ML041890203

Code Relief, Evaluation of Relief Request No. 20, Revision 1 - for the Third 10-Year Inservice Inspection Interval

ML041890203
Person / Time
Site:	Prairie Island
Issue date:	07/27/2004
From:	Mahesh Chawla NRC/NRR/DLPM/LPD3
To:	Solymossy J Nuclear Management Co
Chawla M, NRR/DLPM, 415-8371
References
TAC MC2545, TAC MC2546
Download: ML041890203 (8)

Text

July 27, 2004 Mr. Joseph M. Solymossy Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -

EVALUATION OF RELIEF REQUEST NO. 20, REVISION 1, FOR THE THIRD 10-YEAR INSERVICE INSPECTION (ISI) INTERVAL (TAC NOS. MC2545 AND MC2546)

Dear Mr. Solymossy:

By a letter dated March 30, 2004, as supplemented June 1, 2004, Nuclear Management Company, LLC (the licensee), submitted Relief Request (RR) No. 20, Revision 1, for Prairie Island Units 1 and 2. This request was to use 0.189" root mean square error (RMSE) in lieu of 0.125" required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, Appendix VIII, Supplement 10 for depth sizing of flaws in dissimilar metal weld test specimens during qualification of ultrasonic examination procedure, equipment and personnel.

A request for additional information was sent to NMC on April 19, 2004, to which NMC replied via e-mail on April 23, 2004 (ADAMS Accession Number ML041820031). U.S. Nuclear Regulatory Commission (NRC) staff needed further clarification on the relief request and therefore another request for additional information was sent to the licensee on April 28, 2004 (ADAMS Accession Number ML041800463). During a telephone conference on May 20, 2004, between the NRC staff and the NMC representatives, NMC agreed to revise the relief request which resulted in the submittal of the RR-20, Revision 1, dated June 1, 2004.

The NRC staff has evaluated the licensees request for relief and has authorized the proposed alternative pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section, 50.55a(g)(6)(i). At this time, achieving the 0.125 inch RMSE appears to be impractical. The vendor contracted by the licensee has only been able to achieve an accuracy of 0.189 inch RMSE. The licensee has proposed to use 0.189 inch RMSE to size any detected flaws during the forthcoming outage. The proposed inspection provides reasonable assurance of structural integrity.

J. Solymossy 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 licensees proposed alternative is authorized for Prairie Island Units 1 and 2 for the remainder of the third 10-year ISI interval which ends on December 20, 2004. The NRC staff review and evaluation is contained in the enclosed safety evaluation.

Sincerely,

/RA/

Mahesh L. Chawla, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosure:

Safety Evaluation cc w/encl: See next page

ML041890203 OFFICE PDIII-1/PM PDIII-1/LA EMCB/SC OGC PDIII-1/SC NAME MChawla THarris TChan AHodgdon LRaghavan DATE 07/12/04 07/9/04 07/12/04 07/20/04 07/27/04 Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:

Jonathan Rogoff, Esquire Tribal Council Vice President, Counsel & Secretary Prairie Island Indian Community Nuclear Management Company, LLC ATTN: Environmental Department 700 First Street 5636 Sturgeon Lake Road Hudson, WI 54016 Welch, MN 55089 Manager, Regulatory Affairs Nuclear Asset Manager Prairie Island Nuclear Generating Plant Xcel Energy, Inc.

Nuclear Management Company, LLC 414 Nicollet Mall, R.S. 8 1717 Wakonade Drive East Minneapolis, MN 55401 Welch, MN 55089 John Paul Cowan Manager - Environmental Protection Division Executive Vice President & Chief Nuclear Minnesota Attorney Generals Office Officer 445 Minnesota St., Suite 900 Nuclear Management Company, LLC St. Paul, MN 55101-2127 700 First Street Hudson, WI 54016 U.S. Nuclear Regulatory Commission Resident Inspector's Office Craig G. Anderson 1719 Wakonade Drive East Senior Vice President, Group Operations Welch, MN 55089-9642 Nuclear Management Company, LLC 700 First Street Regional Administrator, Region III Hudson, WI 54016 U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Administrator Goodhue County Courthouse Box 408 Red Wing, MN 55066-0408 Commissioner Minnesota Department of Commerce 121 Seventh Place East Suite 200 St. Paul, MN 55101-2145 November 2003

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION INTERVAL PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 REQUEST FOR RELIEF NO. 20, REVISION 1 NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By letter dated March 30, 2004, as supplemented June 1, 2004, Nuclear Management Company, LLC submitted a request for relief (RR-20, Revision 1) from the requirement specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, Appendix VIII, Supplement 10 to use 0.189" root mean square error (RMSE) in lieu of 0.125" required by the Code for depth sizing of flaws in dissimilar metal weld test specimens during qualification of ultrasonic examination procedure, equipment and personnel. The U.S. Nuclear Regulatory Commission (the Commission or the NRC) staff has evaluated the licensees request for relief pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g)(6)(i).

2.0 REGULATORY EVALUATION

The inservice inspection of ASME Code Class 1, Class 2, and Class 3 components is performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR Section 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except for the design and access provisions and the preservice 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. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the

requirements of the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The inservice inspection (ISI) code of record for Prairie Island Nuclear Generating Plant third 10-year ISI interval is the 1989 Edition which ends December 20, 2004. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.

3.0 TECHNICAL EVALUATION

3.1 Systems/Components for Which Relief is Requested Pressure retaining dissimilar metal piping welds subject to examinations using procedures, personnel, and equipment qualified to the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplement 10, Qualification Requirements for Dissimilar Metal Piping Welds for the remainder of the third 10-year ISI interval.

Code Class: Class 1

Reference:

ASME Section XI, 1989 Edition with no Addenda Examination Category: B-F Item Number: B5.10 Components: Reactor Vessel Nozzle to Safe End Welds identified as W-5, W-7 in Unit 1 and W-5, W-11 in Unit 2 in the June 1, 2004 submittal.

3.2 Code Requirements A volumetric examination of dissimilar metal pressure retaining piping welds is required per the 1989 Edition ASME Section XI Code, Table IWB-2500-1, Examination Category B-F, Item No.

B5.10. In addition, 10 CFR 50.55a requires that these examinations be performed using procedure, personnel, and equipment qualified in accordance with the requirements of the ASME Code,Section XI, 1995 Edition, 1996 Addenda, Appendix VIII, Supplement 10.

Paragraph 3.2(b).

3.3 Licensees Proposed Alternative The inspection vendor to be employed by Prairie Island Nuclear Generating Plant for the remote automated Reactor Vessel examinations did not achieve the 0.125" RMSE Appendix VIII, Supplement 10 acceptance tolerance during procedure qualification. Prairie Island Nuclear Generating Plant proposes using the vendor achieved sizing error of 0.189" RMSE.

The proposed procedure to address sizing of the flaws that may be found during the examination is to add to the measured flaw size the difference between the achieved sizing error and the 0.125" RMSE Appendix VIII Supplement 10 acceptance criteria.

3.4 Licensees Basis for Requesting Relief On September 17, 2003, the Nuclear Management Company, submitted a relief request as alternatives to paragraphs 1.1(b), 1.1(d), 1.1(d)(1), 1.2(b), 1.2(c)(1), 1.3(c), 2.0, 2.2(b), 2.2(c),

2.3(a), and 2.3(b) of the ASME Code,Section XI, Appendix VIII, Supplement 10. On February 26, 2004, the NRC authorized the licensees proposed alternatives.

The 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplement 10, Paragraph 3.2(b), states that the examination procedure, equipment, and personnel are qualified for depth sizing when the RMSE of the flaw depth measurements, as compared to the true flaw depths, is less than or equal to 0.125. An RMSE of 0.189" is based on actual vendor demonstrated in-process field qualifications and is the optimum value that could be achieved. Prairie Island Nuclear Generating Plant believes that the use of 0.189" RMSE as an adjustment to the measured flaw will ensure a conservative bounding flaw depth value. This issue was not addressed in the licensees September 17, 2003 submittal and this was not authorized in the NRC staffs safety evaluation on February 26, 2004.

4.0 STAFF EVALUATION Supplement 10 of Appendix VIII to the ASME Code,Section XI requires that examination procedures, equipment, and personnel meet specific criteria for flaw depth sizing accuracy.

The Code specifies that the maximum error of flaw depth measurements, as compared to the true flaw depths, must be less than or equal to 0.125 inch RMSE. The industry is in the process of qualifying personnel to Supplement 10 as implemented by the PDI (Performance Demonstration Initiative) program. However, for demonstrations performed from the inside surface of a pipe weldment, personnel have been unsuccessful at achieving the 0.125 inch RMSE depth sizing criterion. At this time, achieving the 0.125 inch RMSE appears to be impractical. The vendor contracted by the licensee has only been capable of achieving an accuracy of 0.189 inch RMSE. The licensee has proposed to use 0.189 inch RMSE to size any detected flaws during the forthcoming outage. The licensee would add the difference (0.064 inch) between the Code required RMSE (0.125 inch) and the demonstrated accuracy (0.189 inch RMSE) to the measurements acquired from flaw sizing. The request is for the remainder of the cycle in the third 10-year ISI interval.

From performance demonstration of typical hot leg and cold leg weld examinations with a wall thickness of 2.5 inches, the staff gathered information which suggested that RMSE values are independent of flaw depth. In the thickness range of test specimens, 0.125 inch RMSE of the flaw depth measurement would be approximately 5 percent tolerance on root mean square percent of the typical wall thickness and, likewise, 0.189 inch RMSE would translate to approximately 7.5 percent of the root mean square percent of the typical wall thickness. The increase in error of 2.5 percent of the measured flaw depth is less than the planar flaw acceptance criteria in Table IWB-3514-2. The staff believes that the flaw depth adjustment proposed by the licensee will ensure a conservative bounding flaw depth value and provides reasonable assurance of structural integrity.

5.0 CONCLUSION

Based on the above evaluation, the staff has determined that achieving the 0.125 inch RMSE for depth sizing of flaws in dissimilar metal weld test specimens during qualification of ultrasonic examination procedure, equipment and personnel for the subject welds is impractical at this time. Granting relief pursuant to 10CFR50.55a(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. Moreover, the proposed inspection provides reasonable assurance of structural integrity. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the licensees proposed alternative is authorized for Prairie Island Units 1 and 2 for the remainder of the third 10-year ISI interval which ends on December 20, 2004. All other ASME Code,Section XI 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.