ML13176A400

Catawba Nuclear Station, Units 1 and 2, Proposed Relief Request 13-CN-001, Use of Alternate Depth Sizing Criteria (TAC MF1168)

ML13176A400
Person / Time
Site:	Catawba
Issue date:	07/08/2013
From:	Pascarelli R J Plant Licensing Branch II
To:	Henderson K Duke Energy Carolinas
Kim S J NRR/DORL/LPLII-1
References
13-CN-001, TAC MF1168
Download: ML13176A400 (9)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 July 8, 2013 Mr. K. Henderson Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745 CATAWBA NUCLEAR STATION, UNIT 2, PROPOSED RELIEF REQUEST 13-CN-001, USE OF ALTERNATE DEPTH SIZING CRITERIA (T AC NO. MF1168) Dear Mr. Henderson: By letter dated March 18, 2013, Duke Energy Carolinas, LLC (the licensee) submitted a request, Relief Request (RR) 13-CN-001, to the Nuclear Regulatory Commission (NRC) for the use of an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, related to inservice inspection (lSI) of welds. Specifically, pursuant to Title 10 ofthe Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), the licensee requested to use the alternative in RR 13-CN-001 on the basis that the ASME Code requirement is impractical. RR 13-CN-001 provides an alternative for the depth sizing of indications detected using ultrasonic testing (UT) of the dissimilar metal welds at the reactor coolant system hot leg and cold leg when examined from the inner diameter surface. The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the proposed alternatives for weld lSI would provide reasonable assurance of leak-tightness and structural integrity of the piping and component segments identified in RR 13-CN-001, and that complying with the specified ASME Code,Section XI, requirements is impractical. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff authorizes the licensee's proposed alternatives as described in RR 13-CN-001 for the duration of the Catawba, Unit 2, third 10-year inservice inspection interval, currently scheduled to end on August 19, 2016. All other ASME Code,Section XI, requirements, for which relief was not specifically requested and authorized herein by the NRC staff, remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.

K. Henderson -2 If you have any questions, please contact the Project Manager, James Kim at 301-415-4125 or via e-mail at james.kim@nrc.gov. Sincerely, Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. Safety cc w/encl: Distribution via UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 13-CN-001 DEPTH SIZING EXAMINATION OF THE HOT LEG AND COLD LEG WELDS CATAWBA NUCLEAR STATION UNIT 2 DUKE ENERGY DOCKET NO. SO-414 1.0 INTRODUCTION By letter dated March 18, 2013, Duke Energy Carolinas, LLC (the licensee) submitted a request, Relief Request (RR) 13-CN-001, to the Nuclear Regulatory Commission (NRC) for the use of an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, related to inservice inspection (lSI) of welds. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) SO.SSa(g)(S)(iii), the licensee requested to use the alternative in RR 13-CN-001 on the basis that the ASME Code requirement is impractical. RR 13-CN-001 provides an alternative for the depth sizing of indications detected using ultrasonic testing (UT) of the dissimilar metal welds at the reactor coolant system hot leg and cold leg when examined from the inner diameter surface. 2.0 REGULATORY EVALUATION Pursuant to 10 CFR SO.SSa(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection (lSI) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1 O-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR SO.SSa(b), 12 months prior to the start of the 120-month interval, subject to the conditions listed therein. 10 CFR SO.SSa(g)(S)(iii), states that"... If the licensee has determined that conformance with a code requirement is impractical for its facility, the licensee shall notify the NRC and submit, as specified in Section SO.4, information to support the determinations. Determinations of Enclosure impracticality in accordance with this section must be based on the demonstrated limitations experienced when attempting to comply with the code requirements during the lSI interval for which the request is being submitted. Requests for relief made in accordance with this section must be submitted to the NRC no later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought. .. " 1 0 CFR 50.55a(g)(6)(i), states that " ... The Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines 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 ... " Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC staff to grant the relief requested by the licensee. 3.0 TECHNICAL EVALUATION 3.1 RELIEF REQUEST 13-CN-001 3.1.1 ASME Code Components Affected Class 1 Reactor Coolant System Reactor Vessel Hot Leg Nozzle-to-Safe End and Cold Leg Nozzle-to-Safe End Dissimilar Metal Welds Listed in the table below. Description Size Inside diameter (ID) Nozzle-to-Safe End Weld No. Inspection Plan Summary Number ASME Section XI Item No or Code Case N-770-1 Inspection Item Hot Leg Nozzle 2A to Safe End Weld 29.0" 10 (Nom.) 2RPV-202-121ASE C2.85.10.0006 85.10 C2.G7.2.0001 A-2 Hot Leg Nozzle 2B to Safe End Weld 29.0"10 (Nom.) 2RPV-202-121 BSE C2.B5.10.0005 B5.10 C2.G7.2.0002 A-2 Hot Leg Nozzle 2C to Safe End Weld 29.0" 10 (Nom.) 2RPV-202-121CSE C2.B5.10.0008 85.10 C2.G7.2.0003 A-2 Hot Leg Nozzle 20 to Safe End Weld 29.0" 10 (Nom.) 2RPV-202-1210SE C2.85.10.0007 85.10 C2.G7.2.0004 A-2 Cold Leg Nozzle 2A to Safe End Weld 27.5" 10 (Nom.) 27.5" 10 (Nom.) 2RPV-201-121ASE C2.G7.3.0001 8 Cold Leg Nozzle 28 to Safe End Weld 2R C2.G7.3.0002 8 Cold Leg Nozzle 2C to Safe End Weld 27.5" 10 (Nom.) 2RPV-201-121CSE C2.G7.3.0003 8 Cold Leg Nozzle 20 to Safe End Weld 27.5" 10 (Nom.) 2RPV-201-1210SE C2.G7.3.0004 8

-The safe end is made of SA-182, Type F316. The nozzle-to-safe end weld is made of nickel-based Alloy 182 with a Alloy 82/182 butter. The nozzle is SA-508, Class 2 with stainless steel cladding. 3.1.2 Applicable Code Edition and Addenda The code of record for the third 10-year lSI interval is the ASME Code,Section XI, 1998 Edition through the 2000 Addenda. 3.1.3 Applicable Code Requirements IWA-2232 requires that ultrasonic examinations be conducted in accordance with Mandatory Appendix I. Appendix I, 1-2220 requires that ultrasonic examinations be qualified by performance demonstration in accordance with Mandatory Appendix VIII. Appendix VIII, Supplement 10, "Qualification Requirements For Dissimilar Metal Piping Welds", Paragraph 3.2(b) requires that "Examination procedures, equipment, and personnel are qualified for depth sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, is less than or equal to 0.125 in. (3.2 mm)." Note that volumetric examinations of the Reactor Vessel nozzle-to-safe end dissimilar metal welds are also required to be performed in accordance with Section XI, Appendix VIII, as required by 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xvi). 10 CFR 50.55a(g)(6)(ii)(F) mandates the use of ASME Code Case N-770-1, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1It, and places additional conditions on its use. Code Case N-770-1, Table 1, Footnote (4) applies to volumetric examination of Inspection Items A-2 and B, and requires that "Ultrasonic volumetric examination shall be used and shall meet the applicable requirements of Appendix VIII." ASME Code Case N-695, Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1, provides alternatives to the reqUirements of Appendix VIII, Supplement 10. Paragraph 3.3(c) of this code case requires that "Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)." 3.1.4 Reason for Request The licensee stated that since 2002, the nuclear power industry has attempted to qualify personnel and procedures for depth sizing for indications when examining from the inside diameter surface of dissimilar metal and austenitic stainless steel butt welds in PWR piping using UT. The licensee further stated that as of November 26, 2012, no domestic or international vendor has met the applicable root mean square error (RMSE) requirement of ASME Section XI Appendix VIII, Supplement 10, or the alternative qualification requirements of ASME Code Case N-695. The licensee noted that the vendors being considered for performing these examinations have demonstrated an RMSE of no less than 0.189 inches.

-4 3.1.5 Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee proposed to use the following alternative for flaw depth sizing when dissimilar metal welds are examined from the inside surface of the pipe: 1. Examinations shall be performed using UT techniques that are qualified for flaw detection and sizing using procedures, personnel and equipment qualified by demonstration in all aspects except depth sizing. 2. A correction factor of the RMS Error -0.125 inches shall be added to the depths of any measured flaws. The correction factor shall be applied to the most critical location on the flaw in relation to surface proximity. 3. Eddy Current (ET) examinations shall be used to confirm whether any detected flaws are surface-breaking. 4. If any inner diameter (10) surface-breaking flaws are detected and measured as 50% through-wall depth or greater, the licensee shall repair the indications or shall perform flaw evaluations and shall submit the evaluations to the NRC for review and approval prior to reactor startup. These flaw evaluations shall include the following: Information concerning the mechanism which caused the flaw. Information concerning the surface roughness/profile in the area of the pipe/weld required to perform the examination, and an estimate of the percentage of potential surface areas with UT probe "lift-off. 3.1.6 Duration of Proposed Alternative The licensee stated that the proposed alternative to the ASME Code is applicable for the remainder of the third 10-year lSI Interval at Catawba Nuclear Station, Unit 2, which began on October 15, 2005, and is currently scheduled to end on August 19, 2016. 3.2 NRC Staff Evaluation The licensee has requested relief from the requirements of ASME Code,Section XI, Appendix VIII, Supplement 10, and Code Case N-695. The NRC has accepted for use Code Case 695 in NRC Regulatory Guide 1.147, Revision 16. Code Case N-695 provides that procedures used to inspect welds from the inside surface of the pipe be qualified by performance demonstration. The acceptance criterion in Code Case N-695 and supplement 10 of ASME Code,Section XI, 1998 Edition through 2000 Addenda specifies that the RMSE of the depth sizing procedures, equipment and personnel shall not be greater than 0.125 inches. The NRC staff has confirmed that since 2002, the industry has not been able to satisfy the RMSE acceptance criterion of less than 0.125 inches to qualify the UT inspection procedures for flaw depth sizing performed from the inside surface of a pipe. The NRC staff finds that this inability to qualify UT techniques from the pipe inside surface in accordance with ASME Code Case N-695 constitutes an impracticality as described in 10 CFR 50.55a(g)(5)(iii).

-5 In July 2012, to address under sizing of flaws by inside surface UT procedures which do not meet the ASME Code Case N-695 acceptance criterion, the NRC staff and personnel from the Performance Demonstration Initiative investigated the proprietary UT examination data set compiled from all attempts to date to qualify inside surface UT inspection procedures to the acceptance criterion contained in ASME Code Case N-695. Based on this investigation the NRC staff concluded that: (a) For flaw depths less than 50 percent pipe wall thickness, a flaw could be appropriately depth sized if a correction factor is added to the measured flaw depth such that the adjusted flaw depth is equal to the measured flaw depth plus the difference between the vendor RMSE and 0.125 inches (procedure qualification RMSE -0.125). This corrected depth would then be used in evaluating the flaw (b) For flaw depths greater than 50 percent pipe wall thickness, the variability of sizing errors was sufficiently large that no single mathematic flaw size adjustment formula was sufficient to provide reasonable assurance of appropriate flaw depth sizing. As a result, the NRC staff finds it necessary to evaluate the flaws that have depth greater than 50 percent through wall on a case-by-case basis. Based on the concerted efforts by the industry to meet the acceptance criteria contained in ASME Code Case N-695 and the difficulties associated with other inspection methods, the NRC staff finds that meeting the 0.125-inch acceptance criterion in ASME Code Case N-695 is impractical and represents a burden to the licensee. Additionally the NRC staff finds that the acceptance criteria for ASME Code Case N-695 need not be met to provide reasonable assurance of structural integrity or leak tightness of the subject components and, therefore, reasonable assurance that the subject components "will not endanger life or property" provided that the following alternative requirements which are imposed by the NRC staff in accordance with 10 CFR 50.55a(g)(6)(i) are met. (1) Examine the welds under consideration using a UT technique which is qualified for flaw detection and length sizing. (2) For flaw(s) with measured depth of less than 50 percent of the wall thickness, the depth shall be adjusted by adding the measured flaw depth to the difference between the procedure qualification RMSE and 0.125 inches (Le., Procedure RMSE -0.125). (3) For flaw(s) with measured depth of greater than 50 percent of the wall thickness, either the degraded weld needs to be repaired in accordance with the ASME Code or, a flaw evaluation needs to be submitted to the NRC for review and approval prior to reactor startup. In addition, the flaw depth analyzed in the flaw evaluation shall also be adjusted by adding the measured flaw depth to the difference between the procedure qualification RMSE and 0.125 inches (Le., Procedure RMSE -0.125). In addition to information normally contained in flaw evaluations in accordance with the ASME Code,Section XI, IWB-3600, the submitted flaw evaluation shall include: (a) information concerning the degradation mechanism which caused the crack, (b) information concerning the surface roughness and/or profile in the area of the examined pipe and/or

-weld, and (c) information concerning areas in which the UT probe may "lift off" from the surface of the pipe and/or weld. Perform eddy current examination(s) to confirm whether a flaw is connected to the inside surface of the pipe and/or weld, if the flaw is located close to the inside surface of the pipe in accordance with the proximity rule of ASME Code,Section XI, IWA-3310. An surface-connected flaw in an Alloy 82/182 dissimilar metal weld in a PWR water environment is susceptible to primary water stress corrosion cracking. The crack growth rate of primary stress corrosion cracking is aggressive and such a degradation mechanism poses significant challenges to the structural integrity of the weld. The NRC staff finds that the proposed alternative is consistent with the NRC staff's position because the licensee will (1) incorporate the correction factor into the flaw depth when evaluating the flaw size, (2) repair the flaw or submit a flaw evaluation to the NRC for review and approval prior to reactor startup if any flaw is measured to be 50 percent through-wall depth or greater, and (3) use the eddy current testing to confirm whether the flaw is connecting to the inside surface. Therefore, the NRC staff determines that the proposed alternative to the RMSE acceptance criterion of ASME Code Case N-695 is acceptable and will not significantly challenge the structural integrity or leak tightness of the subject welds. 4.0 CONCLUSION As set forth above, the NRC staff determines 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. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i), and is in compliance with the ASME Code's requirements, for which the relief was not requested. Therefore, the NRC staff grants the use of RR 13-CN-001 for the duration of the Catawba Unit 2 third 10-year lSI interval, currently scheduled to end on August 19, 2016. All other ASME Code,Section XI and 10 CFR 50.55a requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. Principal Contributor: J. Tsao, NRR Date: July 8, 2013 K. Henderson -2 If you have any questions, please contact the Project Manager, James Kim at 301-415-4125 or via e-mail at james.kim@nrc.gov. Sincerely, IRA! Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. Safety cc w/encl: Distribution via DISTRIBUTION: PUBLIC Branch Reading RidsNrrDorlDpr Resource RidsNrrDorlLpl2-1 Resource RidsNrrDeEpnb Resource RidsAcrsAcnw_MailCTR Resource RidsNrrLASFigueroa Resource RidsNrrPMCatawba Resource (hard copy) RidsOgcMailCenter Resource RidsRgn2MailCenter Resource ADAMS Accession No. ML13176A400 *memo dated June 18, 2013 OFFICE NAME DORULPL2-1/PM JKim DORULPL2-1/LA SFigueroa DElEPNB/BC(A} JHuang'" ======i-1/BC RPascarelli I DATE 6/26/13 6/26/13 6/18/13 7/8/13 OFFICIAL RECORD