ML040850668
| ML040850668 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/19/2004 |
| From: | Richard Laufer NRC/NRR/DLPM/LPD4 |
| To: | Kansler M Entergy Nuclear Operations |
| Milano P, NRR/DLPM 415-1457 | |
| References | |
| TAC MC1696, TAC MC1697 | |
| Download: ML040850668 (10) | |
Text
March 19, 2004 Mr. Michael R. Kansler, President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
RELIEF REQUEST NOS. 70 AND 3-39 REGARDING ALTERNATIVE TO DEPTH SIZING CRITERIA, INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 (TAC NOS. MC1696 AND MC1697)
Dear Mr. Kansler:
In a letter dated December 30, 2003, Entergy Nuclear Operations, Inc. (Entergy), submitted Relief Request (RR) Nos. 70 and 3-39 for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and 3), respectively. Relief was requested from the non-destructive examination performance demonstration requirements of Appendix VIII, Supplement 4, to Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Specifically, Entergy proposes to use a depth sizing error not to exceed 0.15 inch root mean square as an alternative to the requirement that the performance demonstration results satisfy the statistical parameters specified in Subparagraph 3.2(c) of Supplement 4 to Appendix VIII.
The Nuclear Regulatory Commission (NRC) staff reviewed the proposed alternative in RR 70 and RR 3-39. The results are provided in the enclosed safety evaluation.
The NRC staff has concluded that the proposed alternative to the ASME Code requirements in RRs 70 and 3-39 provides an acceptable level of quality and safety. Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative is authorized for the remainder of the third inservice inspection interval which is until April 3, 2006, for IP2 and until July 20, 2009, for IP3.
If you should have any questions, please contact Patrick Milano at 301-415-1457. This completes the NRC staffs action on TAC Nos. MC1696 and MC1697.
Sincerely,
/RA/
Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286
Enclosure:
Safety Evaluation cc w/encl: See next page
March 19, 2004 Mr. Michael R. Kansler, President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
RELIEF REQUEST NOS. 70 AND 3-39 REGARDING ALTERNATIVE TO DEPTH SIZING CRITERIA, INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 (TAC NOS. MC1696 AND MC1697)
Dear Mr. Kansler:
In a letter dated December 30, 2003, Entergy Nuclear Operations, Inc. (Entergy), submitted Relief Request (RR) Nos. 70 and 3-39 for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and 3), respectively. Relief was requested from the non-destructive examination performance demonstration requirements of Appendix VIII, Supplement 4, to Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Specifically, Entergy proposes to use a depth sizing error not to exceed 0.15 inch root mean square as an alternative to the requirement that the performance demonstration results satisfy the statistical parameters specified in Subparagraph 3.2(c) of Supplement 4 to Appendix VIII.
The Nuclear Regulatory Commission (NRC) staff reviewed the proposed alternative in RR 70 and RR 3-39. The results are provided in the enclosed safety evaluation.
The NRC staff has concluded that the proposed alternative to the ASME Code requirements in RRs 70 and 3-39 provides an acceptable level of quality and safety. Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative is authorized for the remainder of the third inservice inspection interval, which is until April 3, 2006, for IP2 and until July 20, 2009, for IP3.
If you should have any questions, please contact Patrick Milano at 301-415-1457. This completes the NRC staffs action on TAC Nos. MC1696 and MC1697.
Sincerely,
/RA/
Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PUBLIC PDI-1 R/F A. Howe R. Laufer T. Chan P. Milano D. Votolato B. McDermott, R-I C. Bixler, R-I S. Little G. Hill (2)
J. Jolicoeur, EDO ACRS OGC Accession Number: ML040850668
- See previous concurrence OFFICE PDI-1:PM PDI-1:LA EMCB:SC OGC*
PDI-1:SC NAME PMilano SLittle SE dtd JMcGurren RLaufer DATE 03/17/04 03/17/04 02/20/04 03/17/04 03/17/04 OFFICIAL RECORD COPY
Indian Point Nuclear Generating Unit Nos. 2 & 3 cc:
Mr. Gary Taylor Chief Executive Officer Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. John Herron Senior Vice President and Chief Operating Officer Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Fred Dacimo Vice President, Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 295 Broadway, Suite 2 P.O. Box 249 Buchanan, NY 10511-0249 Mr. Christopher Schwarz General Manager, Plant Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 295 Broadway, Suite 2 P.O. Box 249 Buchanan, NY 10511-0249 Mr. Dan Pace Vice President Engineering Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Randall Edington Vice President Operations Support Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. John McCann Director, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Ms. Charlene Faison Manager, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Director of Oversight Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. James Comiotes Director, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.
Indian Point Energy Center 295 Broadway, Suite 2 P.O. Box 249 Buchanan, NY 10511-0249 Mr. Patric Conroy Manager, Licensing Entergy Nuclear Operations, Inc.
Indian Point Energy Center 295 Broadway, Suite 2 P. O. Box 249 Buchanan, NY 10511-0249 Mr. John M. Fulton Assistant General Counsel Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector, Indian Point 2 U. S. Nuclear Regulatory Commission 295 Broadway, Suite 1 P.O. Box 38 Buchanan, NY 10511-0038
Indian Point Nuclear Generating Unit Nos. 2 & 3 cc:
Senior Resident Inspector, Indian Point 3 U. S. Nuclear Regulatory Commission 295 Broadway, Suite 1 P.O. Box 337 Buchanan, NY 10511-0337 Mr. Peter R. Smith, President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mr. Paul Eddy Electric Division New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mayor, Village of Buchanan 236 Tate Avenue Buchanan, NY 10511 Mr. Ray Albanese Executive Chair Four County Nuclear Safety Committee Westchester County Fire Training Center 4 Dana Road Valhalla, NY 10592 Ms. Stacey Lousteau Treasury Department Entergy Services, Inc.
639 Loyola Avenue Mail Stop: L-ENT-15E New Orleans, LA 70113 Mr. William DiProfio PWR SRC Consultant 139 Depot Road East Kingston, NH 03827 Mr. Dan C. Poole PWR SRC Consultant 20 Captains Cove Road Inglis, FL 34449 Mr. William T. Russell PWR SRC Consultant 400 Plantation Lane Stevensville, MD 21666-3232 Mr. Alex Matthiessen Executive Director Riverkeeper, Inc.
25 Wing & Wing Garrison, NY 10524 Mr. Paul Leventhal The Nuclear Control Institute 1000 Connecticut Avenue NW Suite 410 Washington, DC, 20036 Mr. Karl Coplan Pace Environmental Litigation Clinic 78 No. Broadway White Plains, NY 10603 Mr. Jim Riccio Greenpeace 702 H Street, NW Suite 300 Washington, DC 20001
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
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF NOS. 70 AND 3-39 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286
1.0 INTRODUCTION
The inservice inspection (ISI) of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 1, Class 2, and Class 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Section 50.55a(g) of Title 10 of the Code of Federal Regulations (10 CFR), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) of 10 CFR 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 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 that 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 the third 10-year ISI interval at Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3) is the 1989 Edition (with no addenda) of the ASME Code.
By letter dated December 30, 2003, Entergy Nuclear Operations, Inc. (Entergy, the licensee) submitted a request for relief from certain ASME Code,Section XI requirements for ISI at IP2 and 3. Specifically, the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplement 4, Qualification Requirements for the Clad/Base Metal Interface of Reactor Vessel, requires that performance demonstration results satisfy the statistical parameters specified in Subparagraph 3.2(c). In lieu of Subparagraph 3.2(c), the licensee proposed using a depth sizing error not to exceed 0.15 inch root mean square (RMS).
2.0 DISCUSSION 2.1 Components for Which Relief Is Requested ASME Code,Section XI, Class 1, Examination Category B-A, Item No. B1.10, Circumferential and Longitudinal Shell Welds, and B1.20, Head Welds.
2.2 Code Requirements 10 CFR 50.55a(b)(2) was amended to reference Section XI of the ASME Code through the 1995 Edition, with the 1996 Addenda (64 FR 51370). The 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplement 4 requires that performance demonstration results satisfy the statistical parameters specified in Subparagraph 3.2(c), which states that performance demonstration results reported by the candidate, when plotted on a two-dimensional plot with the depth estimated by ultrasonics plotted along the ordinate and the true depth plotted along the abscissa, satisfy the following statistical parameters: (1) the slope of the linear regression line is not less that 0.7; (2) the mean deviation of flaw depth is less than 0.25 inch; and (3) the correlation coefficient is not less than 0.70.
2.3 Proposed Alternative Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed using the RMS value from the proposed rule of 10 CFR 50.55a(b)(2)(xv)(C)(1) (69 FR 892), which modifies the depth sizing criteria of the 1995 Edition with 1996 Addenda of the ASME Code, Section Xl, Appendix VIll, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c).
2.4 Licensee Basis for Use (As stated)
ASME Code, Section Xl, Appendix VIll, Supplement 4, Subparagraph 3.2(c) imposes three statistical parameters for depth sizing. The first parameter, 3.2(c)(1), pertains to the slope of a linear regression line. The linear regression line is the difference between actual versus true value plotted along a through-wall thickness. For Supplement 4 performance demonstrations, a linear regression line of the data is not applicable because the performance demonstrations are performed on test specimens with flaws located in the 15 percent through-wall. The differences between the actual versus true value produce a tight grouping of results, which resemble a shotgun pattern. The slope of a regression line from such data is extremely sensitive to small variations, thus making the parameter of 3.2(c)([1]), an inappropriate criterion.
The second parameter, 3.2(c)(2), pertains to the mean deviation of flaw depth.
The value used in the Code is too lax with respect to evaluating flaw depths within the inner 15 percent of wall thickness. Therefore, Entergy Nuclear Operations, Inc. (Entergy) proposes to use the more appropriate criterion of 0.15 inch RMS of 10CFR50.55a(b)(2)(xv)(C)(1), which modifies Subparagraph 3.2(a),
as the acceptance criterion. The third parameter, 3.2(c)(3), pertains to a correction coefficient. The value of the correction coefficient in Subparagraph 3.2(c)(3) is inappropriate for this application since it is based on the linear regression from Subparagraph 3.2(c)(1).
Entergy believes the proposed alternative to use the RMS value of 10 CFR 50.55a(b)(2)(xv)(C)(1), which modifies the criterion of ASME Code, Appendix VIIl, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c), will provide an acceptable level of quality and safety.
3.0 EVALUATION Supplement 4, Subparagraph 3.2(c) of Appendix VIII, requires that the ultrasonic performance demonstration results be plotted on a two-dimensional plot, with the measured depth plotted along the ordinate axis and the true depth plotted along the abscissa axis. For qualification, the plot must satisfy the following statistical parameters: (1) slope of the linear regression line is not less than 0.7; (2) the mean deviation of flaw depth is less than 0.25 inch; and (3) correlation coefficient is not less than 0.70.
The licensee proposes to eliminate the use of Supplement 4, Subparagraph 3.2(c), which imposes three statistical parameters for depth sizing. The first parameter, 3.2(c)(1), pertains to the slope of a linear regression line. The linear regression line is the difference between actual versus true value plotted along a through-wall thickness. For Supplement 4 performance demonstrations, a linear regression line of the data is not applicable because the performance demonstrations are performed on test specimens with flaws located in the inner 15 percent through-wall. The difference between actual versus true value produces a tight grouping of results which resembles a shotgun pattern. The slope of a regression line from such data is extremely sensitive to small variations, thus making the parameter of Subparagraph 3.2(c)(1) a poor and inappropriate acceptance criterion. The second parameter, 3.2(c)(2), pertains to the mean deviation of flaw depth. The value used in the Code is too lax with respect to evaluating flaw depths within the inner 15 percent of wall thickness. Therefore, the licensee proposes to use the more appropriate criterion of 0.15 inch RMS of 10 CFR 50.55a(b)(2)(xv)(C)(1), which modifies Subparagraph 3.2(a), as the acceptance criterion. The third parameter, 3.2(c)(3),
pertains to a correlation coefficient. The value of the correlation coefficient in Subparagraph 3.2(c)(3) is inappropriate for this application since it is based on the linear regression from Subparagraph 3.2(c)(1).
Based on the above, the NRC staff has determined that the use of Subparagraph 3.2(c) requirements is inappropriate as a screening parameter for determining the acceptability of Supplement 4 performance demonstration results. The depth sizing requirement of 0.15 inch RMS provides a better measure of NDE performance because the test specimen flaws are not evenly distributed through the entire wall thickness and this depth sizing requirement provides a specific deviation limit for the area of concern. Therefore, the proposed alternative to use the RMS value of the proposed rule 10 CFR 50.55a(b)(2)(xv)(C)(1), which modifies the criterion of Appendix VIII, Supplement 4, Subparagraph 3.2(a), and applies the same criteria to Subparagraph 3.2(c), specifically 0.15 inch RMS, will provide an acceptable level of quality and safety.
4.0 CONCLUSION
Based on the discussion above, the staff concludes that the proposed alternative to use the depth sizing criterion of Appendix VIII, Supplement 4, Subparagraph 3.2(a) as modified by 10 CFR 50.55a(b)(2)(xv)(C)(1), in lieu of Subparagraph 3.2(c), will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative is authorized for the third 10-year ISI interval for IP2 and 3. 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: D. Votolato Date: March 19, 2004