ML090850504
| ML090850504 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/19/2009 |
| From: | Chernoff H Plant Licensing Branch 1 |
| To: | St.Pierre G Florida Power & Light Energy Seabrook |
| Egan, Dennis; NRR/DORL 301-415-2443 | |
| References | |
| TAC MD9785 | |
| Download: ML090850504 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 19, 2009 Mr. Gene F. St. Pierre Site Vice President c/o Michael O'Keefe Seabrook Station FPL Energy Seabrook, LLC P.O. Box 300 Seabrook, NH 03874
SUBJECT:
SEABROOK STATION, UNIT NO.1 - RELIEF REQUEST FOR USE OF AN ALTERNATE FLAW SIZING METHODOLOGY FOR THE SECOND INSERVICE INSPECTION INTERVAL (TAC NO. MD9785)
Dear Mr. St. Pierre:
By letter dated September 30,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082760464), as supplemented by letter dated November 12, 2008 (ADAMS Accession No. ML083220442), FPL Energy Seabrook, LLC, submitted a relief request from certain examination requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) at the Seabrook Station, Unit NO.1.
Specifically, the licensee proposed using a root mean square error criterion for sizing flaws that is greater than stated in ASME Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds" (N-695). The request is for the remainder of the second 1O-year inservice inspection interval.
The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's analysis in support of the request for relief. The staff has determined that it is impractical for the licensee to comply with the requirement and the proposed alternative provides reasonable assurance of structural integrity. The request is granted pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 55a(g)(6)(i) for the remainder of the second 1O-year inservice inspection interval or until such time as ultrasonic techniques are capable of satisfying the 0.125-inch root mean square error requirement of N-695, whichever is less. 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 facility.
The NRC staffs' evaluation and conclusions are contained in the enclosed safety evaluation.
This completes the NRC staff's efforts on TAC No. MD9785.
G. St. Pierre
- 2 If you have any questions, please contact the Seabrook Project Manager, Mr. Dennis Egan, at 301-415-2443.
Sincerely,
?/fU::-/
~~rold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443
Enclosure:
As stated cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST ASSOCIATED WITH THE SECOND INSERVICE INSPECTION INTERVAL FPL ENERGY SEABROOK, LLC SEABROOK STATION, UNIT NO.1 DOCKET NO. 50-443
1.0 INTRODUCTION
By letter dated September 30,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082760464), as supplemented by letter dated November 12, 2008 (ADAMS Accession No. ML083220442), FPL Energy Seabrook, LLC, submitted a relief request from certain examination requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) at the Seabrook Station, Unit NO.1.
Specifically, the licensee proposed using a root mean square (RMS) error criterion for sizing flaws that is greater than stated in 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 second 10-year inservice inspection (lSI) interval which began August 18, 2000, and is scheduled to end on August 17, 2010.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Paragraph 55a(g), "Inservice Inspection Requirements," requires, in part, that ASME Class 1,2, and 3 components must meet the inspection examination requirements set forth in the applicable editions and addenda of the ASME Code, except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Section 50.55a(g)(6)(i) of 10CFR states in part that the Commission will evaluate determinations that code requirements are impractical. 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 to the burden upon the licensee that could result if the requirements were imposed on the facility.
The Code of Record for the second 10-year lSI interval at the Seabrook Station, Unit NO.1 is the 1995 Edition through 1996 Addenda of the ASME Code,Section XI.
Enclosure
- 2
3.0 TECHNICAL EVALUATION
3.1 Affected Components The affected components are dissimilar metal welds (DMWs) at the Reactor Pressure Vessel (RPV) locations identified in the Table below.
Nozzle-to-Safe End Weld Identification Location IWB 2500-1 Catecorv Weld Type RC RPV-SE-301-121-A RPV "A" Outlet Nozzle @ 202 0 R-A Shop RC RPV-SE-302-121-B RPV "B" Inlet Nozzle ~ 247 0 R-A Shop RC RPV-SE-302-121-C RPV "C" Inlet Nozzle @ 293 0 R-A Shop RC RPV-SE-301-121-D RPV "0" Outlet Nozzle ~ 338 0 R-A Shop RC RPV-SE-301-121-E RPV "E" Outlet Nozzle @ 220 R-A Shop RC RPV-SE-302-121-F RPV "F" Inlet Nozzle @ 670 R-A Shop RC RPV-SE-302-121-G RPV "G" Inlet Nozzle @ 113 0 R-A Shop RC RPV-SE-301-121-H RPV "H" Outlet Nozzle @ 158 0 R-A Shop 3.2 Applicable Code The second 10-year lSI interval Code of Record for ultrasonic testing (UT) examinations is the 1995 Edition with 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplement 10.
Supplement 10, Paragraph 3.2(b) states that the RMS error for flaw depths estimated by UT shall not exceed 0.125-inch. However, the Code does not provide criteria for examinations performed from the inside diameter (10) surface.
N-695 is an alternative to Supplement 10 that is endorsed by the NRC in RG 1.147, Revision 15.
N-695, Paragraph 3.3(c), states that "[e]xamination 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, does not exceed 0.125 in." N-695 provides for qualifications performed from either the 10 or outside diameter (00) of DMWs.
3.3 Proposed Alternative For the subject welds, the licensee proposes using 0.189-inch as an alternative depth-sizing RMS error value, which is greater than the 0.125-inch RMS error value stated in N-695. To compensate, the licensee will add the difference between the required RMS error value of 0.125 inch RMS and the actual RMS value achieved by the inspection vendor to the flaw depth as determined during flaw sizing.
3.4 Licensee Basis for the Alternative The licensee stated that the nuclear industry has attempted to qualify personnel and procedures for depth sizing examinations from the inside surface of DMWs since November 2002. The most recent attempt at achieving 0.125-inch RMS was in early 2008. This attempt, as well as all previous attempts, did not achieve the required RMS values for personnel and procedures. The qualification attempts have been significant. The attempts have involved multiple vendors,
- 3 ultrasonic instruments, personnel and flaw depth sizing methodologies, all of which have been incapable of achieving the 0.125-inch RMS value.
The licensee states that adding the difference between the required RMS error and the achieved RMS error to the actual flaw size will provide a conservative flaw bounding approach and provide an acceptable level of quality and safety.
The impracticality of meeting the required 0.125-inch RMS value is demonstrated by the repeated failed industry attempts. Accordingly, approval of this alternative examination and evaluation process is requested pursuant to 10 CFR 50.55a(g)(6)(i).
3.5
NRC Staff Evaluation
The licensee's Code of Record for the second 1O-year lSI interval is the 1995 Edition with 1996 Addenda. The ASME Code requires that OMWs be examined using procedures, equipment, and personnel qualified to Section XI, Appendix VIII, Supplement 10. The Code of Record does not provide criteria for examinations performed from the 10 surface. As an alternative to Supplement 10, the ASME developed N-695 for qualifications performed from either the 10 or 00 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 RMS error. The U.S. nuclear power industry is using the Electric Power Research Institute's Performance Demonstration Initiative (POI) program to implement the performance demonstration required by N-695. The nuclear power industry has been trying to qualify personnel and procedures for 10 pipe examinations of OMWs since November 2002. Since then, UT techniques have undergone incremental improvements.
However, to date, personnel and procedures have not been successful at meeting the 0.125 inch RMS error maximum Code requirement.
The difficulties are associated with surface roughness and pipe misalignment common to field welds, which POI replicated in mockups used for UT qualifications. Currently, the POI program does not have mockups with less severe surface conditions for UT performance demonstrations and subsequent qualifications. In the event that UT qualifications were obtained on less severe surface conditions, the licensee would have to implement a repairlreplacement activity to provide these conditions in their plant. The NRC is involved in discussions at semiannual meetings addressing the considerations of physical modification to address 10 surface roughness and UT depth sizing qualifications.
The licensee proposed adding the depth sizing difference between the performance demonstrated RMS error and the ASME Code required 0.125-inch RMS error to the measured value of any flaw detected during the examination of the subject OMWs. The licensee has stated that the vendor's POI performance demonstration RI\\IIS error was 0.189-inch for N-695.
The NRC staff finds that compliance with the N-695 required 0.125-inch RMS error is impractical at this time and that adding the difference between the performance demonstrated depth sizing RMS error and the ASME Code-required depth sizing RMS error to the actual measured flaw value for determining flaw acceptability according to the standards specified in ASME Section XI, IWB-3500, provides reasonable assurance of structural integrity of the subject welds.
- 4
4.0 CONCLUSION
The staff has determined that it is impractical for the licensee to comply with the 0.125-inch RMS error requirement and the proposed alternative provides reasonable assurance of structural integrity of the subject welds. The request is granted pursuant to 10 CFR 50.55a(g)(6)(i} for the remainder of the second 1O-year inservice inspection interval or until such time as UT techniques are capable of satisfying the 0.125-inch RMS error requirement of N-695, whichever is less. 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 facility.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: D. G. Naujock Date: May 19, 2009
G. St. Pierre
- 2 If you have any questions, please contact the Seabrook Project Manager, Mr. Dennis Egan, at 301-415-2443.
Sincerely,
/raJ Harold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443
Enclosure:
As stated cc w/encls: Distribution via Listserv DISTRIBUTION:
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DATE 4/7/09 4/7/09 4/15/09 4/21/09 5/19/09 OFFICIAL RECORD COpy