ML102210133
| ML102210133 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/20/2010 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Bannister D Omaha Public Power District |
| Wilkins, L E, NRR/DORL/LPL4, 415-1377 | |
| References | |
| TAC ME2512 | |
| Download: ML102210133 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 20, 2010 Mr. David J. Bannister Vice President and CNO Omaha Public Power District Fort Calhoun Station 444 South 16th St. Mall Omaha, NE 68102-2247
SUBJECT:
FORT CALHON STATION, UNIT 1 - RELIEF REQUEST FOR USE OF AN ALTERNATE DEPTH-SIZING QUALIFICATION (TAC NO. ME2512)
Dear Mr. Bannister:
By letter dated October 30, 2009, as supplemented by letter dated February 16, 2010, Omaha Public Power District (the licensee) requested approval of relief from certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),
Section XI. The licensee stated that an ASME Code depth sizing at the present time is impractical. The licensee proposed an alternative to the depth sizing error requirements in ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," for the fourth 10-year inservice inspection (lSI) interval for Fort Calhoun Station (FCS) Unit 1. The fourth 10-year lSI interval began on October 3, 2003, and is scheduled to end on September 25, 2013.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(6)(i),
the licensee requested to use an alternative to the 0.125-inch root mean square error depth sizing requirement of ASME Code,Section XI, Appendix VIII, Supplement 10, and the alternative requirements of Code Case N-695. The licensee stated that relief is requested due to the impracticality of complying with the requirements of Code Case N-695.
Based on the information provided by the licensee in its request, the U.S. Nuclear Regulatory Commission (NRC) staff has determined compliance with the ASME Code requirements is impractical and that the proposed alternative provides reasonable assurance of structural integrity of the subject welds. Therefore, granting relief pursuant to 10 CFR 50.55a(g)(6)(i) for the proposed alternative 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 licensee. Relief is granted for the fourth 1O-year lSI interval for FCS which began on October 3, 2003 and ends on September 25, 2013.
All other ASME Code,Section XI, requirements for which relief has not been specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
D. Bannister
- 2 A copy of the NRC staff's safety evaluation is enclosed. If you have any questions concerning this matter, please contact Lynnea Wilkins at (301) 415-1377 or via e-mail at Lynnea.Wilkins@nrc.gov.
Sincerely, Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285
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 RELIEF REQUEST FOR ALTERNATIVE DEPTH-SIZING REQUIREMENT OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT 1 DOCKET NO. 50-285
1.0 INTRODUCTION
By letter dated October 30, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML093030358), as supplemented by letter dated February 16, 2010 (ADAMS Accession No. ML100480308), Omaha Public Power District (OPPD, the licensee) requested relief from certain requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, pursuant to paragraph 50.55a(a)(6)(i) of Title 10 of the Code of Federal Regulations (10 CFR). The licensee proposed an alternative to the depth-sizing error requirements in ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," for Fort Calhoun Station (FCS) Unit 1. The proposed alternative is applicable to FCS's fourth 10-year inservice inspection (lSI) interval which began on October 3, 2003, and is scheduled to end on September 25, 2013.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g), lSI of nuclear power plant components shall be performed in accordance with the requirements of the ASME Code,Section XI, except where specific written relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). The regulations in 10 CFR 50.55a(g)(6)(i) state that "the Commission may grant such relief and may impose such alternative requirements as it determines 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 of the burden upon the licensee that could result if the requirements were imposed on the facility."
Pursuant to paragraph 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 (lSI) 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 1O-year lSI interval and subsequent intervals comply with the Enclosure
- 2 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.
Pursuant to 10 CFR 50.55a(a)(3), alternatives to requirements may be authorized by the NRC if the licensee demonstrates that: (i) the proposed alternative provides 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. The regulations in 10 CFR 50.55a(g)(5)(iii) state that if the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in 10 CFR 50.4, information to support the determinations.
The lSI Code of record for the fourth 10-year lSI interval at FCS, Unit 1 is the ASME Code,Section XI, 1998 Edition, including Addenda through 2000.
3.0 TECHNICAL EVALUATION
3.1 Components Affected By the Relief Request Code Class:
Class 1 System Welds:
Reactor Coolant System Examination Categories:
Category B-F for dissimilar metal welds Code Item Number:
B5.10 for dissimilar metal welds Location Nozzle-to-Safe End Weld Weld Type N1A Outlet Nozzle (0°)
MRC-1/01 Shop N2A Inlet Nozzle (60°)
MRC-1/18 Shop N2B Inlet Nozzle (120°)
MRC-1/30 Shop N1 B Outlet Nozzle (180°)
MRC-2/01 Shop N2C Inlet Nozzle (240°)
MRC-2/18 Shop N2D Inlet Nozzle (300°)
MRC-2/30 Shop The component materials for each of the nozzle-to-safe end dissimilar metal welds are Alloy 82/182.
3.2
Applicable Code Edition and Addenda
The Code of record for the fourth 10-year lSI interval for FCS is the ASME Code,Section XI, 1998 Edition, including Addenda through 2000. The fourth 10-year lSI interval for FCS, Unit 1 began on October 3,2003, and is scheduled to end on September 25,2013.
3.3 Applicable Code Requirements (As stated by the licensee)
The examination of Class 1 piping welds are required to be performed using procedures, personnel and equipment qualified to the criteria of the applicable
- 3 ASME Code,Section XI, Appendix VIII, Supplements. The applicable supplement to this relief is 10, "QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS."
Paragraph 3.2, "Sizing Acceptance Criteria," Subparagraph (b) of Supplement 10, states that the "examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS [root mean square] error of the flaw depth measurements, as compared to the true flaw depths, is less than or equal to 0.125-inch (3.2mm)."
Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," provides alternative requirements to Appendix VIII, Supplement 10. Paragraph 3.3(c), of Code Case N-695 states, "Examination procedures, equipment, and personnel are qualified for depth sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125-in. (3 mm)." Code Case N-695 is unconditionally approved for use through Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 15.
3.4 Reason for Request (As stated by the licensee)
Impracticality of Compliance OPPD successfully completed volumetric examinations of the FCS reactor pressure vessel (RPV) nozzle-to-safe end dissimilar metal (DM) welds from the inside surface for a Materials Reliability Program (MRP-139) inspection during the Fall 2009 refueling outage that began November 1, 2009. No flaws were found. These examinations are scheduled to be performed again for a 10-year reactor vessel inservice inspection at the end of the current interval. OPPD will implement the NRC approved alternative requirements of Code Case N-695 for the qualification of procedures and personnel for examinations performed during these inspections.
This relief is submitted due to the impracticality of meeting the required 0.125-inch RMS value required by Code Case N-695. Code Case N-695 requires that qualified procedures and personnel shall demonstrate a flaw depth sizing error less than or equal to 0.125-inch RMS. The nuclear power industry has attempted to qualify personnel and procedures for depth sizing examinations performed from the inside surface of dissimilar metal welds (Supplement 10, Code Case N-695) since November 2002. To date, no personnel or procedure has achieved less than or equal to the ASME Code required 0.125-inch RMS error [NRC approved relief request regarding Braidwood Station, Units 1 and 2, dated November 8, 2007 (ADAMS Accession No. ML072760048)].
The inability of examination procedures to achieve the required RMS value is primarily due to a combination of factors such as surface condition, scan access, base materials and the dendritic structure in the welds themselves. The combination of these factors has proven too difficult for procedures and
- 4 personnel to achieve an RMS value that meet current Code requirements or Code Case N-695.
Burden Caused by Compliance The most recent attempt at achieving 0.125-inch RMS was in early 2008. This attempt, as well as previous attempts, did not achieve the required RMS values for personnel or procedures. The qualification attempts have been substantial.
The attempts have involved multiple vendors, ultrasonic instruments, personnel and flaw depth sizing methodologies, all of which have been incapable of achieving the 0.125-inch RMS value.
The process of qualification for this type of flaw sizing is well established. The cost and effort involved to perform a successful demonstration is quantifiable when a capable technique is available. However, when a capable technique is not available, the costs and effort required for a successful demonstration cannot be easily quantified.
3.5 Proposed Alternative (As stated by the licensee)
Fort Calhoun proposes using an alternative depth-sizing RMS error value greater than the 0.125-inch RMS error value stated in Code Case N-695 for the examination of welds listed in Table 1 of this relief request. Pursuant to 10 CFR 50.55a(g)(6)(i), relief is requested to use an alternative depth-sizing RMS error value due to the impracticality of achieving the value stated in Code Case N-695.
As an alternative to the required RMS error stated in Code Case N-695 for procedure and personnel depth sizing, Fort Calhoun will add the difference between the required RMS value of 0.125-inch and the actual RMS value achieved by our inspection vendor to the flaw depth as determined during flaw sizing. The inspection vendor chosen has achieved an RMS of 0.189" for Supplement 10 welds.
Applying the difference between the required RMS error and the achieved RMS error to the actual flaw size, will ensure a conservative flaw bounding approach and provide an acceptable level of quality and safety.
3.6 Duration of Proposed Alternative (As stated by the licensee)
The alternative requirements are requested for the duration of the Fourth 10-year lSI interval.
3.7
NRC Staff Evaluation
ASME Code,Section XI, Appendix VIII, Supplement 10 and Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds," require that examination procedures, equipment, and personnel be qualified for depth sizing such that the RMS error of the flaw depth measurements, as compared to true depths, do not exceed 0.125 inches. ASME Code,
- 5 Section XI, Code Case N-695 is referenced in the licensee's lSI program and has been determined to be an acceptable alternative to Appendix VIII, Supplement 10, per NRC Regulatory Guide 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Table 1, "Acceptable Section XI Code Cases."
The nuclear industry is in the process of qualifying personnel to ASME Code,Section XI, Appendix VIII, Supplement 10/ Code Case N-695 requirements, as implemented through industry's Performance Demonstration Initiative (POI) program which is managed by the Electric Power Research Institute. However, for ultrasonic examinations performed from the inside surface of a pipe weld, personnel have been unsuccessful at achieving the ASME Code required 0.125-inch RMS error flaw depth-sizing criterion. At this time, the NRC staff acknowledges that achieving the 0.125-inch RMS error is not feasible. The examination vendor contracted by the licensee has proposed to use an RMS error of 0.189 inches instead of the 0.125 inches required for Supplement 10 for dissimilar metal welds. In the event an indication is detected that requires depth sizing, the 0.064-inch difference between the required RMS error and the demonstrated RMS error for Supplement 10 will be added to the measured through-wall extent. This total flaw depth will then be assessed against the applicable acceptance criteria specified in Section IWB-3600 of the ASME Code for flaw evaluation.
Based on the above review, the NRC staff has determined that requiring the licensee to qualify procedures, personnel, and equipment to meet the maximum error of 0.125-inch RMS error for crack-depth sizing is impractical at the present time. The licensee's proposal of adding the difference between the ASME Code-required RMS error and the demonstrated RMS error to the measured through-wall extent, in addition to the use of the acceptance standards specified in Section IWB-3600 of the ASME Code, provides reasonable assurance of structural integrity of the supject welds.
4.0 Conclusion Based on the above review, the NRC staff concludes that the applicable ASME Code,Section XI, requirements are impractical for the depth sizing qualification on the FCS reactor pressure vessel nozzle-to-safe end dissimilar metal welds from the inside surface. Furthermore, the staff concludes that the licensee's alternative provides reasonable assurance of structural integrity of the subject welds.
The NRC staff concludes that 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 giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the licensee. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the proposed alternative is granted for the fourth 10-year lSI interval for FCS, Unit 1, which began on October 3, 2003, and ends on September 25, 2013.
- 6 All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: M. Audrain Date: August 20, 2010
D. Bannister
- 2 A copy of the NRC staff's safety evaluation is enclosed. If you have any questions concerning this matter, please contact Lynnea Wilkins at (301) 415-1377 or via e-mail at Lynnea.Wilkins@nrc.gov.
Sincerely, IRA by Carl F. Lyon fori Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrPMFortCalhoun Resource LPLIV rlf RidsNrrLAJBurkhardt Resource RidsAcrsAcnw_MailCTR Resource RidsOgcRp Resource RidsNrrDciCpnb Resource RidsRgn4MailCenter Resource RidsNrrDorlDpr Resource LTrocine, EDO RIV RidsNrrDorlLp/4 Resource I\\I1Audrain, NRRIDCI/CPNB ADAMS Accession No ML102210133
- SE memo dated OFFICE NRR/LPL4/PM NRR/LPL4/LA DCI/CPNB/BC NRR/LPL4/BC NRR/LPL4/PM NAME LWilkins JBurkhardt TLupold*
MMarkley CFLyon for LWilkins DATE 8/18/10 8/17/10 7/14/10 8/19/10 8/20/10 OFFICIAL AGENCY RECORD