ML16043A523

Request for Additional Information 2016/02/12 E-mail from R. Guzman to M.Whitlock Relief Requests IR-3-22 and IR-3-23 (CAC Nos. MF6573 and MF6574)

ML16043A523
Person / Time
Site:	Millstone
Issue date:	02/12/2016
From:	Richard Guzman Plant Licensing Branch 1
To:	Whitlock M Dominion Nuclear Connecticut
Guzman R
References
TAC MF6573, TAC MF6574
Download: ML16043A523 (3)

Text

From: Guzman, Richard Sent: Friday, February 12, 2016 5:22 PM To: Michael L Whitlock (Generation - 6) (michael.l.whitlock@dom.com); Craig D Sly (Generation - 6) (craig.d.sly@dom.com)

Cc: 'wanda.d.craft@dom.com'

Subject:

Millstone Unit 3 - Relief Requests IR-3-22 and IR-3 REQUEST FOR ADDITIONAL INFORMATION (CAC Nos. MF6573 AND MF6574)

Mike, The NRC staff has reviewed the information provided in the subject Relief Requests dated July 30, 2015 (Agencywide Documents Access and Management System Accession No. ML15216A363), and has determined that additional information is needed to complete its review. Shown below are the NRC staffs request for additional information questions. The information was discussed with your staff on February 10, 2016. As agreed, please provide your formal response by March 24, 2016. If you have any questions, please contact me.

Thanks,

~~~~~~~~~~~

Rich Guzman Sr. Project Manager NRR/DORL US NRC 301-415-1030 REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUESTS IR-3-22 AND IR-3-23 REGARDING WELD EXAMINATION COVERAGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION, UNIT 3 DOCKET NUMBER 50-423 CAC NOS. MF6573 AND MF6574 By letter dated July 30, 2015 (Accession Number ML15216A363), Dominion Nuclear Connecticut, Inc. (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code) specifically related to ASME Section XI, Examination Category C-F-1 and C-F-2, as well as Code Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division

1. Relief requests IR-3-22 and IR-3-23 pertain to the examination coverage of the reactor coolant system and main steam pipe to valve, elbow, flange or weldolet welds at the Millstone Power Station (MPS), Unit 3.

To complete its review, the U.S. Nuclear Regulatory Commission (NRC) staff requests the following additional information:

1. Pertaining to relief request IR-3-22:

1.1. While the NRC staff notes that examination of the far side of certain austenitic welds is not a qualified exam, the NRC staff also notes that best effort exams consisting of refracted longitudinal (L)-waves have been effective in identifying indications in the far side of welds, particularly in areas near the root of the weld. Given the sensitivity of these areas to degradation, the NRC staff requests clarification as to whether the ASME Code required ultrasonic testing (UT), (as well as the "Best Effort" examinations if performed) covered the regions that are typically susceptible to higher stresses and, therefore, potential degradation (i.e., the weld root and HAZ of the base material near the inside diameter surface of the joint).

1.2. For the welds for which relief is requested, please discuss: (a) the inservice inspection history (i.e., inspection years, disposition of detected flaws, extent of condition assessment, and correct of actions); (b) whether the licensee identified any indications during construction and preservice inspections (i.e., radiographic testing or surface examination, or both) on the volume not covered by UT; and (c) disposition of identified flaws.

1.3. Given the reduced inspection coverage of the weld under consideration: (a) Discuss any walkdowns (e.g., under Boric Acid Corrosion Control program or normal operator round) usually performed to monitor and identify leakage in an unlikely event of a through wall leak; and (b) Discuss reactor coolant system leakage detection capabilities at the plant, or any measures taken, to monitor and identify leakage during operation in an unlikely event of a through wall leak in the weld under consideration.

1.4. In an unlikely event of a potential through-wall flaw and leakage, discuss how such a leak will be addressed to minimize its impact on plant operations.

1.5. (a) Discuss any industry or plant-specific operating experience regarding potential degradation (e.g., stress corrosion cracking and corrosion) and potential severe loading (e.g., vibration, water hammer, and overloading) for the subject weld and associated components.

(b) The NRC staff notes that due to recent operating experience regarding thermal fatigue cracking in some plants, the Electric Power Research Institute (EPRI) issued an interim guidance, EPRI-MRP Interim Guidance for Management of Thermal Fatigue, (ADAMS Accession No. ML15189A100) that supplemented the existing industry thermal fatigue guidelines (e.g., MRP-146 and MRP-192) to better manage thermal fatigue cracking.

i. Please discuss whether any of these incomplete-coverage joints are affected by the instructions in the latest version of MRP-146 and the latest MRP Interim Guidance ii. If the inspection requirements of any of the subject joints are affected by MRP-146 or the Interim Guidance, please discuss whether there is any safety significance related to the areas of incomplete coverage.

1.6. Discuss whether the use of alternative volumetric examination techniques (e.g., the radiographic testing and phased array UT) would increase examination coverage.

2. Pertaining to relief request IR-3-23:

2.1. Clarify whether the ASME Code required UT, covered the regions that are typically susceptible to higher stresses and, therefore, potential degradation (i.e., the weld root and HAZ of the base material near the inside diameter surface of the joint).

2.2. For the welds for which relief is requested, please discuss: (a) the in-service inspection history (i.e., inspection years, disposition of detected flaws, extent of condition assessment, and correct of actions); (b) Discuss whether the licensee identified any indications during construction and preservice inspections (i.e., radiographic testing or surface examination, or both) on the volume not covered by UT; and (c) Discuss disposition of identified flaws.

2.3. Given the reduced inspection coverage of the weld under consideration: (a) Discuss any walkdowns (e.g., under Boric Acid Corrosion Control program or normal operator round) usually performed to monitor and identify leakage in an unlikely event of a through wall leak; and (b) Discuss reactor coolant system leakage detection capabilities at the plant, or any measures taken, to monitor and identify leakage during operation in an unlikely event of a through wall leak in the weld under consideration.

2.4. In an unlikely event of a potential through-wall flaw and leakage, discuss how such a leak will be addressed to minimize its impact on plant operations.

2.5. Discuss any industry or plant-specific operating experience regarding potential degradation (e.g., stress corrosion cracking and corrosion) and potential severe loading (e.g., vibration, water hammer, and overloading) for the subject weld and associated components.

2.6. Discuss whether use of alternative volumetric examination techniques (e.g., the radiographic testing and phased array UT) would increase examination coverage.