ML063180250

Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Vessel Code, Relief Request No. RI-06, Revision 3, for Weld HMC-BB-1

ML063180250
Person / Time
Site:	Cooper
Issue date:	12/05/2006
From:	Terao D NRC/NRR/ADRO/DORL/LPLIV
To:	Edington R Nebraska Public Power District (NPPD)
Vaidya B, NRR/DORL/LP4, 415-3308
References
TAC MD2316
Download: ML063180250 (10)

Text

December 5, 2006 Mr. Randall K. Edington Vice President-Nuclear and CNO Nebraska Public Power District P.O. Box 98 Brownville, NE 68321

SUBJECT:

COOPER NUCLEAR STATION RE: REQUEST FOR RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND VESSEL CODE (CODE), RELIEF REQUEST NO. RI-06, REVISION 3, FOR WELD HMC-BB-1 (TAC NO. MD2316)

Dear Mr. Edington:

By letter dated May 30, 2006, Nebraska Public Power District (the licensee) submitted Relief Request No. RI-06, Revision 3, to use alternate examination for weld HMC-BB-1, for the third 10-year inservice inspection (ISI) interval at Cooper Nuclear Station (CNS).

The Nuclear Regulatory Commission (NRC) staff has completed its review of Relief Request No. RI-06, Revision 3, for weld HMC-BB-1, and the safety evaluation is enclosed.

Based on the information provided, the NRC staff has concluded that it is impractical for the licensee to comply with the examination requirements for RPV shell welds specified in Item B1.10 of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, in Table IWB-2500-1 of subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Code, subject to conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(3) and (4), and that the proposed examination provides reasonable assurance of pressure boundary integrity. The alternate examination proposed in Relief Request No. RI-06, Revision 3, for weld HMC-BB-1, for the third 10-year ISI interval at CNS, will provide reasonable assurance of the structural integrity of the subject weld. Therefore, for Relief Request RI-06, Revision 3, relief is granted pursuant to paragraph 50.55a(g)(6)(i) of Title 10 of the Code of Federal Regulations (10 CFR) for the third ISI interval at CNS. The NRC staff has also determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) for the third 10-year ISI interval at CNS 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.

R. K. Edington 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 reviews by the authorized Nuclear Inservice Inspector.

Please contact Bhalchandra K. Vaidya at (301)-415-3308, if you have any questions.

Sincerely,

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosure:

Safety Evaluation cc w/encl: See next page

R. K. Edington 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 reviews by the authorized Nuclear Inservice Inspector.

Please contact Bhalchandra K. Vaidya at (301)-415-3308, if you have any questions.

Sincerely,

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosure:

Safety Evaluation cc w/encl: See next page DISTRIBUTION:

PUBLIC LPLIV R/F RidsAcrsAcnwMailCenter RidsRgn4MailCenter RidsNrrDorl (CHaney/JLubinski) RidsOgcRp JLamb, EDO RIV RidsNrrDorlLpl4 (DTerao) RidsNrrPMBVaidya RidsNrrLALFeizollahi CFairbanks, DCI/CVIB RidsNrrDciCvib (MMichell)

Accession No. ML063180250 OFFICE LPL4/PM LPL4/LA DCI/CVIB OGC -NLO LPL4/BC NAME BVaidya LFeizollahi MMitchell TCampbell DTerao DATE 11/17/06 11/17/06 11/8/06 12/5/06 12/5/06 OFFICIAL RECORD COPY

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. RI-06, REVISION 3, FOR WELD HMC-BB-1 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Class 1 components requires that inservice inspections (ISI) be performed in accordance with the applicable edition of Section XI of the ASME Code and any relevant 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 Commission pursuant to 10 CFR 50.55a(g)(6)(i). It is stated in 10 CFR 50.55a(a)(3) that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (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 components shall 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 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 the 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.

2.0 RELIEF REQUEST NO. RI-06, REVISION 3 By letter dated May 30, 2006, the Nebraska Public Power District (NPPD, licensee) requested relief to change the NRC allowable ISI examination coverage in the lower reactor vessel circumferential weld HMC-BB-1 from 86 percent total composite coverage to 75 percent total composite coverage.

2.1 Code Requirements for which Relaxation is Requested The ISI code of record for the Cooper Nuclear Station (CNS) third 10-year ISI interval is the 1989 Edition of Section XI of the ASME Code. Relief Request No. RI-06, Revision 3, updates the information for the examination coverage of weld HMC-BB-1, contained in Relief Request No. RI-06, Revision 2 (dated April 24, 2001), which addressed Table IWB-2500-1, Category B-A, Item Nos. B1.11, B1.12, B1.21, B1.22 and B1.30 for volumetric examination requirements of the ASME Code for the reactor pressure vessel shell and head. Relief Request No. RI-06, Revision 2, was approved by the NRC staff by letter dated November 30, 2001.

2.2 Licensees Proposed Alternative Examination (As Stated)

In accordance with 10 CFR 50.55a(g)(5)(iii), CNS proposes to examine the accessible portions of the reactor vessel welds in lieu of the impractical [ASME Code-required examinations.

Using the provisions of this relief request as an alternative to the specific requirements of ASME Table IWB-2500-1, identified above [in the licensees May 30, 2006, submittal], will continue to provide reasonable assurance of structural integrity since the percent of examination coverage already obtained would have identified any pattern of degradation should one develop. Therefore, pursuant to 10 CFR 50.55a, Codes and Standards, Paragraph (a)(3), NPPD requests relief from the specific IWB requirements identified in this request.

Based on the above, the proposed alternative inspection will continue to provide an acceptable level of quality and safety.

2.3 Licensees Basis for Proposed Alternative (As Stated)

The Cooper Nuclear Station construction permit was issued before the effective date of implementation for ASME Section XI and thus the plant was not designed to meet requirements of inservice inspection; therefore, 100% compliance is not feasible or practicable.

The CRD [control rod drive] and instrument penetrations prevent direct access to most of the bottom head. Circumferential weld HMD-BB-1 is located inside the skirt and is inaccessible for examination. Portions of the Bottom Head Meridional welds, HMB-BB-1, HMB-BB-2, HMB-BB-3, HMB-BB-4, HMB-BB-5, and HMB-BB-6 are located inside the vessel skirt and are inaccessible for examination. Access to weld HMC-BB-1 is limited due to the proximity of the vessel skirt. The configuration limits scanning with the 60 degree probe. The total composite coverage achieved for HMC-BB-1 in the third ISI inspection interval was 75% with no recordable indications.

Access to the reactor vessel shell welds from the exterior is limited. Below the top of the biological shield, most of the reactor vessel is insulated with permanent reflective insulation and surrounded by a concrete biological shield.

Penetrations through the biological shield provide limited access to some welds.

The annular space between the inside diameter of the insulation and the outside diameter of the reactor vessel is a nominal 2 inches. There is no working space

to remove the insulation panels from the vessel, which precludes both direct and remote examination [from] the outside surface.

In accordance with 10 CFR 50.55a(g)(6)(ii)(A), an examination of the Reactor Vessel shell welds was performed during RFO-18 [Refueling Outage-18] using PDI [Performance Demonstration Initiative] qualified procedures (see Relief Request RI-04) and the GERIS 2000 ID [inside diameter] Scanner.

Supplemental manual examinations were performed to the extent practical.

Weld coverage is identified in the attached table1.

Proposed Relief Request RI-06, Revision 3, is a request to change the NRC allowable examination in lower reactor vessel circumferential weld HMC-BB-1 from 86% total composite coverage to 75% coverage. The only weld affected by this relief request revision is HMC-BB-1. All other welds listed in this request met the NRC allowable examination coverage for the third ten-year interval and are not affected by this revision.

In addition to the physical limitation due to the proximity of the weld to the vessel skirt, the reduction in total composite coverage (i.e., 75% as opposed to previous 86%) is believed to be due to difference in procedure and equipment qualifications through the Performance Demonstration Initiative (PDI) as compared to previous requirements. Limitations [on] qualified transducer sizes through PDI contributed to the reduction of coverage compared to the previous examination of HMC-BB-1. This weld was examined in 1993 and again in 1995 using 0-, 45- and 60-degree transducers. In 2001, the examination for the third interval was performed in accordance with ASME Section XI, 1995 Edition, 1996 Addenda, Appendix VIII, using only a 60-degree transducer. This examination achieved 75% total composite coverage . . . However, based on the demonstrated qualification of the examiners, a more reliable examination was performed even though less coverage was achieved.

Since 10 CFR 50.55a requires use of ASME [Section] XI Appendix VIII in detecting flaws in the welds at CNS, only qualified PDI procedures and specific qualified transducers may be used for the examinations. Previous examinations

[of] the weld may have reported additional coverage, but since the previous examiners and equipment were not qualified by PDI, the 86% coverage reported during the second interval cannot be considered as reliable as the 75% achieved in the third interval. The PDI examination methods maximize the coverage that can be reliably obtained, even though PDI-qualified transducers and associated procedures limit how much coverage an examiner is able to achieve.

1 The attached table refers to a table included in the licensees May 30, 2006, submittal.

The table is not included in this safety evaluation.

3.0 STAFF EVALUATION Paragraph 50.55a(g)(6)(ii)(A)(2) of 10 CFR requires all licensees to augment their reactor pressure vessel (RPV) examinations by implementing once, as part of the ISI interval in effect on September 8, 1992, the examination requirements for RPV shell welds specified in Item B1.10 of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, in Table IWB-2500-1 of subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Code, subject to conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(3) and (4). The licensee is requesting NRC staff approval for changing the NRC allowable examination coverage for lower reactor vessel circumferential weld HMC-BB-1, which is in the above mentioned Examination Category, from 86 percent total composite coverage to 75 percent total composite coverage.

Weld HMC-BB-1 is the only weld affected by the relief request. NRC staff previously approved 86 percent total composite coverage for weld HMC-BB-1 in a safety evaluation dated September 26, 2001 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML012700361). All other welds met the NRC allowable examination coverage for the third 10-year ISI interval.

Although the applicable ASME Code for the CNS third 10-year interval was the 1989 Edition, the examination of weld HMC-BB-1 was performed in accordance with the ASME Code,Section XI, 1995 Edition, 1996 Addenda, Appendix VIII, due to implementation of the PDI requirements which allow only qualified PDI procedures and specific qualified transducers to be used for performing the examination. The change from the 86 percent coverage reported in the second 10-year interval to the 75 percent coverage reported in the third 10-year interval may have been affected by the fact that previous examiners and equipment were not qualified by PDI. Results from the third 10-year interval, using PDI-qualified procedures and equipment, are considered to be more reliable than the previous results, which did not use PDI-qualified procedures and equipment. The third 10-year examination of weld HMC-BB-1 found no recordable indications.

Weld HMC-BB-1 is a lower reactor vessel circumferential weld. Full inspection coverage is not achievable for weld HMC-BB-1 due to the proximity of the vessel skirt. The configuration limits scanning with the 60 degree probe. Additionally, performance of remote, automated ultrasonic testing with GERIS 2000 ID equipment is a difficult and intensive examination. Previously, the NRC staff concluded that there was sufficient coverage of the other RPV welds such that 86 percent total composite coverage for weld HMC-BB-1 was found to be acceptable in the safety evaluation dated September 26, 2001.

Based on the percentage of RPV weld volume examined, including the change in coverage for weld HMC-BB-1 from 86 percent to 75 percent coverage, the NRC staff finds that any patterns of degradation would be detected and the licensee has performed the examination to the extent practical. The change in coverage for weld HMC-BB-1 from 86 percent composite coverage to 75 percent composite coverage may be attributed to the use of PDI-qualified procedures and transducers for the third interval.

4.0 CONCLUSION

Based on the evaluation above, the NRC staff has concluded that it is impractical for the licensee to comply with the requirements and the examinations performed provide reasonable assurance of the structural integrity of the subject weld. Therefore, for Request for

Relief RI-06, Revision 3, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i) for the third ISI interval at the CNS.

The NRC staff has also determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) for the third 10-year ISI interval at CNS 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 ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party reviews by the authorized Nuclear Inservice Inspector.

Principal Contributor: C. J. Fairbanks, DCI/CVIB Date: December 5, 2006

Cooper Nuclear Station cc:

Mr. Ronald D. Asche Mr. H. Floyd Gilzow President and Chief Executive Officer Deputy Director for Policy Nebraska Public Power District Missouri Department of Natural Resources 1414 15th Street P.O. Box 176 Columbus, NE 68601 Jefferson City, MO 65102-0176 Mr. Gene Mace Senior Resident Inspector Nuclear Asset Manager U.S. Nuclear Regulatory Commission Nebraska Public Power District P.O. Box 218 P.O. Box 98 Brownville, NE 68321 Brownville, NE 68321 Regional Administrator, Region IV Mr. John C. McClure U.S. Nuclear Regulatory Commission Vice President and General Counsel 611 Ryan Plaza Drive, Suite 400 Nebraska Public Power District Arlington, TX 76011 P.O. Box 499 Columbus, NE 68602-0499 Director, Missouri State Emergency Management Agency Mr. Paul V. Fleming P.O. Box 116 Licensing Manager Jefferson City, MO 65102-0116 Nebraska Public Power District P.O. Box 98 Chief, Radiation and Asbestos Brownville, NE 68321 Control Section Kansas Department of Health Mr. Michael J. Linder, Director and Environment Nebraska Department of Environmental Bureau of Air and Radiation Quality 1000 SW Jackson P.O. Box 98922 Suite 310 Lincoln, NE 68509-8922 Topeka, KS 66612-1366 Chairman Mr. Donald A. Flater Nemaha County Board of Commissioners Radiation Control Program Director Nemaha County Courthouse Bureau of Radiological Health 1824 N Street Iowa Department of Public Health Auburn, NE 68305 Lucas State Office Building, 5th Floor 321 East 12th Street Ms. Julia Schmitt, Manager Des Moines, IA 50319 Radiation Control Program Nebraska Health & Human Services R & L Mr. Daniel K. McGhee Public Health Assurance Bureau of Radiological Health 301 Centennial Mall, South Iowa Department of Public Health P.O. Box 95007 Lucas State Office Building, 5th Floor Lincoln, NE 68509-5007 321 East 12th Street Des Moines, IA 50319 October 2006

Cooper Nuclear Station cc:

Mr. Keith G. Henke, Planner Mr. John F. McCann, Director Division of Community and Public Health Licensing, Entergy Nuclear Northeast Office of Emergency Coordination Entergy Nuclear Operations, Inc.

930 Wildwood P.O. Box 570 440 Hamilton Avenue Jefferson City, MO 65102 White Plains, NY 10601-1813 Jerry C. Roberts, Director of Nuclear Safety Assurance Nebraska Public Power District P.O. Box 98 Brownville, NE 68321 October 2006