ML071620072

Relief Requests NDE-SLH U2 and NDE-LSL U2 for Second 10-year Inservice Inspection Interval Re Reactor Vessel Shell Weld Inspections

ML071620072
Person / Time
Site:	Diablo Canyon
Issue date:	06/25/2007
From:	Hiltz T NRC/NRR/ADRO/DORL/LPLIV
To:	Keenan J Pacific Gas & Electric Co
Wang, A B, NRR/DORL/LPLIV, 415-1445
References
TAC MD2921
Download: ML071620072 (9)

Text

June 25, 2007 Mr. John S. Keenan Senior Vice President and CNO Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 770000 San Francisco, CA 94177-0001

SUBJECT:

DIABLO CANYON POWER PLANT, UNIT 2 - RELIEF REQUESTS NDE-SLH U2 AND NDE-LSL U2 FOR REACTOR VESSEL SHELL WELD INSPECTIONS (TAC NO. MD2921)

Dear Mr. Keenan:

By letter dated August 24, 2006, as supplemented on March 23, 2007, Pacific Gas and Electric Company (PG&E/licensee) requested relief from certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI inspection requirements for the reactor pressure vessels welds at the Diablo Canyon Power Plant (DCPP), Unit 2. The licensee submitted Relief Requests NDE-SLH U2 and NDE-LSL U2 requesting relief from the inspection requirements of the ASME Code pursuant to paragraph 50.55a(g)(5)(iii) of Title 10 of the Code of Federal Regulations (10 CFR), for the second 10-year inservice inspection interval.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and based on the information provided, has concluded that compliance with the ASME Code requirements are impractical and that the volumetric examinations performed provide reasonable assurance of structural integrity of these welds. Therefore, the NRC staff concludes that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

The staff's related safety evaluation is enclosed. If you have any questions or comments regarding the evaluation, please contact Alan Wang at (301) 415-1445.

Sincerely,

/RA/

Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-323

Enclosure:

Safety Evaluation

cc w/encl: See next page

ML071620072 *Safety Evaluation Memo OFFICE DORL/LPL4/PM DORL/LPL4/LA DCI/CVIB/BC OGC - NLO w/comment DORL/LPL4/BC NAME AWang JBurkhardt MMitchell* JBonanno THiltz DATE 6/18/07 6/18/07 6/5/07 6/21/07 6/25/07 Diablo Canyon Power Plant, Units 1 and 2 cc:

NRC Resident Inspector Jennifer Post, Esq.

Diablo Canyon Power Plant Pacific Gas & Electric Company c/o U.S. Nuclear Regulatory Commission P.O. Box 7442 P.O. Box 369 San Francisco, CA 94120 Avila Beach, CA 93424 City Editor Sierra Club San Lucia Chapter The Tribune ATTN: Andrew Christie 3825 South Higuera Street P.O. Box 15755 P.O. Box 112 San Luis Obispo, CA 93406 San Luis, Obispo, CA 94306-0112 Ms. Nancy Culver Director, Radiologic Health Branch San Luis Obispo State Department of Health Services Mothers for Peace P.O. Box 997414, MS 7610 P.O. Box 164 Sacramento, CA 95899-7414 Pismo Beach, CA 93448 Mr. James Boyd, Commissioner Chairman California Energy Commission San Luis Obispo County 1516 Ninth Street MS (31)

Board of Supervisors Sacramento, CA 95831 1055 Monterey Street, Suite D430 San Luis Obispo, CA 93408 Mr. James R. Becker, Vice President Diablo Canyon Operations and Mr. Truman Burns Station Director Mr. Robert Kinosian Diablo Canyon Power Plant California Public Utilities Commission P.O. Box 56 505 Van Ness, Room 4102 Avila Beach, CA 93424 San Francisco, CA 94102 Jennifer Tang Diablo Canyon Independent Safety Field Representative Committee United States Senator Barbara Boxer Attn: Robert R. Wellington, Esq. 1700 Montgomery Street, Suite 240 Legal Counsel San Francisco, CA 94111 857 Cass Street, Suite D Monterey, CA 93940 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 June 2007

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SECOND 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF NO. NDE-SLH U2 AND NDE-LSL U2 DIABLO CANYON POWER PLANT, UNIT 2 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NUMBER NO. 50-323

1.0 INTRODUCTION

By letter dated August 24, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML062420240), Pacific Gas and Electric Company (PG&E, the licensee) proposed Request for Relief NDE-SLH U2 and NDE-LSL U2 for Diablo Canyon Power Plant, Unit 2 (DCPP-2) for its second 10-year interval inservice inspection (ISI) program plan.

The licensee provided additional information by letter dated March 23, 2007 (ADAMS Accession No. ML070880743).

2.0 REGULATORY REQUIREMENTS The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) for Class 1, 2, and 3 components requires that ISI be performed in accordance with Section XI of the ASME Code and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(g), except where specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i).

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. 10 CFR 50.55a(g)(4)(i) requires that inservice examinations of components and system pressure tests conducted during the initial 10-year inspection interval comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months before the issuance of the operating license. The regulation at 10 CFR 50.55a(g)(4)(ii) requires that inservice examinations and system pressure tests during successive 10-year inspection intervals comply with the reference in 10 CFR 50.55a(b) 12 months before the start of the 10-year inspection interval, subject to the limitations and modifications listed therein. The ASME Code of record for the Diablo Canyon Unit 2 second 10-year ISI program, in which interval these welds were inspected, is the 1989

Edition of Section XI of the ASME Code, with no addenda. As required by 10 CFR 50.55a(g)(6)(ii)(C)(2), the licensee is using the 1995 Edition of Section XI of the ASME Code through the 1996 Addenda for the volumetric examinations of the reactor pressure vessel (RPV) welds.

3.0 EVALUATION 3.1 Component Identification The following are the affected welds in the DCPP-2 RPV; specifically, the ASME Code,Section XI, Examination Categories and Item Numbers covering examinations of the RPV welds. These Examination Categories and Item Numbers are from Table IWB-2500-1 of the ASME Code,Section XI.

Relief Examination Request Item Number Description Weld Numbers Category Number RPV Shell-to-Bottom B-A NDE-SLH U2 B1.11 Head Circumferential 10-201 Weld B1.12 RPV Lower Shell 3-201A, 3-201B, NDE-LSL U2 B-A Longitudinal Welds and 3-201C 3.2 ASME Code Requirements Category B-A, Item B1.11 of Table IWB-2500-1 to Section XI of the ASME Code requires that essentially 100 percent of the length of DCPP-2 RPV shell-to-bottom head circumferential weld number 10-201 be volumetrically examined during each 10-year ISI interval as shown in Figure IWB-2500-1, using the acceptance standard of IWB-3510. Category B-A, Item B1.12 of Table IWB-2500-1 to Section XI of the ASME Code requires that essentially 100 percent of the length of DCPP-2 RPV lower shell longitudinal weld numbers 3-201A, 3-201B, and 3-201C be volumetrically examined during each 10-year ISI interval as shown in Figure IWB-2500-2, using the acceptance standard of IWB-3510.

3.3 Licensees Basis for Relief Request Relief is requested from performing a portion of the volumetric examination where access is restricted by core support lugs and the bottom head taper.

The design of the RPV shell-to-bottom head circumferential weld and RPV lower shell longitudinal welds precludes the required examination of portions of these welds due to the presence of the six core barrel support lug locations. These portions of the welds are physically inaccessible due to the core support lug design and the bottom head taper. The support lugs and bottom head taper limit access of the vendor's RPV examination tool to a portion of the examination volume. All areas of the weld (75.36 percent for the RPV shell-to-bottom head

circumferential weld and 80.01 percent for the RPV lower shell longitudinal welds) accessible to the vendor's examination tool for ASME Code volumetric examination were examined.

Consideration was also given to examining the weld from the vessel outside diameter. The bottom head and shell insulation in this area is not designed to be removable and the close proximity of the insulated vessel to the concrete shield wall prohibits access. Providing access at this location would require redesign of the building concrete structure and the vessel insulation, and is impractical.

The inspection of the RPV lower shell-to-bottom head circumferential weld and RPV lower shell longitudinal welds were conducted using a procedure and techniques qualified by demonstration to the Electric Power Research Institute Performance Demonstration Initiative (PDI) consistent with the 1995 Edition with the 1996 Addenda of ASME Code,Section XI, Appendix Vlll, Supplements 4 and 6. The examinations used a combination of 45-degree angle search units with shear and longitudinal wave propagation and 30 millimeters (mm) or greater focal depths. The sound beams from these transducers were directed in four orthogonal directions parallel and perpendicular to the weld in as close proximity to the six core support lugs as the inspection device and transducer sled would allow.

The physical size of the transducer sled limited the approach to the core support lugs without hazarding the assembly and causing damage and loose parts concerns. Calculation of the weld and volume coverage afforded by each set of transducers by the automated data analysts resulted in a combined coverage of 75.36 percent for the RPV shell-to-bottom head circumferential weld and 80.01 percent for the RPV lower shell longitudinal welds.

3.4 Licensees Proposed Alternative Examination All accessible areas (75.36 percent of the ASME Code-required volume for the RPV shell-to-bottom head circumferential weld and 80.01 percent of the ASME Code-required volume for the RPV lower shell longitudinal welds) have been examined.

3.5 Evaluation The ASME Code,Section XI, Table IWB-2500-1, Category B-A, Items B1.11 and B1.12 requires that essentially 100 percent of the length of the DCPP-2 RPV welds be volumetrically examined during each 10-year ISI interval as shown in Figures IWB-2500-1 and IWB-2500-2.

The licensee is requesting relief from the ASME Code requirements as it was unable to obtain essentially 100 percent coverage of the weld length due to restrictions by the core support lugs and the bottom head taper. The licensee noted that the design of the RPV shell-to-bottom head circumferential weld and the RPV lower shell longitudinal welds precludes a portion of the welds from the required examination due to the presence of the six core barrel support lugs.

Furthermore, in addition to the support lugs, the bottom head taper limits access of the vendor's RPV examination tool to a portion of ASME Code-required examination weld length. The licensee performed the examinations from inside of the RPV and found that the physical size of the transducer sled limited the approach to the core support lugs without causing possible damage to the examination tool, which could cause loose parts to fall into the RPV. The licensee also considered examining the subject weld from the RPV outside diameter; however, in this area, the insulation is not designed to be removable and the close proximity of the

insulated vessel to the concrete shield wall prohibits the licensee access to perform the examination. Therefore, the NRC staff determined that the ASME Code-required examinations are impractical based on the RPV drawings and the licensees description of the RPV examination area. To require the licensee to perform the ASME Code-required examinations would be a burden because the RPV, associated attachments, and insulation would have to be redesigned.

The licensee conducted the ASME Code-required examinations using techniques qualified by PDI as required in the 1995 Edition through the 1996 Addenda of the ASME Code,Section XI, Appendix Vlll, Supplements 4 and 6. The licensee noted that the examination used a combination of 45-degree angle search units with shear and longitudinal wave propagation and 30 mm or greater focal depths. The sound beams from these transducers were directed in four orthogonal directions parallel and perpendicular to the weld in as close proximity to the six core support lugs as the inspection device and transducer sled would allow.

The licensee obtained a significant combined volumetric coverage of 75.36 percent for the RPV shell-to-bottom head circumferential weld and 80.01 percent for the RPV lower shell longitudinal welds. The NRC staff concludes that the volumetric coverage obtained for the RPV shell-to-bottom head circumferential weld and the RPV lower shell longitudinal welds by the licensee represents a significant portion of the ASME Code-required volume. As stated in its letter dated March 23, 2007, the licensee did not find any indications during its examinations of the RPV shell-to-bottom head circumferential weld. The licensee did find three indications in one of the RPV lower shell longitudinal welds (weld number 3-201-A), but these indications were found to be acceptable based on the criteria for planar flaws in Table IWB-3510-1 of Section XI to the ASME Code, 1989 Edition. The NRC staff notes that these three indications (first indication is 2.1 inches long by 0.13 inches deep, second indication is 1.0 inches long by 0.13 inches deep, and the third indication is 0.75 inches long by 0.13 inches deep) are not surface breaking and are minor when compared to the overall circumferential weld length and weld thickness. In addition to the ultrasonic examination, the licensee performed VT-3 examinations of the reactor vessel internal surface using remote underwater cameras. As indicated in its letter dated March 23, 2007, the licensee found no recordable indications during the VT-3 examination performed in accordance with Table IWB-2500-1, Category B-N-1, Item B13.10, and Category B-N-2, Item B13.60 of Section XI to the ASME Code. The licensee also found no indication of leakage during a system leakage test at full system temperature and pressure in accordance with Section XI of the ASME Code. Based on the above, the NRC staff determined that these ultrasonic examinations would have detected any significant patterns of degradation in these welds. Therefore, the NRC staff has determined that the ultrasonic examinations performed on 75.36 percent for the RPV shell-to-bottom head circumferential weld and 80.01 percent for the RPV lower shell longitudinal welds along with a VT-2 visual examination of the bottom of the RPV performed every refueling outage provides reasonable assurance of structural integrity of the subject welds.

4.0 CONCLUSION

S The NRC staff has reviewed the licensee's submittal for Relief Requests NDE-SLH U2 (RPV shell-to-bottom head weld number 10-201) and NDE-LSL U2 (RPV longitudinal lower shell weld numbers 3-201A, 3-201B, and 3-201C). Based on the information provided, the staff has concluded that compliance with the ASME Code requirements is impractical and that the

volumetric examinations performed provide reasonable assurance of structural integrity of these welds. Therefore, the NRC staff concludes that the licensees request for relief may be granted pursuant to 10 CFR 50.55a(g)(6)(i) for the second 10-year ISI interval.

The NRC staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. All other requirements of the ASME Code, Sections III and XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: J. Honcharik, NRR/DCI/CVIB Date: