ML15317A155

Request for Additional Information Email, Relief Request No. B-11 for Reactor Coolant System Pipe to Valve Weld, Second 10-year Inservice Inspection Interval

ML15317A155
Person / Time
Site:	Comanche Peak
Issue date:	11/13/2015
From:	Balwant Singal Plant Licensing Branch IV
To:	Hope T Luminant Generation Co
Singal B
References
CAC MF6556
Download: ML15317A155 (4)

Text

Burkhardt, Janet From: Singal, Balwant Sent: Friday, November 13, 2015 8:36 AM To: 'Hope, Timothy' (Timothy.Hope@luminant.com)

Cc: 'Jack.Hicks@luminant.com' (Jack.Hicks@luminant.com); Burkhardt, Janet

Subject:

Request for Additional Information (RAI) - Relief Request B-11 (CAC No. MF6556)

Attachments: RAI-MF6556-RR-B11.docx

Tim, By letter dated August 3, 2015 (Agencywide Documents Access ad Management System (ADAMS) Accession Number ML15224B364), Luminant Generation Company LLC requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Relief Request (RR) B-11 pertains to the examination coverage of the reactor coolant system pipe to valve weld at the Comanche Peak Nuclear Power Plant, Unit 2.

The NRC staff requests for the attached additional information to complete the review of this RR.

Draft RAI were transmitted on November 6, 2015. Jack Hicks of your organization informed the NRC staff on November 12, 2015 that a clarification call is not needed.

Please treat this e-mail as formal transmittal of RAIs. You are requested to respond to this RAI request within 30 days from the date of this e-mail.

Thanks.

1

REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST B-11 LUMINANT GENERATION COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT 2 DOCKET NUMBER 50-446 By letter dated August 3, 2015 (Agencywide Documents Access ad Management System (ADAMS) Accession Number ML15224B364), Luminant Generation Company LLC (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Relief Request (RR) B-11 pertains to the examination coverage of the reactor coolant system pipe to valve weld at the Comanche Peak Nuclear Power Plant, Unit 2.

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

1. The NRC staff notes that in a single-sided access or a limited access to far-side, the Best Effort examination using the refracted longitudinal (L) waves may be performed as an extra effort to interrogate the required examination volume on the far-side particularly the root of weld and the heat affected zone (HAZ) of base materials typically susceptible to high stresses and potential degradation. The L-waves have been shown capable of detecting planar flaws in the far-side volume of cast austenitic stainless steel and austenitic stainless steel weld and base materials in a single-sided examination. Given the reduced inspection coverage of the weld under consideration:

a. Please discuss whether the license performed the Best Effort examination as an extra effort to interrogate the examination volume on the far-side (particularly the root of weld and HAZ of base materials typically susceptible to high stresses and potential degradation).

b. Please provide percentage of coverage obtained from the Best Effort examination if this examination was performed.

c. Please clarify 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).

2. Section 2 of the RR states; The applicable ASME Boiler and Pressure Vessel Code (hereafter referred to as the Code) edition and addenda is ASME Section XI, Rule for Inservice Inspection of Nuclear Power Plant Components, 1998 Edition, through 2000 Addenda. In addition, as required by 10 CFR [Title 10 of Code of Federal Regulations] 50.55a, ASME Section XI, 1995

Edition, 1996 Addenda is used for Appendix VIII, Performance Demonstration for Ultrasonic Examination System.

The NRC staff notes that pursuant to 10 CFR 50.55a(b)(2)(xv),Section XI condition:

Appendix VIII specimen set and qualification requirements, licensees using Appendix VIII in the 1995 Edition through the 2001 Edition of the ASME Code may elect to comply with all of the provisions in paragraphs (b)(2)(xv)(A) through (M) of 50.55a, except for paragraph (b)(2)(xv)(F) of 50.55a, which may be used at the licensee's option.

Licensees using editions and addenda after 2001 Edition through the 2006 Addenda must use the 2001 Edition of Appendix VIII and may elect to comply with all of the provisions in paragraphs (b)(2)(xv)(A) through (M) of 50.55a, except for paragraph (b)(2)(xv)(F) of 50.55a, which may be used at the licensees option.

a. Please clarify if 1998 Edition through 2000 Addenda to the ASME Code,Section XI, Appendix VIII, was used for ultrasonic testing (UT) personnel qualification and procedures demonstration.

b. Please provide justification if other editions and addenda were used..

3. Please provide the following information:

a. Material specifications (e.g., austenitic stainless steel pipes SA-376, TP-304 and austenitic stainless steel ER-308 weldment) for the welds and associated components (e.g., pipe, and valves).

b. Thickness of each pipe.

4. Please describe the following:

a. The inservice inspection (ISI) 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.

c. Disposition of identified flaws.

5. Given the reduced inspection coverage of the welds under consideration, please discuss the following:

a. Any walkdowns (e.g., under Boric Acid Corrosion Control Program or normal operator rounds) usually performed to monitor and identify leakage in an unlikely event of a through wall leak.

b. 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.

6. In an unlikely event of a potential though wall flaw and leakage, please discuss significance of the leak and potential for structural failure of the subject weld.

7. Please discuss any industry or plant-specific operating experience regarding potential degradation (e.g., stress corrosion cracking, corrosion, and fatigue) and potential severe

loading (e.g., vibration, water hammer, and overloading) for the subject weld and associated components.

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 [Material Reliability Program] Interim Guidance for Management of Thermal Fatigue, (ADAMS Accession Number ML15189A100), that supplemented the existing industry thermal fatigue guidelines (e.g., MRP-1461 and MRP-1922) to better manage thermal fatigue cracking. Please discuss whether the licensee will take any compensatory measures to better manage thermal fatigue cracking in the subject weld to ensure structural integrity and leak tightness since essentially 100 coverage was not achieved by the UT.

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

8. It appears that the licensee reported in Attachments 1 and 2 to the RR that the UT was performed only in axial direction. Given that the licensee preformed a single-sided examination due to the valve body, please clarify whether scanning was done in both circumferential and axial directions, and the 50 percent coverage reported in Section 4 of the RR is the aggregate coverage of all ASME Code required UT performed in a single-sided examination (i.e., scanning of the weld in the circumferential and axial directions).

1 Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines (MRP-146) (Non-Publically Available) 2 Materials Reliability Program: Assessment of Residual Heat Removal Mixing Tee Thermal Fatigue in PWR [Pressurized Water Reactor] Plants (MRP-192) (Non-Publically Available)