ML042520392

Relief, Requirements of American Society of Mechanical Engineers Boiler & Pressure Vessel Code,Section III, 1965 Edition, & Section XI, 1989 Edition, for Repair & Inspection of Reactor Pressure Vessel Head Penetrations

ML042520392
Person / Time
Site:	Indian Point
Issue date:	10/05/2004
From:	Richard Laufer NRC/NRR/DLPM/LPD1
To:	Kansler M Entergy Nuclear Operations
Milano P, NRR/DLPM , 415-1457
References
TAC MC3281, TAC MC3282
Download: ML042520392 (10)

Text

October 5, 2004 Mr. Michael R. Kansler, President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

RELIEF REQUEST NOS. 62 AND 3-32, INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND NO. 3 (TAC NOS. MC3281AND MC3282)

Dear Mr. Kansler:

By letter dated May 25, 2004, as supplemented on August 9, 2004, Entergy Nuclear Operations, Inc. (the licensee), requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section III, 1965 Edition, and Section XI, 1989 Edition, for the repair and inspection of reactor pressure vessel head penetrations for the Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and 3). In this regard, the licensee proposed alternatives for design, implementation of repairs, and inspections of the head penetrations.

The Nuclear Regulatory Commission (NRC) has reviewed the proposed alternatives in the subject relief requests. The results are provided in the enclosed safety evaluation.

The NRC staff has concluded that the proposed alternatives to the ASME Code requirements in RR Nos. 62 and 3-32 provides an acceptable level of quality and safety and are acceptable.

Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternatives are authorized for the remainder of the third 10-year inservice inspection interval, which is until April 3, 2006, for IP2, and until July 20, 2009, for IP3.

If you have any questions regarding this approval, please contact the IP2 and IP3 Project Manager, Patrick Milano, at 301-415-1457.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Reactor Regulation Docket Nos. 50-247 and 50-286

Enclosure:

Safety Evaluation cc w/encl: See next page

ML042520392 *No substantive changes made OFFICE PDI-1/PM PDI-1/LA EMCB OGC PDI-1/SC NAME PMilano SLittle SE dtd HMcGurren RLaufer DATE 09/08/04 09/10/04 08/18/04 09/14/04 10/01/04 Indian Point Nuclear Generating Unit Nos. 2 & 3 cc:

Mr. Gary Taylor Ms. Charlene Faison Chief Executive Officer Manager, Licensing Entergy Operations, Inc. Entergy Nuclear Operations, Inc.

1340 Echelon Parkway 440 Hamilton Avenue Jackson, MS 39213 White Plains, NY 10601 Mr. John Herron Director of Oversight Senior Vice President and Entergy Nuclear Operations, Inc.

Chief Operating Officer 440 Hamilton Avenue Entergy Nuclear Operations, Inc. White Plains, NY 10601 440 Hamilton Avenue White Plains, NY 10601 Mr. James Comiotes Director, Nuclear Safety Assurance Mr. Fred Dacimo Entergy Nuclear Operations, Inc.

Vice President, Operations Indian Point Energy Center Entergy Nuclear Operations, Inc. 295 Broadway, Suite 2 Indian Point Energy Center P.O. Box 249 295 Broadway, Suite 2 Buchanan, NY 10511-0249 P.O. Box 249 Buchanan, NY 10511-0249 Mr. Patric Conroy Manager, Licensing Mr. Christopher Schwarz Entergy Nuclear Operations, Inc.

General Manager, Plant Operations Indian Point Energy Center Entergy Nuclear Operations, Inc. 295 Broadway, Suite 2 Indian Point Energy Center P. O. Box 249 295 Broadway, Suite 2 Buchanan, NY 10511-0249 P.O. Box 249 Buchanan, NY 10511-0249 Mr. John M. Fulton Assistant General Counsel Mr. Dan Pace Entergy Nuclear Operations, Inc.

Vice President Engineering 440 Hamilton Avenue Entergy Nuclear Operations, Inc. White Plains, NY 10601 440 Hamilton Avenue White Plains, NY 10601 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. Randall Edington 475 Allendale Road Vice President Operations Support King of Prussia, PA 19406 Entergy Nuclear Operations, Inc.

440 Hamilton Avenue Senior Resident Inspector, Indian Point 2 White Plains, NY 10601 U. S. Nuclear Regulatory Commission 295 Broadway, Suite 1 Mr. John McCann P.O. Box 38 Director, Nuclear Safety Assurance Buchanan, NY 10511-0038 Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

Indian Point Nuclear Generating Unit Nos. 2 & 3 cc:

Senior Resident Inspector, Indian Point 3 Mr. William DiProfio U. S. Nuclear Regulatory Commission PWR SRC Consultant 295 Broadway, Suite 1 139 Depot Road P.O. Box 337 East Kingston, NH 03827 Buchanan, NY 10511-0337 Mr. Dan C. Poole Mr. Peter R. Smith, President PWR SRC Consultant New York State Energy, Research, and 20 Captains Cove Road Development Authority Inglis, FL 34449 Corporate Plaza West 286 Washington Avenue Extension Mr. William T. Russell Albany, NY 12203-6399 PWR SRC Consultant 400 Plantation Lane Mr. Paul Eddy Stevensville, MD 21666-3232 Electric Division New York State Department Mr. Alex Matthiessen of Public Service Executive Director 3 Empire State Plaza, 10th Floor Riverkeeper, Inc.

Albany, NY 12223 25 Wing & Wing Garrison, NY 10524 Mr. Charles Donaldson, Esquire Assistant Attorney General Mr. Paul Leventhal New York Department of Law The Nuclear Control Institute 120 Broadway 1000 Connecticut Avenue NW New York, NY 10271 Suite 410 Washington, DC, 20036 Mayor, Village of Buchanan 236 Tate Avenue Mr. Karl Coplan Buchanan, NY 10511 Pace Environmental Litigation Clinic 78 No. Broadway Mr. Ray Albanese White Plains, NY 10603 Executive Chair Four County Nuclear Safety Committee Mr. Jim Riccio Westchester County Fire Training Center Greenpeace 4 Dana Road 702 H Street, NW Valhalla, NY 10592 Suite 300 Washington, DC 20001 Ms. Stacey Lousteau Treasury Department Entergy Services, Inc.

639 Loyola Avenue Mail Stop: L-ENT-15E New Orleans, LA 70113

Indian Point Nuclear Generating Unit Nos. 2 & 3 cc:

Mr. Robert D. Snook Assistant Attorney General State of Connecticut 55 Elm Street P.O. Box 120 Hartford, CT 06141-0120 Mr. David Lochbaum Nuclear Safety Engineer Union of Concerned Scientists 1707 H Street NW, Suite 600 Washington, DC 20006

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS NOS. 62, REVISION 1, AND 3-32, REVISION 2 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NUMBERS 50-247 AND 50-286

1.0 INTRODUCTION

By letter dated May 25, 2004, as supplemented on August 9, 2004, Entergy Nuclear Operations, Inc. (Entergy, the licensee), submitted Requests for Relief Nos. RR-62, Revision 1, and RR 3-32, Revision 2, for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3),

respectively. The licensee proposed the use of two alternatives in lieu of certain defect repair requirements in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Sections III and XI for the reactor pressure vessel head (RPVH). The RPVH is an ASME Code,Section III, Class 1 component. These relief requests apply for the remainder of the third 10-year inservice inspection (ISI) intervals at IP2 and 3.

2.0 REGULATORY EVALUATION

The ISI of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components, and applicable edition and addenda as required by 10 CFR 50.55a(g), except when specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 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 applicant 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 the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, 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) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ISI Code of record for IP2 and 3 during the third 10-year ISI interval is the ASME Code,Section XI, 1989 Edition, No Addenda.

Enclosure

3.0 TECHNICAL EVALUATION

3.1 System/Component(s) for which Relief is Requested The IP2 RPVH, which includes control rod drive mechanism (CRDM) penetrations (90), in-core Instrumentation (ICI) penetrations (7), and one head vent penetration.

The IP3 RPVH, which includes CRDM penetrations (73), ICI penetrations (5), and one head vent penetration.

3.2 Code Requirements for which Relief is Requested The 1989 Edition, with No Addenda, of the ASME Code,Section XI, IWA-4120(a) states that:

"Repairs shall be performed in accordance with the Owners Design Specification and the original Construction Code of the component or system. Later Editions and Addenda of the Construction Code or of Section III, either in their entirety or portions thereof, and Code Cases may be used. The applicable Construction Code for IP2 and 3 is the ASME Code,Section III, 1965 Edition, with addenda through the Summer 1965 Addenda, including Code Cases 1332, 1335, 1339, and 1359. In accordance with IWA-4120(a), the licensee will follow the applicable requirements of the 1989 Edition of the ASME Code,Section III, in conjunction with the proposed alternatives described in Section 3.1.4 below, for RPVH penetration repairs.

The licensee requested relief from the following requirements in the 1989 Edition, with No Addenda, of the ASME Code:

a. Subsection IWA-4120(a) of Section XI to perform repairs on the RPVH penetrations per the rules of the Construction Code.

b. Subsections NB-4131, NB-2538, and NB-2539.1 of Section III to eliminate base material defects prior to repair welding.

3.3 Licensees Reason for Request (as stated by the licensee)

Entergy will be performing RPVH inspections during refueling outages to meet the requirements of the NRC First Revised Order EA-03-009 (Reference 1).

Entergy is requesting this relief as a contingency in the event that flaws requiring repair are identified during these inspections. The proposed embedded flaw process as described in [Westinghouse Electric Company Topical Report]

WCAP-15987-P-A (Reference 2), which incorporates the NRC safety evaluation (SE), and Reference 3 regarding J-groove weld inspection provide an acceptable alternative to repair reactor vessel head penetrations.

3.4 Licensees Proposed Alternative to the Code (as stated by the licensee)

Design, implementation of repairs, and inspections will be consistent with the information contained in References 2 and 3.

The embedded flaw repair overlay welds on the penetration J-groove welds will consist of a minimum of 3 deposited layers. The embedded flaw repair overlay

welds on the inside diameter (ID) and the outside diameter (OD) of the penetration tube material will consist of a minimum of 2 deposited layers of weld, consistent with References 2 and 3 to minimize welding induced residual stresses and material distortion. In the case of repairs on the ID surface, the 2 layer approach results in a reduced inlay excavation depth.

3.5 Licensees Basis for Use (as stated by the licensee)

In the NRC SE incorporated in Reference 2, the NRC staff concluded that, subject to the specified conditions and limitations, the embedded flaw process proposed in the WCAP provides an acceptable level of quality and safety. The staff also concluded that the WCAP is acceptable for referencing in licensing applications.

In both cases of the ID and the OD overlay repair welds, the proposed substitute examination methods have been previously demonstrated to be adequate for flaw detection and sizing as shown in Reference 3.

The embedded flaw repair process is considered a permanent repair that will last through the useful life of the RPVH. As long as a primary water stress corrosion cracking (PWSCC) flaw remains isolated from the primary water environment the only known mechanism for any further potential propagation is fatigue. The calculated fatigue usage in this region is very low, because the reactor vessel head region is isolated from the transients that affect the hot leg or cold leg piping.

The thickness of the weld used to embed the flaw has been set to provide a permanent embedment of the flaw. The embedded flaw process imparts less residual stresses than weld repair following the complete removal of the flaw.

Since Alloy 52/152 (690) weldment is considered highly resistant to PWSCC, a new PWSCC crack should not initiate and grow through the Alloy 52/152 overlay to reconnect the primary water environment with the embedded flaw. The resistance of Allow 690 material has been demonstrated by laboratory testing, and in approximately 10 years of operational service in steam generator tubes, where no PWSCC has been found.

Therefore, the embedded flaw repair process is considered to be an alternative to Code requirements that provides an acceptable level of quality and safety, as required by 10 CFR 50.55a(a)(3)(i).

3.6 Duration of the Proposed Alternative The licensee proposed to use the alternative for the remainder of the third 10-year ISI interval for IP2 and IP3. The third 10-year ISI interval for IP2 started July 1, 1994, and will end on April 3, 2006. The third 10-year ISI interval for IP3 started July 21, 2000, and will end on July 20, 2009.

3.7 Staff Evaluation The NRC staff has reviewed Westinghouse Topical Report WCAP-15987-P, Revision 2. In its July 3, 2003, letter (Reference 2), the staff accepted the referencing of this topical report for use with the following conditions and limitations:

1. Licensees must follow the NRC flaw evaluation guidelines provided in an NRC letter dated April 11, 2003 (Reference 4).

2. The crack growth rate referenced in WCAP-15987-P, Revision 2, is not applicable to Alloy 600 or Alloy 690 weld material, i.e., Alloy 52, 82, 152, and 182 filler material.

3. The nondestructive examination (NDE) requirements listed in the table below must be implemented for examinations of repairs made using the embedded flaw process.

Repair Location Flaw Orientation Repair Repair NDE ISI NDE of the Weld repair, Note 2 VHP Nozzle ID Axial Seal UT (ultrasonic testing) UT or Surface and Surface VHP Nozzle ID Circumferential Note 1 Note 1 Note 1 VHP Nozzle OD Axial or Note 1 Note 1 Note 1 above J-groove Circumferential weld VHP Nozzle OD Axial or Seal UT or Surface UT or Surface below J-groove Circumferential weld J-groove weld Axial Seal UT and Surface, UT and Surface, Note 3 Note 3 J-groove weld Circumferential Seal UT and Surface, UT and Surface, Note 3 Note 3 Notes: 1. Repairs must be reviewed and approved separately by the NRC.

2. Inspection consistent with the NRC Order EA-03-009 dated February 11, 2003, and any subsequent changes.

3. Inspect with personnel and procedures qualified with UT performance-based criteria.

Examine the accessible portion of the repaired region. The UT coverage plus surface coverage must equal 100 percent.

The licensee stated that it would follow the applicable requirements in the 1989 Edition, with No Addenda, of the ASME Code,Section III, in conjunction with the proposed alternatives, for RPVH penetration repairs. Since this edition of the Code is the same used as technical basis for Westinghouse Topical Report WCAP-15987-P, Revision 2, the licensee is not required to

perform a Code reconciliation between the original Code of Construction and the 1989 Edition, with No Addenda, of the ASME Code,Section III.

The licensee also stated that it would use the information in October 1, 2003, Westinghouse letter (Reference 3) as part of the proposed alternative. The letter was included in Westinghouse Topical Report WCAP-15987-P, Revision 2, which was reviewed and accepted by the NRC staff.

Therefore, the NRC staff finds the use of WCAP-15987-P, Revision 2, and the October 1, 2003, Westinghouse letter (References 2 and 3) for design, implementation of repairs, and inspections of the RPVH in lieu of applicable Code requirements to be acceptable for IP2 and IP3 during the third 10-year ISI interval.

3.8 Conclusion Based upon review of the information provided by the licensee in support of its relief request Nos. RR-62, Revision 1, and RR 3-32, Revision 2, the NRC staff concludes that the proposed alternative for design, implementation of repairs, and inspections of the RPVH in lieu of the defect repair requirements in the 1989 Edition, with No Addenda, of Sections III and XI of the ASME Code will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the use of the proposed alternatives at IP2 and IP3 during their respective third 10-year ISI intervals. All other ASME Code, Sections III and 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.

4.0 REFERENCES

1. U.S. NRC Order EA-03-009, Issuance of First Revised Order Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors, dated February 20, 2004. (Agencywide Documents Access and Management System (ADAMS) Accession No. ML040220181)

2. Westinghouse Topical Report, WCAP-15987-P-A, Revision 2, Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, dated December 2003. (ADAMS Accession No. ML031840237)

3. Westinghouse Letter, LTR-NRC-03-61, J.S. Galembush to T, Chan and B. Benney, NRC, dated October 1, 2003;

Subject:

Inspection of Embedded Flaw Repair of a J-groove Weld. (ADAMS Accession No. MLML032810457

4. NRC letter, R. J. Barrett to A. Marion, Nuclear Energy Institute, Flaw Evaluation Guidelines, dated April 11, 2003. (ADAMS Accession No. ML030980322)

Principal Contributor: R. Rodriguez Date: October 5, 2004