ML13268A079

Request for Additional Information Regarding Relief Request IR-03-17

ML13268A079
Person / Time
Site:	Millstone
Issue date:	10/22/2013
From:	James Kim Plant Licensing Branch 1
To:	Heacock D Dominion Nuclear Connecticut
Kim J
References
IR-03-17, TAC MF1314
Download: ML13268A079 (4)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 22, 2013 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.

lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT 3 - REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST IR-03-17 (TAC NO. MF1314)

Dear Mr. Heacock:

By letter dated March 28, 2013, Dominion Nuclear Connecticut, Inc. (the licensee) submitted alternative request IR-03-17, requesting relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code to permit certain degraded valves in the service water system to remain in service until they can be replaced.

The U.S. Nuclear Regulatory Commission staff has reviewed the information provided by the licensee and has determined that the enclosed request for additional information (RAI) is needed in order to complete the review. A response to this RAI is requested to be provided by February 7, 2014.

If you have any questions regarding this matter, please contact me at 301-415-4125.

Sincerely, James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosure:

As stated cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION REGARDING MILLSTONE POWER STATION, UNIT 3 RELIEF REQUEST IR-03-17 DOCKET NO. 50-423 By letter dated March 28, 2013 (Agencywide Document Access and Management System (ADAMS) Accession No. ML13091A038), Dominion Nuclear Connecticut Inc. (the licensee),

requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to permit certain degraded valves in the service water system to remain in service until they can be replaced. In order to complete our review, the U.S.

Nuclear Regulatory Commission (NRC) staff requests additional information.

1. Figure 5 of the Surry report shows a micrograph of a Surry valve which experienced selective leaching and was subsequently destructively analyzed. In this figure approximately 50 percent of the area appears to be occupied by phases not subject to selective leaching.

In this micrograph the non-selectively leachable phases appear to be continuous. Given the apparent continuity of the non-leachable phases, the NRC suspects that this sample will retain at least some strength even when fully leached. Alternatively, if the leachable phase was continuous, the NRC suspects that little or no strength would remain in a fully selectively leached specimen. Minor changes in chemistry or processing conditions of a component may significantly change the amount of each phase present. The NRC is aware that the Millstone valves have received a heat treatment which the Surry valves did not. The NRC is also aware that some heat treatments should reduce the formation of leachable phases while others increase the formation of leachable phases. The significance of such a treatment relative to chemistry differences between the valves remains unclear to the NRC.

Given the apparent tenuous balance between continuous and discontinuous phases in the Surry valve and the chemistry and processing differences between the Surry and Millstone valves, please justify why it should be assumed that the non-leachable phase is continuous in all locations in all valves at Millstone.

2. Figure 1 to enclosure 1 to the relief request (p 14 of 56) shows a graph of the tensile strength of samples taken from the Surry valve. These test specimens are represented as having varying degrees of dealloying. It is assumed that this was estimated optically based on the color change between leached and non-leached material. Only one sample tested is represented as being 100 percent dealloyed. One sample is not statistically significant.

Please justify why this single value is used as the basis for all structural calculations.

3. As indicated in item 2, above, the extent of selective leaching is often identified visually, by color change. Given that selective leaching is a chemical process in which aluminum is removed gradually from the selectively leachable phases, it is possible that the color change associated with selective leaching will occur long before all the aluminum is removed from the leachable phases. In this case a sample of material which is optically judged to be 100 percent leached may still contain aluminum in the leachable phases and may retain more strength than a sample of similar appearance from which all aluminum has been Enclosure

removed from the leachable phases. In order to consider a sample to be 100 percent leached, the NRC believes that it is necessary to demonstrate that the sample is 100 percent visibly discolored and to demonstrate that all of the aluminum has chemically been removed from the leachable phases. Such a demonstration may involve polishing the end of a broken tensile specimen and chemically examining the polished surface using an electron microscope. In light of the apparent lack of chemical analyses in support of the present data, please explain why the data provided in support of the relief request should be considered to represent fully dealloyed material.

4. The Surry report (Enclosure 2 to the relief request) contains a table of tensile test results.

This table reports that several of the specimens tested broke outside the gage marks. Given the design of tensile bars, this indicates the possibility that the measured tensile strengths overstate the actual tensile strengths of the material. Of particular interest is sample M. This sample is reported to be 100 percent dealloyed, to have a tensile strength of approximately 9 KSI, and to have failed in the grips of the tensile test machine. Given the clamping power and resulting friction of the grips, the actual tensile strength of this sample may have approached 0 KSI. Given the potential that the tensile strength of this sample was much lower than reported, please justify the tensile strength used in the structural analysis.

5. Since there are differences in the heat treatment and chemistry potential differences in continuous vs. discontinuous phases between the Millstone and Surry valves may exist. The validity of the test data as noted in the sample that failed in the grips and the lack of statistical significance of the Surry data is not conducive to concluding the valves have or will continue to have adequate strength. Additional tensile test data, chemistry data is needed to ensure strength of the material is adequate for structural integrity. Please provide additional information that will demonstrate adequate strength of the valves can be maintained.

October 22, 2013 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.

lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT 3- REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST IR-03-17 (TAC NO. MF1314)

Dear Mr. Heacock:

By letter dated March 28, 2013, Dominion Nuclear Connecticut, Inc. (the licensee) submitted alternative request IR-03-17, requesting relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code to permit certain degraded valves in the service water system to remain in service until they can be replaced.

The U.S. Nuclear Regulatory Commission staff has reviewed the information provided by the licensee and has determined that the enclosed request for additional information (RAI) is needed in order to complete the review. A response to this RAI is requested to be provided by February 7, 2014.

If you have any questions regarding this matter, please contact me at 301-415-4125.

Sincerely, Ira/

James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosure:

As stated cc w/encl: Distribution via Listserv Distribution:

PUBLIC LPLI-1 R/F RidsAcrsAcnw MaiiCTR Resource RidsNrrDoriDpr Resource RidsNrrDorllpl1-1 Resource RidsNrrDeEicb-Resource RidsNrrLKGoldstein Resource RidsNrrPMMillstone Resource RidsOgcRp Resource RidsRgn 1Mail Center Resource DAlley, NRR RidsNrrDeEpnb Resource ADAMS A ccess1on No.: ML13268A079 *S ee memo dated s eQ_tem ber 19, 2013 OFFICE LPL 1-1/PM LPL 1-1/LA DE/EPNB/BC LPL 1-1/BC (A)

NAME JKim KGoldstein Tlupold* RBeall DATE 10/21/13 09/30/13 9/19/2013 10/22/13 OFFICIAL RECORD COPY