ML063060478

Relief Request 13R-04 for the Third 10-Year Interval Inservice Inspection

ML063060478
Person / Time
Site:	Callaway
Issue date:	12/07/2006
From:	Terao D NRC/NRR/ADRO/DORL/LPLIV
To:	Naslund C Union Electric Co
Donohew J N, NRR/DORL/LP4, 415-1307
References
TAC MD1158
Download: ML063060478 (7)

Text

December 7, 2006 Mr. Charles D. Naslund Senior Vice President and Chief Nuclear Officer Union Electric Company Post Office Box 620 Fulton, MO 65251

SUBJECT:

CALLAWAY PLANT, UNIT 1 - RELIEF REQUEST I3R-04 FOR THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION (TAC NO. MD1158)

Dear Mr. Naslund:

By letter dated March 28, 2006 (ULNRC-05271), Union Electric Company submitted three Relief Requests (RRs) for its third 10-year inservice inspection (ISI) program interval at the Callaway Plant, Unit 1 (Callaway). The three RRs are I3R-01, I3R-02, and I3R-04. This letter addresses RR I3R-04. The remaining RRs will be addressed in future letters.

Based on the enclosed safety evaluation, the NRC staff concludes that compliance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements results in a hardship without a compensating increase in the level of quality and safety, and the licensees proposed alternative provides reasonable assurance of the structural integrity of the reactor pressure vessel support structures. Based on this determination, pursuant to 10 CFR 50.55a(a)(3)(ii), the Commission authorizes the proposed alternatives in RR I3R-04 for the third 10-year ISI interval at Callaway. 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.

Sincerely,

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosure:

Safety Evaluation cc w/encl: See next page

ML063060478 OFFICE NRR/LPL4/PM NRR/LPL4/LA LPL4/PM CVIB/BC OGC-NLO subj to edits NRR/LPL4/BC NAME JDonohew LFeizollahi RCureton MMitchell TCampbell DTerao DATE 12/6/06 12/7/06 11/20/06 6/13/06 12/5/06 12/6/06 Callaway Plant, Unit 1 cc:

Professional Nuclear Consulting, Inc. Mr. Rick A. Muench 19041 Raines Drive President and Chief Executive Officer Derwood, MD 20855 Wolf Creek Nuclear Operating Corporation P.O. Box 411 John ONeill, Esq. Burlington, KA 66839 Pillsbury Winthrop Shaw Pittman LLP 2300 N. Street, N.W. Mr. Dan I. Bolef, President Washington, D.C. 20037 Kay Drey, Representative Board of Directors Coalition for the Mr. Keith A. Mills, Supervising Engineer Environment Regional Regulatory Affairs/Safety Analysis 6267 Delmar Boulevard AmerenUE University City, MO 63130 P.O. Box 620 Fulton, MO 65251 Mr. Lee Fritz, Presiding Commissioner Callaway County Court House U.S. Nuclear Regulatory Commission 10 East Fifth Street Resident Inspector Office Fulton, MO 65151 8201 NRC Road Steedman, MO 65077-1302 Mr. David E. Shafer Superintendent, Licensing Mr. Les H. Kanuckel Regulatory Affairs Manager, Quality Assurance AmerenUE AmerenUE P.O. Box 66149, MC 470 P.O. Box 620 St. Louis, MO 63166-6149 Fulton, MO 65251 Manager, Regulatory Affairs Missouri Public Service Commission AmerenUE Governor Office Building P.O. Box 620 200 Madison Street Fulton, MO 65251 Jefferson City, MO 65102-0360 Mr. Keith G. Henke, Planner Regional Administrator, Region IV Division of Community and Public Health U.S. Nuclear Regulatory Commission Office of Emergency Coordination 611 Ryan Plaza Drive, Suite 400 930 Wildwood P.O. Box 570 Arlington, TX 76011-4005 Jefferson City, MO 65102 Mr. H. Floyd Gilzow Certrec Corporation Deputy Director for Policy 4200 South Hulen, Suite 630 Missouri Department of Natural Resources Fort Worth, TX 76109 P.O. Box 176 Jefferson City, MO 65102-0176 Director, Missouri State Emergency Management Agency P.O. Box 116 Jefferson City, MO 65102-0116 June 2006

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST I3R-04 FOR THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By letter dated March 28, 2006 (Agencywide Documents Access and Management System Accession No. ML061010704), Union Electric Company (the licensee) submitted three Relief Requests (RRs) for its third 10-year inservice inspection (ISI) program interval at the Callaway Plant, Unit 1 (Callaway). The three RRs are I3R-01, I3R-02, and I3R-04. This Safety Evaluation (SE) only addresses RR I3R-04.

2.0 REGULATORY EVALUATION

Inservice inspection of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, 2, and 3 components is 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) 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). Paragraph 50.55a(a)(3) of 10 CFR states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (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) 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. 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 in 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 code of record for the Callaway third 10-year interval ISI program, which began on December 19, 2005, is the 1998 Edition through the 2000 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code).

3.0 TECHNICAL EVALUATION

ASME Code Components The code components are the Reactor Pressure Vessel Supports (RPV), Component Numbers 2-RBB01-01, 2-RBB01-02, 2-RBB01-03, and 2-RBB01-04.

Applicable ASME Code Requirement The ASME Code,Section XI, Table IWF-2500-1, Examination Category F-A, Item Number F1.40 requires 100 percent of Class 1 supports, other than piping supports, be subject to a VT-3 visual examination once every inspection interval.

Licensees Proposed Alternative Examination In lieu of implementing the requirements of ASME Code,Section XI, Table IWF-2500-1, Category F-A, Item No. F1.40, the licensee proposes to perform a limited VT-3 visual examination, with the walk plate and insulation installed, on the accessible NF portions of the Reactor Vessel support assemblies. If conditions are discovered during this limited VT-3 examination that do not meet the acceptance standards of IWF-3400, the walk plate or insulation will, if necessary, be removed in order to meet the requirements of IWF-3122.2 or IWF-3122.3, as applicable.

Licensees Basis for Relief Request (As stated in the licensees March 28, 2006, submittal)

Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested on the basis that compliance with the specified requirements is impractical. Conformance with the applicable inservice inspection requirements would necessitate a design modification to the reactor pressure vessel supports and associated insulation/walkplate to allow 100% visual examination of the subject supports. In addition, limited accessibility and high radiation levels in the area where these supports are located further reduces the percentage of the supports available for visual examination.

The Callaway reactor vessel is supported by two cold leg nozzles and two hot leg nozzles. There is a support assembly at each of these nozzles that consists of a nozzle weld build-up, shoe plate, air-cooled box, and steel support structure embedded in the primary shield wall. Figures 1 and 2 [Figures 1 and 2 are included in the licensees letter dated March 28, 2006] depict these support assemblies. As shown in these figures, only the nozzle weld build-up and shoe plate are completely accessible for a visual VT-3 examination. Most of the air-cooled box and the entire steel support structure are located beneath a steel walk plate, and only the top of the air-cooled box is directly accessible. An additional 20 to 30 percent of the air-cooled box and a very small percentage of the steel support structure would be made accessible if the steel walk plate and insulation were removed.

The reactor vessel supports are located in a confined space below the refueling pool permanent seal ring. The area can only be accessed through four seal ring hatches. In addition to difficult access, the radiation level in the area is between 1.5 and 2.0 man-rem per hour.

The large cost1 of a design modification to the reactor pressure vessel supports and associated insulation/walkplate to allow 100% visual examination of the subject supports is deemed an undue burden. Further, it is estimated that the removal and re-installation of the walk plate and insulation in this confined space, combined with the performance of the visual VT-3 examination, would result in an exposure of approximately 36 man-rem. Removal of the walk plate and insulation, under these conditions, in order to increase the examination coverage of the air-cooled box by approximately 20 to 30 percent and a very small percentage of the steel support structure is considered impractical without a commensurate increase in quality or safety.

NRC Staff Evaluation

The ASME Code of record for Callaway in this inspection interval requires that 100 percent of Class 1 supports, other than piping supports, be subject to a visual, VT-3 examination once every inspection interval. As an alternative to the ASME Code requirements, the license proposes to perform a limited VT-3 visual examination, with the walk plate and insulation installed, on the accessible portions of the RPV support assemblies. In addition, the licensee proposed that if conditions are discovered during this limited VT-3 examination that do not meet the acceptance standards of IWF-3400, the walk plate or insulation will, if necessary, be removed in order to meet the requirements of IWF-3122.2 or IWF-3122.3, as applicable.

As described by the licensee in its March 28, 2006, submittal, limited accessibility and high-radiation levels in the area where the subject supports are located reduces the percentage of the supports available for visual examination. According to Figures 1 and 2 of the licensee's submittal, the RPV is supported by two cold-leg nozzles and two hot-leg nozzles. In addition, there is a support assembly at each of these nozzles that consists of a nozzle weld build-up, shoe plate, air-cooled box, and steel support structure embedded in the primary shield wall.

The licensee noted that the nozzle weld build-up and shoe plate are accessible for a visual VT-3 examination. The air-cooled box steel support structure is located beneath a steel walk plate and only the top of the air-cooled box is accessible to perform a VT-3 visual examination.

If the steel walk plate and insulation were removed, only an additional 20 to 30 percent of the air-cooled box and a small percentage of the steel support structure would be made accessible for examination.

Furthermore, the subject RPV supports are located in a confined space that is below the refueling pool permanent seal ring. This area is only accessible through four seal-ring hatches and access in this area would cause the licensee's personnel to be exposed to a radiation level between 1.5 and 2.0 man-rem per hour. The licensee estimated that the removal and reinstallation of the walk plate and insulation combined with the performance of the visual VT-3 examination, would result in an exposure of approximately 36 man-rem.

1. Cost of a modification is not considered a basis for relief either by 10 CFR 50.55a(a)(3)(i),

10 CFR 50.55a(a)(3)(ii), or 10 CFR 50.55a(g)(5)(iii).

In its submittal, requesting relief pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee had determined that the ASME Code requirements are impractical for its facility because a design modification was needed to allow 100 percent visual examination of the subject reactor vessel supports. The licensee showed that the design modification was needed because the radiation levels in the area are high and the access to the supports is difficult.

Based on the radiation exposure and the difficulty in obtaining access to the RPV support area, the NRC staff concludes that compliance with the ASME Code requirements results in a hardship without a compensating increase in the level of quality and safety. The NRC staff further concludes that the alternative provides reasonable assurance of the structural integrity of the RPV support structures, based on the licensees proposed alternative in RR I3R-04.

4.0 CONCLUSION

For RR I3R-04, the NRC staff concludes that compliance with the ASME Code requirements results in a hardship without a compensating increase in the level of quality and safety, and the licensees proposed alternative provides reasonable assurance of the structural integrity of the RPV support structures. Therefore, the NRC staff concludes that the licensees proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the Callaway third 10-year ISI interval. 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.

Principal Contributor: Thomas K. McLellan Date: December 7, 2006