ML16088A208

Response to Request for Additional Information Regarding ASME Section XI Inservice Inspection Program, Relief Requests for Limited Coverage Examinations Performed in the Third 10-Year Inspection Interval

ML16088A208
Person / Time
Site:	Millstone
Issue date:	03/22/2016
From:	Mark D. Sartain Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
16-068
Download: ML16088A208 (13)

Text

'~~

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060

• -~'Dominion

-~'

Web Address: www.dom.com

. 'L*, March 22, 2016 U.S. Nuclear Regulatory Commission Serial No.16-068 Attention: Document Control Desk NLOS/WDC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUESTS FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE THIRD 10-YEAR INSPECTION INTERVAL By letter dated July 30, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted relief requests for limited coverage examinations performed in the first inspection period of the third 10-year inservice inspection interval for Millstone Power Station Unit 3 (MPS3). In an email dated February 12, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to Relief Requests IR-3-22 and IR-3-23. DNC agreed to respond to the RAI by March 24, 2016.

The attachments to this letter provide DNC's response to the NRC's RAI.

If you have any questions regarding this response, please contact Wanda Craft at (804) 273-4687.

Sincerely,

~-

Mark D. Sartain Vice President - Nuclear Engineering Commitments made in this letter: None

Serial No.16-068 Docket No. 50-423 Page 2 of 2 Attachments:

1. Response to Request for Additional Information Regarding ASME Section XI lnservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Third 10-Year Inspection Interval - IR-3-22

2. Response to Request for Additional Information Regarding ASME Section XI lnservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Third 10-Year Inspection Interval - IR-3-23 cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No.16-068 Docket No. 50-423 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUESTS FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE THIRD 10-YEAR INSPECTION INTERVAL RELIEF REQUEST IR-3-22 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 1 of 6 By letter dated July 30, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted relief requests for limited coverage examinations performed in the first inspection period of the third 10-year inservice inspection (ISi) interval for Millstone Power Station Unit 3 (MPS3). In an email dated February 12, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to Relief Requests IR-3-22 and IR-3-23. This attachment provides DNC's response to the NRC's RAI.

Pertaining to Relief Request IR-3-22 RAI 1.1 While the NRG staff notes that examination of the far side of certain austenitic welds is not a qualified exam, the NRG 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 NRG 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).

DNC Response As shown in the examination coverage plots and summary in Relief Request IR-3-22 for each of the welds with limited examination coverage, the examination volume on the far side of the weld is limited. The extent of the limitation is based on the component geometry. The examination volume on the far side of the weld centerline has been examined to the extent possible with qualified techniques and "best effort" techniques.

As a minimum, the region of the weld root has been examined.

RAI 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 corrective 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.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 2 of 6 DNC Response The ISi history of weld CHS-31-FW-1 included ultrasonic testing (UT) and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-31-FW-3 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected.

The radiographic testing (RT) performed to meet the construction code requirements was initially unacceptable due to incomplete fusion and burn-through identified in the weld. The weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

The ISi history of weld CHS-32-1-SW-D included UT and surface examinations performed in 1991, 2001 and 2011. The surface examination performed during 2011 identified a 5/64 inch base metal indication that was evaluated as acceptable to ASME code acceptance standards. All other examinations were acceptable with no flaws detected during these examinations. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-32-FW-1 included UT and surface examinations performed in 1991, 2001 and 2011. The surface examination performed during 2001 identified a 0.1 inch base metal indication that was evaluated as acceptable to ASME code acceptance standards. All other examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-33-1-SW-B included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-33-FW-1 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-33-FW-17 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected. The RT performed to meet the construction code requirements identified tungsten inclusions within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 3 of 6 The ISi history of weld CHS-33-FW-4 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected.

The RT performed to meet the construction code requirements was initially unacceptable due to incomplete fusion identified in the weld. The weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

The ISi history of weld RHS-9-2-SW-K included UT and surface examinations performed in 1999 and 2010. These examinations were acceptable with no flaws detected during these examinations. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld RHS-9-3-SW-B included UT and surface examinations performed in 1989, 1999 and 2010. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld RHS-9-FW-2 included UT and surface examinations performed in 1989, 1999 and 2010. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld SIL-157-FW-3 included UT and surface examinations performed in 1987, 2001 and 2010. These examinations were acceptable with no flaws detected.

The RT performed to meet the construction code requirements was initially unacceptable due to incomplete fusion identified in the weld. This weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

The ISi history of weld SIL-43-FW-1 included UT and surface examinations performed in 1987, 1999 and 2010. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld SIL-43-FW-16 included UT and surface examinations performed in 1987, 1999 and 2010. These examinations were acceptable with no flaws detected.

The RT performed to meet the construction code requirements identified porosity within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

The ISi history of weld SIL-504-1-SW-7 included UT and surface examinations performed in 1987, 1999 and 2010. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 4 of 6 The ISi history of weld SIL-504-FW-15 included UT and surface examinations performed in 1987, 1999 and 2010. The surface examination performed in 2010 identified a 1/8 inch rounded indication in the weld that was evaluated as acceptable to ASME code acceptance standards. All other examinations were acceptable with no flaws detected. The RT performed to meet the construction code requirements identified porosity within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

The ISi history of weld SIL-8-FW-3 included UT and surface examinations performed in 1987, 1999 and 2011. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

RAI 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.

DNC Response a) Welds SIL-157-FW-3, SIL-504-1-SW-7, SIL-504-FW-15, SIL-43-FW-1 and SIL FW-16 are located inside containment and therefore normally inaccessible during plant operation. Walkdowns required by the Boric Acid Program are performed at the beginning of each refueling outage. These welds are located within the general inspection areas covered by the Boric Acid Program.

The remaining welds identified in Relief Request IR-3-22 are located outside containment. These areas are walked down once per shift during operator rounds. Any significant leakage would have a high probability of being identified in the unlikely event of through-wall leakage.

b) The RCS leakage detection system provides for the detection of unidentified leakage in the unlikely event that a significant through-wall leak occurs inside containment. RCS leakage detection consists of containment sump level and sump pump run time monitoring; containment airborne particulate and gaseous radioactivity monitoring; and containment pressure, temperature, and humidity monitoring.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 5 of 6 RAI 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.

DNC Response Potential through-wall flaws and leakage would be addressed through established station procedures and the corrective action program. If a leak affects plant equipment governed by Technical Specifications (TS), an immediate operability determination would be performed and the appropriate TS action statement would be entered if operability is not reasonably assured. For some piping locations, the station may be able to perform evaluations as permitted by the ASME Code or approved alternatives to establish structural integrity and limit the effects of the leakage such that the leak may be repaired at a later time to minimize the impact on plant operations.

RAI 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 NRG 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, "EPRl-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.

DNC Response a) The weld locations are constructed of austenitic stainless steel, which has shown good performance in industry operating experience. As a group, none

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 6 of 6 of the welds are located near the reactor coolant loop piping or, for the Residual Heat Removal locations, in thermal mixing locations, and are therefore not subject to the thermal fatigue issues that have been the subject of industry operating experience. One set of the welds are located near the centrifugal charging pumps and may experience a moderate level of high frequency vibration. However, these welds are full penetration girth welds, not socket fitting welds, and therefore are not typically susceptible to high cycle vibration fatigue failure. The subject lines are maintained water-filled. During normal operation the lines are at ambient temperature conditions, which results in a low potential for stress corrosion cracking. The piping systems have been reviewed for transient loading conditions and are designed to accommodate the transient loading conditions, as applicable.

b) The weld joint locations are not within the areas of concern identified by the latest version of MRP-146 and MRP interim guidance.

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

DNC Response The use of alternative volumetric examination techniques was considered as part of the evaluation process to obtain the maximum examination coverage. The use of RT is not

• a desired alternative because RT is limited in its ability to detect service-induced flaws.

Use of other conventional or phased array techniques was considered, but would not increase the coverage due to the limitations created by the component configurations.

Serial No.16-068 Docket No. 50-423 ATTACHMENT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUESTS FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE THIRD 10-YEAR INSPECTION INTERVAL RELIEF REQUEST IR-3-23 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.16-068 Docket No. 50-423 Attachment 2, Page 1 of 3 By letter dated July 30, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted relief requests for limited coverage examinations performed in the first inspection period of the third 10-year inservice inspection (ISi) interval for Millstone Power Station Unit 3 (MPS3). In an email dated February 12, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to Relief Requests IR-3-22 and IR-3-23. This attachment provides DNC's response to the NRC's RAI.

Pertaining to Relief Request IR-3-23 RAI 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).

DNC Response As shown in the examination coverage plots and summary in Relief Request IR-3-23 for each of the welds with limited examination coverage, the examination volume on the far side of the weld is limited. The extent of the limitation is based on the component geometry. The examination volume on the far side of the weld centerline has been examined to the extent possible with qualified techniques. For the welds covered by this relief request, the region of the weld root and base material heat affected zone (HAZ) on the far side of the weld has been effectively examined for circumferential flaws.

RAI 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 corrective 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.

DNC Response The ISi history of weld DTM-25-FW-1 included Ultrasonic Testing (UT) and surface examinations performed in 1999, 2004 and 2011. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

Serial No.16-068 Docket No. 50-423 Attachment 2, Page 2 of 3 The ISi history of weld MSS-32-FW-2 included UT and surface examinations performed in 1995, 2004 and 2011. These examinations were acceptable with no flaws detected.

The radiographic testing (RT) performed to meet the construction code requirements identified tungsten inclusions and porosity within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

The ISi history of weld MSS-32-FW-3 included UT examination and surface examinations performed in 1999, 2004 and 2011. These examinations were acceptable with no flaws detected. The RT performed to meet the construction code requirements was initially unacceptable due to incomplete fusion identified in the weld. This weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

RAl2.3 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.

DNC Response a) The welds are located outside containment. These areas are walked down once per week during operator rounds. Any significant leakage would be identified in the unlikely event of through-wall leakage.

b) As the welds are located outside containment, the leakage detection capabilities of the reactor coolant system located inside containment do not apply.

RAl2.4 Jn 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.

DNC Response Potential through-wall flaws and leakage would be addressed through established station procedures and the corrective action program. If a leak affects plant equipment

Serial No.16-068 Docket No. 50-423 Attachment 2, Page 3 of 3 governed by the Technical Specifications, an immediate operability determination would be performed and the appropriate action statement would be entered if operability is not reasonably assured. For some piping locations, the station may be able to perform evaluations as permitted by the ASME Code or approved alternatives to establish structural integrity and limit the effects of the leakage such that the leak may be repaired at a later time to minimize the impact on plant operations.

RAI 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.

DNC Response The three welds are located on carbon steel main steam piping and are not susceptible to stress corrosion cracking mechanisms for austenitic and nickel-based materials. The flow accelerated corrosion (FAC) program considers this piping to have low potential for wall loss. The piping is designed for all anticipated loading conditions, including steam hammer due to valve closing.

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

DNC Response The use of alternative volumetric examination techniques was considered as part of the evaluation process to obtain the maximum examination coverage. The use of RT is not a desired alternative because RT is limited in its ability to detect service-induced flaws.

Use of other conventional or phased array techniques was considered, but would not increase the coverage due to the limitations created by the component configuration.

'~~

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060

• -~'Dominion

-~'

Web Address: www.dom.com

. 'L*, March 22, 2016 U.S. Nuclear Regulatory Commission Serial No.16-068 Attention: Document Control Desk NLOS/WDC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUESTS FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE THIRD 10-YEAR INSPECTION INTERVAL By letter dated July 30, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted relief requests for limited coverage examinations performed in the first inspection period of the third 10-year inservice inspection interval for Millstone Power Station Unit 3 (MPS3). In an email dated February 12, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to Relief Requests IR-3-22 and IR-3-23. DNC agreed to respond to the RAI by March 24, 2016.

The attachments to this letter provide DNC's response to the NRC's RAI.

If you have any questions regarding this response, please contact Wanda Craft at (804) 273-4687.

Sincerely,

~-

Mark D. Sartain Vice President - Nuclear Engineering Commitments made in this letter: None

Serial No.16-068 Docket No. 50-423 Page 2 of 2 Attachments:

1. Response to Request for Additional Information Regarding ASME Section XI lnservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Third 10-Year Inspection Interval - IR-3-22

2. Response to Request for Additional Information Regarding ASME Section XI lnservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Third 10-Year Inspection Interval - IR-3-23 cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No.16-068 Docket No. 50-423 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUESTS FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE THIRD 10-YEAR INSPECTION INTERVAL RELIEF REQUEST IR-3-22 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 1 of 6 By letter dated July 30, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted relief requests for limited coverage examinations performed in the first inspection period of the third 10-year inservice inspection (ISi) interval for Millstone Power Station Unit 3 (MPS3). In an email dated February 12, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to Relief Requests IR-3-22 and IR-3-23. This attachment provides DNC's response to the NRC's RAI.

Pertaining to Relief Request IR-3-22 RAI 1.1 While the NRG staff notes that examination of the far side of certain austenitic welds is not a qualified exam, the NRG 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 NRG 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).

DNC Response As shown in the examination coverage plots and summary in Relief Request IR-3-22 for each of the welds with limited examination coverage, the examination volume on the far side of the weld is limited. The extent of the limitation is based on the component geometry. The examination volume on the far side of the weld centerline has been examined to the extent possible with qualified techniques and "best effort" techniques.

As a minimum, the region of the weld root has been examined.

RAI 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 corrective 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.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 2 of 6 DNC Response The ISi history of weld CHS-31-FW-1 included ultrasonic testing (UT) and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-31-FW-3 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected.

The radiographic testing (RT) performed to meet the construction code requirements was initially unacceptable due to incomplete fusion and burn-through identified in the weld. The weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

The ISi history of weld CHS-32-1-SW-D included UT and surface examinations performed in 1991, 2001 and 2011. The surface examination performed during 2011 identified a 5/64 inch base metal indication that was evaluated as acceptable to ASME code acceptance standards. All other examinations were acceptable with no flaws detected during these examinations. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-32-FW-1 included UT and surface examinations performed in 1991, 2001 and 2011. The surface examination performed during 2001 identified a 0.1 inch base metal indication that was evaluated as acceptable to ASME code acceptance standards. All other examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-33-1-SW-B included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-33-FW-1 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld CHS-33-FW-17 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected. The RT performed to meet the construction code requirements identified tungsten inclusions within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 3 of 6 The ISi history of weld CHS-33-FW-4 included UT and surface examinations performed in 1991, 2001 and 2011. These examinations were acceptable with no flaws detected.

The RT performed to meet the construction code requirements was initially unacceptable due to incomplete fusion identified in the weld. The weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

The ISi history of weld RHS-9-2-SW-K included UT and surface examinations performed in 1999 and 2010. These examinations were acceptable with no flaws detected during these examinations. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld RHS-9-3-SW-B included UT and surface examinations performed in 1989, 1999 and 2010. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld RHS-9-FW-2 included UT and surface examinations performed in 1989, 1999 and 2010. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld SIL-157-FW-3 included UT and surface examinations performed in 1987, 2001 and 2010. These examinations were acceptable with no flaws detected.

The RT performed to meet the construction code requirements was initially unacceptable due to incomplete fusion identified in the weld. This weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

The ISi history of weld SIL-43-FW-1 included UT and surface examinations performed in 1987, 1999 and 2010. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

The ISi history of weld SIL-43-FW-16 included UT and surface examinations performed in 1987, 1999 and 2010. These examinations were acceptable with no flaws detected.

The RT performed to meet the construction code requirements identified porosity within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

The ISi history of weld SIL-504-1-SW-7 included UT and surface examinations performed in 1987, 1999 and 2010. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 4 of 6 The ISi history of weld SIL-504-FW-15 included UT and surface examinations performed in 1987, 1999 and 2010. The surface examination performed in 2010 identified a 1/8 inch rounded indication in the weld that was evaluated as acceptable to ASME code acceptance standards. All other examinations were acceptable with no flaws detected. The RT performed to meet the construction code requirements identified porosity within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

The ISi history of weld SIL-8-FW-3 included UT and surface examinations performed in 1987, 1999 and 2011. These examinations were acceptable with no flaws detected.

No indications were identified during construction or preservice examinations on the volume not covered by UT.

RAI 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.

DNC Response a) Welds SIL-157-FW-3, SIL-504-1-SW-7, SIL-504-FW-15, SIL-43-FW-1 and SIL FW-16 are located inside containment and therefore normally inaccessible during plant operation. Walkdowns required by the Boric Acid Program are performed at the beginning of each refueling outage. These welds are located within the general inspection areas covered by the Boric Acid Program.

The remaining welds identified in Relief Request IR-3-22 are located outside containment. These areas are walked down once per shift during operator rounds. Any significant leakage would have a high probability of being identified in the unlikely event of through-wall leakage.

b) The RCS leakage detection system provides for the detection of unidentified leakage in the unlikely event that a significant through-wall leak occurs inside containment. RCS leakage detection consists of containment sump level and sump pump run time monitoring; containment airborne particulate and gaseous radioactivity monitoring; and containment pressure, temperature, and humidity monitoring.

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 5 of 6 RAI 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.

DNC Response Potential through-wall flaws and leakage would be addressed through established station procedures and the corrective action program. If a leak affects plant equipment governed by Technical Specifications (TS), an immediate operability determination would be performed and the appropriate TS action statement would be entered if operability is not reasonably assured. For some piping locations, the station may be able to perform evaluations as permitted by the ASME Code or approved alternatives to establish structural integrity and limit the effects of the leakage such that the leak may be repaired at a later time to minimize the impact on plant operations.

RAI 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 NRG 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, "EPRl-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.

DNC Response a) The weld locations are constructed of austenitic stainless steel, which has shown good performance in industry operating experience. As a group, none

Serial No.16-068 Docket No. 50-423 Attachment 1, Page 6 of 6 of the welds are located near the reactor coolant loop piping or, for the Residual Heat Removal locations, in thermal mixing locations, and are therefore not subject to the thermal fatigue issues that have been the subject of industry operating experience. One set of the welds are located near the centrifugal charging pumps and may experience a moderate level of high frequency vibration. However, these welds are full penetration girth welds, not socket fitting welds, and therefore are not typically susceptible to high cycle vibration fatigue failure. The subject lines are maintained water-filled. During normal operation the lines are at ambient temperature conditions, which results in a low potential for stress corrosion cracking. The piping systems have been reviewed for transient loading conditions and are designed to accommodate the transient loading conditions, as applicable.

b) The weld joint locations are not within the areas of concern identified by the latest version of MRP-146 and MRP interim guidance.

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

DNC Response The use of alternative volumetric examination techniques was considered as part of the evaluation process to obtain the maximum examination coverage. The use of RT is not

• a desired alternative because RT is limited in its ability to detect service-induced flaws.

Use of other conventional or phased array techniques was considered, but would not increase the coverage due to the limitations created by the component configurations.

Serial No.16-068 Docket No. 50-423 ATTACHMENT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUESTS FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE THIRD 10-YEAR INSPECTION INTERVAL RELIEF REQUEST IR-3-23 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.16-068 Docket No. 50-423 Attachment 2, Page 1 of 3 By letter dated July 30, 2015, Dominion Nuclear Connecticut, Inc. (DNC) submitted relief requests for limited coverage examinations performed in the first inspection period of the third 10-year inservice inspection (ISi) interval for Millstone Power Station Unit 3 (MPS3). In an email dated February 12, 2016, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to Relief Requests IR-3-22 and IR-3-23. This attachment provides DNC's response to the NRC's RAI.

Pertaining to Relief Request IR-3-23 RAI 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).

DNC Response As shown in the examination coverage plots and summary in Relief Request IR-3-23 for each of the welds with limited examination coverage, the examination volume on the far side of the weld is limited. The extent of the limitation is based on the component geometry. The examination volume on the far side of the weld centerline has been examined to the extent possible with qualified techniques. For the welds covered by this relief request, the region of the weld root and base material heat affected zone (HAZ) on the far side of the weld has been effectively examined for circumferential flaws.

RAI 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 corrective 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.

DNC Response The ISi history of weld DTM-25-FW-1 included Ultrasonic Testing (UT) and surface examinations performed in 1999, 2004 and 2011. These examinations were acceptable with no flaws detected. No indications were identified during construction or preservice examinations on the volume not covered by UT.

Serial No.16-068 Docket No. 50-423 Attachment 2, Page 2 of 3 The ISi history of weld MSS-32-FW-2 included UT and surface examinations performed in 1995, 2004 and 2011. These examinations were acceptable with no flaws detected.

The radiographic testing (RT) performed to meet the construction code requirements identified tungsten inclusions and porosity within the weld. These indications were evaluated as acceptable to ASME code acceptance standards. No preservice indications were identified.

The ISi history of weld MSS-32-FW-3 included UT examination and surface examinations performed in 1999, 2004 and 2011. These examinations were acceptable with no flaws detected. The RT performed to meet the construction code requirements was initially unacceptable due to incomplete fusion identified in the weld. This weld was repaired, reexamined and determined to be acceptable. No preservice indications were identified.

RAl2.3 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.

DNC Response a) The welds are located outside containment. These areas are walked down once per week during operator rounds. Any significant leakage would be identified in the unlikely event of through-wall leakage.

b) As the welds are located outside containment, the leakage detection capabilities of the reactor coolant system located inside containment do not apply.

RAl2.4 Jn 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.

DNC Response Potential through-wall flaws and leakage would be addressed through established station procedures and the corrective action program. If a leak affects plant equipment

Serial No.16-068 Docket No. 50-423 Attachment 2, Page 3 of 3 governed by the Technical Specifications, an immediate operability determination would be performed and the appropriate action statement would be entered if operability is not reasonably assured. For some piping locations, the station may be able to perform evaluations as permitted by the ASME Code or approved alternatives to establish structural integrity and limit the effects of the leakage such that the leak may be repaired at a later time to minimize the impact on plant operations.

RAI 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.

DNC Response The three welds are located on carbon steel main steam piping and are not susceptible to stress corrosion cracking mechanisms for austenitic and nickel-based materials. The flow accelerated corrosion (FAC) program considers this piping to have low potential for wall loss. The piping is designed for all anticipated loading conditions, including steam hammer due to valve closing.

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

DNC Response The use of alternative volumetric examination techniques was considered as part of the evaluation process to obtain the maximum examination coverage. The use of RT is not a desired alternative because RT is limited in its ability to detect service-induced flaws.

Use of other conventional or phased array techniques was considered, but would not increase the coverage due to the limitations created by the component configuration.