NL-07-1491, Response to Request for Additional Information Concerning Projected Fluence Values for the Limiting Circumferential Weld

Response to Request for Additional Information Concerning Projected Fluence Values for the Limiting Circumferential Weld

ML072150606
Person / Time
Site:	Hatch
Issue date:	08/03/2007
From:	George B Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-07-1491
Download: ML072150606 (6)

Text

Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201 -1295 COMPANY August 3, 2007 Energy to Serve Your Worldsu Docket No.: 50-321 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 1 Response to Request for Additional lnformation Concerning Projected Fluence Values for the Limitina CircumferentialWeld Ladies and Gentlemen:

Pursuant to your request for additional information of July 6, 2007 Southern Nuclear Operating Company (SNC) hereby provides its response concerning the projected fluence values for the limiting circumferential weld. ISI-ALT-8 (letter NL-07-0270, dated March 8,2007) was submitted to extend the authorization for the elimination of the RPV circumferential shell weld examinations required by ASME Code through the renewed license period of extended operation (PEO) at Plant Hatch-Unit 1.

NRC's request for additional information and SNC's response are provided in the enclosure to this letter.

This letter contains no NRC commitments. If you have any questions, please advise.

B. J. ~ e o j ~ e Manager, Nuclear Licensing

Enclosure:

Restatement of NRC Request for lnformation and SNC Response for Plant Hatch - Unit 1 Concerning Projected Fluence Values for the Limiting Circumferential Weld

U. S. Nuclear Regulatory Commission NL-07-1491 Page 2 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, Vice President - Hatch Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch

Edwin I. Hatch Nuclear Plant Unit 1 Response to Request for Additional lnformation Concerning Projected Fluence Values for the Limiting Circumferential Weld Enclosure Restatement of NRC Request for lnformation and SNC Response for Plant Hatch Unit 1

Enclosure Restatement of NRC Request for lnformation and SNC Response for Plant Hatch Unit 1 Concerning Projected Fluence Values for the Limiting Circumferential Weld Restatement of NRC Request for Additional Information Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i),

Request for Alternative ISI-ALT-8 proposed an alternative to extend the authorization for the elimination of the Reactor Vessel (RV) circumferential shell weld examinations required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI through the end of the extended period of operation at Hatch 1. In its January 28, 2005 Safety Evaluation Report (SER) for your original relief request on this topic (RR-38), the NRC staff authorized the elimination of these circumferential shell weld examinations only through July 31, 2007. The staff approval of RR-38 for this limited duration was based on a finding that the neutron fluence values used in RR-38 to justify elimination of the RV circumferential weld examinations through the end of the extended period of operation were based on a fluence calculational methodology (RAMA) that had not yet been reviewed and approved by the NRC staff at that time.

The use of the RAMA ,fluence methodology was subsequently reviewed and conditionally approved by the NRC in an SER dated May 13, 2005. The conditions for the application of the RAMA fluence methodology to the RV fluence calculations at Hatch 1 have been met.

The staff requests that you confirm whether the projected fluence value for the limiting circumferential weld at the end of the extended period of operation at Hatch 1 (i.e., the fluence intended for use as a basis for the current request, ISI-ALT-08) is the same as that reported for Hatch 1 in Enclosure 1 of RR-38 for the period corresponding to 54 effective full power years (EFPY) of facility operation.

If the 54 EFPY fluence for Request ISI-ALT-8 is different than what was originally reported in Enclosure 1 of RR-38 for 54 EFPY, please provide an updated mean RTNDT value for the limiting Hatch 1 circumferential shell weld and compare it to the acceptance criterion for the circumferential weld RTNDT value specified for Combustion Engineering fabricated RVs in Table 2.6-5 of the July 28, 1998, NRC SER for the Boiling Water Reactor Vessel and lnternals Project (BWRVIP)-05 report, "BWR Vessel and lnternals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)."

Enclosure Restatement of NRC Request for Information and SNC Response for Plant Hatch Unit 1 Concerning Projected Fluence Values for the Limiting Circumferential Weld SNC Response for Plant Hatch Unit 1 The projected fluence value calculated by RAMA for the limiting circumferential weld at the end of the period of extended operation (PEO) at Hatch 1 is higher than that previously reported for Hatch 1 in Enclosure 1 of Relief Request 38 (RR-38). As noted in the request for additional information (RAI), Relief Request RR-38, which used GE fluence methodology, was approved on a limited time basis (through July 2007) pending review and approval by the NRC of the RAMA fluence methodology. ISI-ALT-8 proposed a code alternative to examination of the circumferential welds that would extend through the PEO based on calculations using the RAMA methodology. As shown in the table below, the fluence at the end of the PEO was previously stated in RR-38 as 0.236 x 10

n/cm2while the RAMA calculated fluence used in ISI-ALT-8 at the end of the PEO is 0.296 x 10" n/cm2. 'The end of the PEO for Plant Hatch Unit 1 in RR-38 was based on an assumed 90% capacity factor, or 54 EFPY (effective full power years of operation). For the RAMA calculations, a more detailed evaluation of the capacity factor was performed, which resulted in the end of the PEO being defined at 49.3 EFPY.

For comparison purposes, updated values for the limiting Hatch 1 circumferer~tial shell weld at the end of the PEO and the values specified for Combustion Engineering (CE) fabricated RVs in Table 2.6-5 of the July 28, 1998, NRC SER are shown in the below table. Also included for comparison are the values used as the basis for RR-38.

As demonstrated in the table, the increase in projected fluence using the RAMA methodology has resulted in a slight increase in mean RTNDT at the end of the PEO (53.3 vs. 48.5). However, the Hatch 1 RTNDT is still bounded by the mean RTNDT calculated using either CE(VIP) or CE(CE0G) chemistry data as provided in the 1998 NRC SER for BWRVlP-05.

Group CE(VIP) CE(CE0G) Hatch 1 Hatch 1 End of PEO End of 64 EFPY 64 EFPY 54 EFPY PEO (Table 2.6-5) (Table 2.6-5) (RR-38) 49.3 EFPY (G E) (RAMA)

Cu% 0.13 0.183 0.197 0.197 Ni% 0.71 0.704 0.060 0.060 CF 151.7 172.2 91.O 91.O Fluence 0.40 0.40 0.236 0.296 (10 n/cm2)

ARTNDT 113.2 128.5 48.5 53.3 (OF)

RTNDT(U) 0 0 -10 -10 (OF)

Mean RTNDT 113.2 128.5 38.5 43.3 (OF)

P(F/E) 1.99E-4 4.38E-4 --- ---

NRC P(F/E) --- --- --- ---

BWRVIP