ML071770257

2007/06/22-Comment (21) of L. M. Stinson on Behalf of Southern Nuclear Operating Company, Inc., on Proposed Rule Pr 50 Regarding Industry Codes and Standards; Amended Requirements

ML071770257
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	06/22/2007
From:	Stinson L M Southern Nuclear Operating Co
To:	NRC/SECY/RAS
SECY RAS
References
72FR16731 00021, NL-07-1232, PR-50, RIN 3150-AH76
Download: ML071770257 (7)

Text

DOCKETED USNRC L M. Stinson (Mike) Southern Nuclear Vice President Operating Companr Inc. Fleet Operations Support 40 lnverness Center Parkway Post Office Box 1295 Birmingham.

Alabama 35201 Tel205.992.5181 Fax 205.992.0341 June 22, 2007 Docket Nos.: 50-32 1 50-348 50-424 50-366 50-364 50-425 June 26,2007 (10:l lam) OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF SOUTHERN a COMPAlYY Energy to Serve Your NL-07-1232 Secretary, U. S. Nuclear Regulatory Commission ATTN: Rulemakings and Adjudications Staff Washington, DC 20555-0001 Southern Nuclear Operating Cornpa~iy Comments on NRC Proposed Rulemaking RIN 31 50-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 1673 1-1 6741 Ladies and Gentlemen: Southern Nuclear Operating Company (SNC) is pleased to have the opportunity to provide comments on proposed changes to I0 CFR 50.55a as described in RIN 3 150- AH76, published in the Federal Register, Vol.

72, No. 65, April 5,2007, PP. 16731- 16741. SNC's comments are provided in the enclosure to this letter. Additionally, SNC also endorses the comments provided by NEI in their letter of June 19,2007, from James H. Riley, Director Engineering, regarding this subject rulemaking.

If you have any questions, please advise. Sincerely, L. M. Stinson Vice President Fleet Operations Support

Enclosure:

Comments on NRC Proposed Rulemaking RIN 3 150-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 16731-16741 U. S. Nuclear Regulatory Commission NL-07-1232 Page 2 cc: Southern Nuclear Operating Comvany Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. R. Madison., Vice President - Hatch Mr. T. E. Tynan, Vice President - Vogtle Mr. D. H. Jones, Vice President - Engineering RType: CFA04.054; CHA02.004; CVC7000; LC# 14597 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Ms. K. R. Cotton, NRR Project Manager - Farley Mr. R. E. Martin, NRR Project Manager - Hatch Mr. B. K. Singal, NRR Project Manager - Vogtle Mr. E. L. Crowe, Senior Resident Inspector - Farley Mr. J. A. Hickey, Senior Resident Inspector - Hatch Mr. G. J. McCoy, Senior Resident Inspector - Vogtle Nuclear Enerw Institute Mr. A. Marion, Senior Director - Engineering Enclosure Comments on NRC Proposed Rulemaking RIN 3 150-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 16731-16741 Enclosure Comments on NRC Proposed Rulemaking RIN 3 150-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 1673 1-1 6741 10 CFR 50.55a(el16)(ii)(D)-Auemented Insnection of PWR RPV Heads: ASM E was requested by the N RC to develop Code Case N-729- 1 and worked over several years utilizing volunteer resources to complete this effort as a top priority task. Also, ASME has made special effort to interface with NRC staff, including meetings at NRC headquarters, to address concerns related to cracking in Alloy 600 materials.

In this proposed rulemaking the NRC has discounted these efforts by proposing additional conditions. The results would make Code Case N-729-1 more burdensome than the existing NRC Order EA-03-09.

In our view, the technical rationale for imposing the additional conditions does not support their imposition. Southern Nuclear Operating Company (SNC) is disappointed that the NRC has taken this position, given that Code Case N-729-1 was developed and approved using the latest available technical information via an ANSI-approved consensus process (ASME Code Committees) of which the NRC is a participant and voting member. Accelerated PWSCC test results for Alloy 690, 52, and 152 were publicly presented at the Electric Power Research Institute's (EPRIs) 2007 International PWSCC of Alloy 600 Conference in Atlanta, Georgia on June 1 1 - 14,2007. Presentations on Alloy 690,52, and 152 corrosion tests were given by Mitsubishi Heavy Industries, Pacific Northwest National Laboratory, and GE Global Research.

NRC representatives were also in attendance at this presentation. The Mitsubishi Heavy Industries PWSCC test results show no cracks have initiated in Alloy 690,52, and 152 materials in a high temperature simulated PWR environment over periods ranging from approximately 73,000 hrs, 84,000 hrs, and 85,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, respectively.

All testing was performed at 360 OC (680 OF). When adjusted for temperature, the test times completed to date are equivalent to over 50 years for base material and over 140 years for the weld materials.

No crack initiation has occurred in the Alloy 690, 52, and I52 materials. Alloy 600, 82, and 182 materials included in the test matrix initiated cracks consistent with industry experience.

Mitsubishi is evaluating the need to continue testing of these Alloy 690, 52, and 152 specimens that continue to show no crack initiation. The materials tested apply to the entire replacement reactor vessel head population supplied by Mitsubishi Heavy Industries to the USA market from 2003 to date. In addition, Nuclear Management Company provided Alloy 690, 52, and 152 materials fabricated by Mitsubishi Heavy Industries to the EPRI. Accelerated crack growth rate testing of these materials has been performed by Pacific Northwest National Laboratory and GE Global Research.

Crack growth rates of highly cold worked material are so low (10'~ mmls) that they have no engineering significance.

Page 1 of 4 Enclosure Comments on NRC Proposed Rulemaking RIN 3 150-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 1673 1-16741 Corrosion tests performed by multiple laboratories demonstrate Alloy 690,52, and 152 materials are highly resistant to PWSCC. Based on these test results, the proposed initial inspection and re-inspection intervals of 10 and 7 years, respectively, are overly conservative.

At the 2007 International PWSCC of Alloy 600 Conference, it was stated that all high susceptibility reactor vessel heads in the USA have been replaced. Therefore, requiring such a conservative inspection interval and inspection requirements for replacement heads will cause significant unnecessary radiation exposure in the performance of these inspections. Consequently, Code Case N-729-1 should not be included in 1 OCFR50.55a rulemaking.

Page 2 of 4 Enclosure Comments on NRC Proposed Rulemaking RIN 3 150-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 1673 1-1 6741 10 CFR 50.55a(g)(6)(iiWE)-Au~mented Inswction of Class 1 Allov-600 Commnents:

This paragraph would implement a requirement to perform a bare metal visual examination of Alloy-600 welds on PWR Class 1 components.

In this case, the NRC does not reference the industry efforts, especially those made through the EPRl Materials and Reliability Program (MRP) to address this issue. Every PWR in the United States has agreed to the implementation of MRP-139, which requires an augmented program to perform bare metal visual examinations on the large diameter Alloy-600 welds on a frequency that is almost identical to the schedule mandated in ASME Code Case N-722. Typically, utilities are allowed the option to assess each code case and determine if that code case should be adopted for use. By mandating the use of the Case the NRC is, in effect, writing their own Code and deviating from using guidance from an international consensus standard body (ASME Code Committees, of which the NRC is a participant and voting member). The NRC and the industry have been working on this issue, and industry programs are in place to cover these examinations. Additional time should be provided to allow the MRP and ASME to develop the necessary enhancements.

Page 3 of 4 Enclosure Comments on NRC Proposed Rulemaking RIN 3 150-AH76, Dated April 5,2007 Proposed Rules, Federal Register, Vol. 72, No. 65, PP. 1673 1-1 674 1 10 CFR50.55a(b)(21(xx)-Svstem Leabe Tests: Regardless of the edition or addenda of Section XI by which repairlreplacement activities are performed, NDE following welding or brazing would be performed immediately following the welding or brazing activity and prior to the system leakage test, rather than following the system leakage test. In accordance with IWA-4520 ofthe 2003 Addenda through 2004 Edition of Section XI, NDE following welding or brazing is required to be performed per the Construction Code noted within the repairlreplacement plan in order to assure structural integrity of the respective system boundary, and consequently public safety. Additional NDE requirements previously imposed in order to justify not performing a hydrostatic pressure test were removed when it was seen that the additional requirements did not provide increased verification of structural integrity.

Likewise, since the performance of a system leakage test and hydrostatic test provided the equivalent verification of leak tightness, requirements related to the performance of elevated pressure hydrostatic testing were similarly addressed. Therefore, reinstating the requirement to perform NDE per the 1992 Edition of Section Ill or later edition of Section 111, when the Construction Code was not Section Ill, is not commensurate with the increased burden and resource restraints (manpower, dose, outage scheduling) the requirements would cause, with no increased assurance of structural integrity.

Since some plants have already updated to the 2003 Addenda of Section XI, to impose NDE per the 1992 Edition of Section Ill or later edition of Section 111 for those plants would constitute a backfit based upon adding a new requirement which does not presently exist per the 2003 Addenda. The endorsement of IWA-4520(a) and IWA-4540(a) of the 2003 Addenda without any similar limitation provides further basis on which to consider this NDE imposition as a backfit. Page 4 of 4