NP Rev. 2).

I S

I; Initial Overlay Initial Crack Depth ckness of 0.156" Iniia 2.32" l."

Note: The remaining overlay thickness is 0.142" after 20 years of fatigue crack growth Page 1 of 2

Document Control Desk CR-12-04775 / RC-12-0179 Page 3 of 3 Which two or three of the reactor coolant system transients given in Table 2-2 of the analysis are the most significant drivers of fatigue crack growth? What are the stress and delta K associated with each of these transients?

The initial AK for the three thermal transients that have the most significant contribution to fatigue crack growth are as follows:

1. Loss of Load (200 cycles for 40 years), AK = 20.8 ksi-in 1/2

2. Unit Unloading (18300 cycles for 40 years), AK = 4.9 ksi-in112

3. Heatup (200 cycles for 40 years), AK = 16.6 ksi-in112 Please note that only AK is provided since the time history stress data are too voluminous and would not be helpful in assessing the magnitude of fatigue crack growth.

Please explain why the analysis for the fatigue life of the Byron head with the assumed 2.54 inch J-groove weld flaw predicted "at least 10 years of service life time" (WCAP-16401) and the present analysis for the fatigue life of VC Summer head with the slightly smaller 2.32 inch assumed flaw size showed that the "fatigue crack growth for 20 years is insignificant."

A more detailed set of thermal transients was analyzed for V.C. Summer as compared to WCAP-16401. The thermal transient stresses used in WCAP-16401 are based only on the five most significant thermal transients out of the full set of normal/upset thermal transients. The effects of the remaining thermal transients are conservatively accounted for simply by adding the design cycles of these unanalyzed thermal transients to the design cycles of the five most significant transients analyzed.

For V. C. Summer, the full set of normal/upset thermal transients was analyzed and provided a more representative set of thermal transient stresses for the fatigue crack growth analysis. This more precise set of stresses resulted in a longer service life for the weld repairs.

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