ML083540071
| ML083540071 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/22/2009 |
| From: | Lois James Plant Licensing Branch III |
| To: | Jensen J Indiana & Michigan Electric Co |
| beltz T, NRR/DORL/LPL3-1, 301-415-3049 | |
| References | |
| ISIR-27, TAC ME0158 | |
| Download: ML083540071 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 22, 2009 Mr. Joseph N. Jensen Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 SUB.IECT:
DONALD C. COOK NUCLEAR PLANT, UNIT 1 (CNP-1)- EVALUATION OF INSERVICE INSPECTION (lSI) TESTING PROGRAM RELIEF REQUEST (ISIR-27)(TAC NO. ME0158)
Dear Mr. Jensen:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 7, 2008, Indiana Michigan Power Company (the licensee) submitted a request for relief from certain examination requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) at CNP-1. The licensee proposed using a root mean square error criterion for sizing flaws that are greater than the ASME Section XI Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds" (N-695). This Code Case is referenced in Regulatory Guide 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1."
As documented in the enclosed Safety Evaluation, the NRC staff concludes that compliance with the AMSE Code for depth sizing is impractical, and that granting relief 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 NRC staff grants the alternative pursuant to 10 CFR 50.55a(g)(6)(i) for the remainder of the third 1a-year lSI interval at CNP-1 which began on July 1, 1996, and is scheduled to end on February 28, 2010, or until such time as an ultrasonic testing technique is capable of satisfying the requirements as specified in 1\\1-695.
J.Jensen
-2 If you have any questions, please contact Terry Beltz of my staff at (301) 415-3049.
Sincerely, 2A",,~
Lois M. James, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-315
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INSERVICE INSPECTION PROGRAM INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-315
1.0 INTRODUCTION
By letter dated October 7, 2008 (Agencywide Documents Access and Management System Accession No. ML083010237), Indiana Michigan Power Company (the licensee) requested relief from certain examination requirements of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (Code) at the Donald C. Cook Nuclear Plant, Unit 1 (CNP-1) through two relief requests (ISIR-26 and ISIR-27). This safety evaluation pertains only to ISIR-27, and a separate safety evaluation will be issued for ISIR-26.
The licensee proposed in ISIR-27 to use a root mean square error (RMSE) criterion for sizing flaws that are greater than the ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," (N-695). N-695 is referenced in Regulatory Guide (RG) 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1."
The request is for the remainder of the third 1O-yearinservice inspection (lSI) interval which began on July 1, 1996, and is scheduled to end on February 28, 2010.
2.0 REGULATORY REQUIREMENTS The lSI of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
In 10 CFR 50.55a(a)(3), it states, in part, that alternatives to the requirements of paragraph (g) may be used when authorized by the U.S. Nuclear Regulatory Commission (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.
In 10 CFR 50.55a(g)(6)(i), it states, in part, that the NRC may determine if a code requirement is impractical. Pursuant to 10 CFR 50.55a(g)(6)(i), the NRC may grant relief and impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Enclosure
- 2 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 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. As stated in 10 CFR 50.55a(g)( 4)(iv), inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph 10 CFR 50.55a(b), subject to the limitations and modification listed in 10 CFR 50.55a(b) and subject to Commission approval. Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met.
The code of record for the third 10-year lSI interval at CNP-1 is the 1989 Edition of the ASME Code.
3.0 TECHNICAL EVALUATION
FOR RELIEF REQUEST (ISIR-27) 3.1 Component Function/Description This relief request supports examination of ASME Code Class 1, reactor pressure vessel inlet and outlet nozzle to safe-end (dissimilar metals) butt welds, Table IWB-2500-1, Category B-F, Item Number B5.10.
3.2 Code Requirement for Which Relief is Requested The third 10-year lSI interval Code of Record is the 1989 Edition of the ASIVIE Code,Section XI.
For ultrasonic testing (UT) examinations, 10 CFR 50.55a(g)(6)(ii)(C)(2) requires the licensee to use the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplement 10.
N-695 is a Supplement 10 alternative that is endorsed by the NRC in RG 1.147, Revision 15.
N-695, Paragraph 3.3(c), states that "Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMSE of the flaw depth measurements as compared to the true flaw depths, do not exceed 0.125 in."
3.3 Licensee Proposed Alternative The licensee proposes to use the demonstrated 0.224-inches instead of the 0.125-inches specified in N-695 for depth sizing. In the event an indication is detected that requires depth sizing, the 0.099-inch difference between the required RMSE and the demonstrated RMSE (0.224 inches
- 0.125 inches = 0.099 inches) will be added to the measured through-wall extent for comparison with applicable acceptance criteria. If the examination vendor demonstrates an improved depth sizing RMSE prior to the examination, tho excess of that improved RMSE over the 0.125-inch RMSE requirement, if any, will be added 1) the measured value for comparison with the applicable acceptance criteria.
- 3 3.4 Licensee Basis for the Alternative To date, although vendors are qualified for detection and length sizing on these welds, vendors have not met the established RMSE requirement in 1\\1-695 for depth sizing (0.125-inches) when examining from the inner diameter. The licensee's examination vendor has demonstrated an ability to meet the depth sizing qualification requirement with an RMSE of 0.224-inches instead of the required 0.125-inches.
The licensee has determined that the alternative in this request will result in an acceptable level of quality and safety, pursuant to the provisions of 10 CFR 50.55a(a)(3)(i). The proposed alternative assures that the subject welds will be fully examined by procedures, personnel, and equipment qualified by demonstration in all aspects except depth sizing. For depth sizing, the proposed addition of the difference between the qualified by demonstrated sizing tolerance to any flaw that is required to be sized, compensates for the potential variation, and likewise assures an acceptable level of quality and safety.
3.5
NRC Staff Evaluation
The licensee's Code of record for the third 1a-year ISI interval as required by 10 CFR 50.55a(g)(6)(ii)(C)(2) for UT examinations is the 1995 Edition with 1996 Addenda. The ASME Code requires that dissimilar metal welds (OMW) be examined using procedures, equipment, and personnel qualified to Section XI, Appendix VIII, Supplement 10. However, the 1995 Edition with 1996 Addenda does not provide criteria for examinations performed from the inside diameter (10) of nozzles and piping. As an alternative to Supplement 10, the ASME Code developed N-695 for qualifications performed from either the 10 or outside diameter surfaces of OMWs. N-695 is endorsed in RG-1.147, Revision 15 with no conditions.
N-695 requires that the maximum error for flaw depth measurements, when compared to the true flaw depth, not exceed 0.125-inch RMSE. The U.S. nuclear power industry is using the Electric Power Research Institute (EPRI) Performance Oemonstration Initiative (POI) program to implement the performance demonstration required by N-695. To date, personnel and procedures have not been successful at meeting the 0.125-inch RMSE maximum N-695 qualification requirement for examinations performed from the 10.
The nuclear power industry uses the EPRI POI program to administer the performance demonstration testing required by N-695. The difficulties in meeting the RMSE requirement are associated with surface roughness and pipe misalignment that are common to field welds, and are replicated in mockups used in the EPRI POI program. The EPRI POI mockups contain the bounding OMW surface conditions found in nuclear power plants. There is the possibility that performance demonstrations performed on mockups with less severe surface conditions could meet the RMSE requirement; however, such rnockups are not available in the EPRI POI program.
In the event that vendors were able to qualify their UT techniques on mockups with less severe surface conditions than those used in the current EPRI POI mockups, the less severe surface conditions would be necessary for depth sizing flaws at nuclear power plants. The !\\IRC and EPRI-POI have been discussing the RMSE issue at semiannual meetings with industry representatives.
-4 The licensee proposed using the vendor's RMSE achieved in an EPRI PDI performance demonstration to approximate an actual flaw depth. The licensee has stated that the vendor's RMSE was 0.224-inches for examinations performed from the ID. The licensee proposed adding the depth sizing difference between the demonstrated 0.224-inch RMSE and the ASIVIE Code-required 0.125-inch RMSE to the measured value of any flaw detected during the examination of DMWs. Although the 0.224-inch RMSE can result in undersizing a flaw by an amount greater than the /'J-695-required error, the probability of a flaw occurring precisely when surface roughness is affecting UT is considered small.
The NRC staff finds that compliance with the N-695 required 0.125-inch RMSE, at this time, is impractical. Adding the difference between the performance-demonstrated depth sizing RMSE and the N-695 required depth sizing RMSE to an actual measured flaw size for determining flaw acceptability according to the standards specified in ASME Section XI, IWB-3500, provides reasonable assurance of structural integrity of the DI'v1Ws.
4.0 CONCLUSION
Based on the above review and evaluation, the NRC staff concludes that compliance with the N-695 required 0.125-inch RMSE for depth sizing is impractical, and that the proposed alternative to use 0.224-inch RMSE provides reasonable assurance of structural integrity of the DMWs that will be examined during the remainder of the third 1O-year lSI interval.
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 given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
The third 10-year lSI interval is scheduled to end February 28,2010. Pursuant to 10 CFR 50.55a(g)(6)(i), relief is granted to CNP-1 for the remainder of the third 10-year lSI interval or until such time as a UT technique is capable of satisfying the 0.125-inch RMSE requirement of N-695.
All other requirements of the ASIVIE Code,Section XI for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Donald Naujock, NRR Date: May 22, 2009
May 22, 2009 J.Jensen
- 2 If you have any questions, please contact Terry Beltz of my staff at (301) 415-3049.
Sincerely, IRAJ Lois M. James, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-315
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ Distribution:
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