ML061780172
| ML061780172 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/18/2006 |
| From: | Raghavan L NRC/NRR/ADRO/DORL/LPLIII-1 |
| To: | Conway J Nuclear Management Co |
| Tam P | |
| References | |
| TAC MC8882 | |
| Download: ML061780172 (10) | |
Text
July 18, 2006 Mr. John T. Conway Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT (MNGP) - FOURTH 10-YEAR INTERVAL INSERVICE INSPECTION (ISI) PROGRAM PLAN, RELIEF REQUEST NO. 13 (TAC NO. MC8882)
Dear Mr. Conway:
By letter dated September 27, 2005, as supplemented on May 17, 2006, Nuclear Management Company (NMC) proposed its Fourth 10-Year Interval ISI Program Plan, Relief Request No. 13 for MNGP. NMC requested relief from the American Society of Mechanical Engineers (ASME)
Code,Section XI, requirement which specifies 100 percent volumetric examination coverage of all Class 1 reactor pressure vessel nozzle-to-shell welds.
The Nuclear Regulatory Commission (NRC) staff completed its review of the submittals and concluded that ASME Code requirements are impractical. The NRC staff further concluded that the examinations already performed by NMC would have detected any significant degradation that might have been present, providing reasonable assurance of the continued structural integrity of welds N-1A NV, N-2D NV, N-2E NV, N-2J NV, N-3A NV, N-4C NV, N-5B NV, and N-8A NV. The proposed relief is authorized by law and will not endanger life or property or the common defense or 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.
Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), relief is granted for the MNGP fourth 10-year ISI interval.
Details of the staff's review are set forth in the enclosed safety evaluation. If you have any questions, please call the Project Manager, Mr. Peter Tam at 301-415-1451.
Sincerely,
/RA/
L. Raghavan, Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
Safety Evaluation cc w/encl: See next page
ML061780172 OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/CVIB/SC OGC NRR/LPL3-1/BC NAME PTam THarris MMitchell* MLemoncelli LRaghavan DATE 7/7/06 6/28/06 6/9/06 7/13/06 7/18/06 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOURTH 10-YEAR INTERVAL INSERVICE INSPECTION RELIEF REQUEST NO. 13 MONTICELLO NUCLEAR GENERATING PLANT (MNGP)
NUCLEAR MANAGEMENT COMPANY DOCKET NO. 50-263
1.0 INTRODUCTION
By letter dated September 27, 2005 (Accession No. ML052760169) Nuclear Management Company (NMC, the licensee) proposed its Fourth 10-Year Interval Inservice Inspection Program Plan Request for Relief (RR) No. 13, for MNGP. The licensee provided additional information in its letter dated May 17, 2006 (Accession No. ML061420153).
The Nuclear Regulatory Commission (NRC) staffs evaluation of the licensees proposed relief follows.
2.0 REGULATORY EVALUATION
Inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable 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 NRC pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (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) shall meet the requirements, except the design and access provisions and the pre-service 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.
ENCLOSURE
The ASME Code of record for the MNGP fourth 10-year interval ISI program, which ends on May 31, 2012, is the 1995 Edition through the 1996 Addenda of Section XI of the ASME Code.
3.0 TECHNICAL EVALUATION
3.1 ASME Code Components The components affected by RR No. 13 are ASME Code,Section XI, Class 1, Reactor Vessel Nozzle-to-Vessel Shell welds specified in detail in Table A of the licensees application:
Recirculation Suction Nozzle N-1A, Weld N-1A NV Recirculation Inlet Nozzle N-2D, Weld N-2D NV Recirculation Inlet Nozzle N-2E, Weld N-2E NV Recirculation Inlet Nozzle N-2J, Weld N-2J NV Main Steam Discharge Nozzle N-3A, Weld N-3A NV Feedwater lnlet Nozzle N-4C, Weld N-4C NV Core Spray Inlet Nozzle N-58, Weld N-5B NV Jet Pump Instrumentation Nozzle N-8A, Weld N-8A NV 3.2 ASME Code Requirement The ASME Code,Section XI, Examination Category B-D, Item B3.90, requires 100 percent volumetric examination, as defined in Figures IWB-2500-7, a through d, as applicable, of Class 1 reactor pressure vessel (RPV) full penetration nozzle-to-shell welds. The licensee invoked ASME Code Case N-613-1 Ultrasonic Examination of Full Penetration Nozzles to Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Nozzle-In-Vessel Welds, Figures IWB-2500-7(a), (b) and (c),Section XI, Division 1, which was approved in an NRC safety evaluation (SE) dated October 6, 2004. Subsequent to the NRCs October 6, 2004 SE, Regulatory Guide (RG) 1.147, Revision 14, ISI Code Case Acceptability, ASME Section XI, Division 1, was issued in August 2005, in which ASME Code Case N-613-1 has been approved for general use without limitations. ASME Code Case N-613-1 allows the reduction of the examination volume next to the widest part of the weld from half of the vessel wall thickness to one-half (1/2) inch.
The licensee also invoked ASME Code Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds, ASME Section XI, Division 1 which is endorsed by the NRC in RG 1.147. Code Case N-460 states, in relevant part, "when the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent."
3.3 Licensees Proposed Alternative Examination In the application, the licensee stated:
In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested for the components listed in Table A on the basis that the required examination coverage of "essentially 100 percent" is impractical due to physical obstructions and the limitations imposed by design, geometry and materials of construction.
NMC performed qualified examinations that achieved the maximum, practical amount of coverage obtainable within the limitations imposed by the design of the components. Additionally, as Class 1 Examination Category B-P components, a VT-2 examination is performed on the subject components of the Reactor Coolant Pressure Boundary [RCPB] during system pressure tests each refueling outage. This was completed during the 2005 refueling outage and no evidence of leakage was identified for these components.
Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NMC requests relief from the requirements of ASME [Code] Section XI, Table IWB-2500-1, Category B-D, Item B3.90 and associated [ASME] Code Cases1, and proposes to utilize these completed exams as an acceptable alternative that provides reasonable assurance of continued structural integrity.
3.4 Licensees Basis for Relief Request The licensee based its relief request on the following:
The MNGP Nondestructive Examination (NDE) procedures incorporate improved inspection techniques qualified under Appendix VIII of the ASME Section XI Code by the Performance Demonstration Initiative (PDI) for examination of the subject nozzle-to-shell welds.
Coverage was obtained by following the scan parameters defined by the MNGP specific Electric Power Research Institute (EPRI) computer modeling report
[EPRI Internal Report IR-2004-63, "Monticello Nozzle Inner Radius and Nozzle-to-Shell Weld Examinations"] for each nozzle configuration and angle, and as designated within MNGP NDE procedures.
The examinations were performed using a manual contact method from the nozzle outside blend and vessel shell surfaces as discussed in the EPRI modeling report and as stated in MNGP procedures. The shear wave mode of propagation was used for each of the transducer and wedge combinations required for the inner 15 percent of the required parallel scan volume. The refracted longitudinal wave mode of propagation was used for the remaining outer 85 percent of the volume for parallel scans, and all of the perpendicular scans.
- 1. Relief can not be given for the requirements in an ASME Code Case pursuant to 10 CFR 50.55a(g)(5)(iii).
The NRC staffs evaluation of this relief will be evaluated only for the ASME Code,Section XI requirements that are determined to be impractical pursuant to 10 CFR 50.55a(g)(5)(iii).
According to the licensee, due to the design of these welds, a volumetric examination of 100 percent of the volume was not feasible to effectively perform as described in lWB-2500-7(b).
The nozzle-to-vessel welds are accessible from the vessel shell side of the weld, but examinations cannot be performed from the nozzle side of the weld because of the forging curvature. In addition, due to component configuration, certain nozzle-to-vessel weld examinations are further limited by the RPV design obstructions (such as appurtenances).
The licensee stated that the subject components received the required examination(s) to the extent practical within the limited access of the component design. For the examinations conducted, the licensee stated that satisfactory results were achieved, and no evidence of unacceptable flaws was detected with the improved inspection techniques.
The licensee stated that additional coverage for the limited areas was not achievable or practical, based on the latest qualified ultrasonic technology, nor by other considered examination methods such as radiography. The licensee has concluded that if significant degradation existed in the subject welds, it should have been identified by the examinations performed. Additionally, as Class 1 examination category B-P components, the licensee performed VT-2 examinations on the subject components in association with the RCPB system pressure test performed during the 2005 refueling outage, and identified no evidence of leakage.
The materials for the subject components are A533 CI I nozzle forgings welded to A502B CI II vessel shell plate. The licensee's review of operating experience within the nuclear industry did not reveal any instances of cracking in this location and type of weldment.
The licensee stated that the MNGP reactor vessel water chemistry is controlled in accordance with the 2004 revision to the BWRVIP-130, BWR [Boiling-Water Reactor] Water Chemistry Guidelines - 2004 Revision (EPRI Topical Report TR-1008192). Also, a hydrogen water chemistry system is used to reduce the oxidizing environment in the reactor coolant. The licensee stated that these additional measures provide added assurance against the initiation of cracking or corrosion from the inside surface of the reactor vessel for the subject components listed in this relief request. An inerted primary containment environment during operation provides assurance of corrosion protection on the outside surface of the reactor vessel.
3.5 Licensees Additional Information Additional Information was provided by the licensee in its letter dated May 17, 2006, to clarify its reference to the EPRI Internal Report IR-2004-63, "Monticello Nozzle Inner Radius and Nozzle-to-Shell Weld Examinations" regarding computer modeling for each nozzle configuration and angle as designated within MNGP NDE procedures. The licensee stated:
The Nondestructive Examination (NDE) procedures used at the Monticello Nuclear Generating Plant (MNGP) incorporate examination techniques qualified under Appendix VIII of the ASME Section XI Code by the Performance Demonstration Initiative (PDI) for examination of the subject nozzle-to-vessel shell welds.
The Electric Power Research Institute (EPRI) computer modeling report [EPRI Internal Report IR-2004-63, Monticello Nozzle Inner Radius and Nozzle-to-Shell Weld Examinations] was generated to assist NMC in developing and qualifying
Ultrasonic Test (UT) examination techniques for the MNGP nozzle inner corner regions and nozzle-to-vessel shell welds. The examinations were performed using a manual contact method from the nozzle outside blend radius and vessel shell surfaces as discussed in the EPRI modeling report and as stated in MNGP procedures. The UT scanning methodology modeled in the EPRI modeling report was applicable to the coverage for the inner corner regions and for the inner 15 percent volume of the nozzle-to-vessel shell welds when scanning parallel to the weld. The examination of the remaining outer 85 percent volume of the nozzle-to-vessel shell welds was based on a separate qualified technique and procedure which did not require use of the EPRI computer modeling report to validate.
The examinations for which relief was requested were not those modeled in the EPRI report for the inner 15 percent of the nozzle-to-vessel shell welds when scanning parallel to the weld. The UT examinations which were limited in coverage involved the remaining outer 85 percent of the required volume when scanning parallel to the weld, and the exam volume required when scanning normal to the weld. Therefore, the utilization of the EPRI computer modeling report for the MNGP has no bearing on the UT examination limitations included in the requested relief.
3.6 NRC Staff Evaluation The ASME Code requires 100 percent volumetric coverage of all Class 1 RPV nozzle-to-shell welds. The subject welds are carbon steel-to-carbon steel with a "set-in" nozzle configuration having a short radius of curvature on the nozzle side. This geometry limits scanning from the nozzle side of the weld such that 100 percent of the required examination coverage cannot be completed. For the licensee to achieve the ASME Code-required volumetric coverage, the subject nozzles would have to be redesigned and modified. This would place an undue burden on the licensee. Therefore, based on provided drawings and technical description of the nozzles, the NRC staff determined that the ASME Code requirements are impractical.
Ultrasonic examination of these welds was conducted using personnel, equipment, and procedures qualified through the EPRI PDI Program for ferritic pressure vessel welds. As shown on the sketches and technical descriptions provided by the licensee, a significant amount (approximately 78 percent to 82 percent) of the required volumetric coverage was obtained for nozzle-to-shell welds N-1A NV, N-2D NV, N-2E NV, N-2J NV, N-3A NV, N-4C NV, N-5B NV, and N-8A NV. This aggregate coverage includes greater than 90 percent of the examination volume using both 45- and 60-degree ultrasonic beam angles from the vessel side of the weld.
Round robin tests, as reported in NUREG/CR-5068, Piping Inspection Round Robin, have demonstrated that ultrasonic examinations of ferritic material from a single side provide high probabilities of detection (usually 90 percent or greater) for both near- and far-side cracks in blind inspection trials. While the licensee may not have achieved complete examination coverage (from both sides) as required by the ASME Code, the ultrasonic examinations performed by the licensee from the vessel side of the carbon steel weld meet the inspection guidelines documented in NUREG/CR-5068. Additionally, these examinations were performed with personnel, equipment, and procedures that have been demonstrated to meet EPRI PDI Program qualification requirements.
For these reasons, the examinations performed are expected to detect any significant degradation that might have been present, thus providing reasonable assurance of the continued structural integrity of welds N-1A NV, N-2D NV, N-2E NV, N-2J NV, N-3A NV, N-4C NV, N-5B NV, and N-8A NV.
4.0 CONCLUSION
The NRC staff has reviewed the licensees fourth 10-Year Interval ISI Program Plan, RR No. 13 for MNGP. The NRC staff concluded that ASME Code-requirements are impractical and to require the licensee to perform required ASME Code examinations would be a burden as the nozzles would have to be redesigned or modified. The NRC staff further concluded that the examinations already performed would have detect any significant degradation that might have been present, providing reasonable assurance of the continued structural integrity of welds N-1A NV, N-2D NV, N-2E NV, N-2J NV, N-3A NV, N-4C NV, N-5B NV, and N-8A NV. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), relief is granted for the MNGP fourth 10-year ISI interval. The NRC staff has determined 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 facility.
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: T. McLellan Date: July 18, 2006
TABLE A Code System Component Code Component Percent* Limitations Exam Category and ID and Examination Coverage Report and Component Volume Required Obtained Number Item No. Description B-D Reactor Vessel N-1A NV Nozzle-to-Vessel Weld, 83% Limited due to nozzle 2005UT041 B3.90 Recirculation Code Case N-613-1 configuration. Also, small Suction Figure 2 reduction due to interference Nozzles N-1A from welded thermocouple attachments.
B-D Reactor Vessel, N-2D NV Nozzle-to-Vessel Weld, Limited due to nozzle 2005UT028 B3.90 Recirculation Code Case N-613-1 82% configuration. Also, small Inlet Figure 2 reduction due to interference Nozzle N-2D from welded thermocouple attachments B-D Reactor Vessel, N-2E NV Nozzle-to-Vessel Weld, Limited due to nozzle 2005UT16 B3.90 Recirculation Code Case N-613-1 78% configuration.
Inlet Figure 2 Nozzle N-2E B-D Reactor Vessel, N-2J NV Nozzle-to-Vessel Weld, Limited due to nozzle B3.90 Recirculation Code Case N-613-1 78% configuration. 2005UT005 Inlet Figure 2 Nozzle N-2J B-D Reactor Vessel, N-3A NV Nozzle-to-Vessel Weld, Limited due to nozzle B3.90 Main Steam Code Case N-613-1 83% configuration. 2005UT023 Discharge Figure 2 Nozzle N-3A B-D Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B3.90 Feedwater Inlet N-4C NV Code Case N-613-1 79% configuration. 2005UT025 Nozzle N-4C Figure 2 B-D Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B3.90 Core Spray Inlet N-5B NV Code Case N-613-1 81% configuration. Also, small Nozzle N-5B Figure 2 reduction due to interference 2005UT018 from welded thermocouple attachments.
B-D Reactor Vessel, Nozzle-to-Vessel Weld Limited due to nozzle B3.90 Jet Pump N-8A NV Code Case N-613-1 83% configuration.
Instrumenttion Figure 2 2005UT037 Nozzle N-8A
- Due to the nozzle design, it was not feasible to effectively examine essentially 100 percent of the required examination volume as defined in figure 2 of Code Case N-613-1.
Percentages are conservatively rounded down to the nearest whole number. It should be noted that 100 percent of the inner 15 percent was examined in the parallel scans for all components listed above.
Monticello Nuclear Generating Plant cc:
Jonathan Rogoff, Esquire Commissioner Vice President, Counsel & Secretary Minnesota Department of Commerce Nuclear Management Company, LLC 85 7th Place East, Suite 500 700 First Street St. Paul, MN 55101-2198 Hudson, WI 54016 Manager - Environmental Protection Division U.S. Nuclear Regulatory Commission Minnesota Attorney Generals Office Resident Inspector's Office 445 Minnesota St., Suite 900 2807 W. County Road 75 St. Paul, MN 55101-2127 Monticello, MN 55362 Michael B. Sellman Manager, Nuclear Safety Assessment President and Chief Executive Officer Monticello Nuclear Generating Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 2807 West County Road 75 Hudson, MI 54016 Monticello, MN 55362-9637 Nuclear Asset Manager Robert Nelson, President Xcel Energy, Inc.
Minnesota Environmental Control 414 Nicollet Mall, R.S. 8 Citizens Association (MECCA) Minneapolis, MN 55401 1051 South McKnight Road St. Paul, MN 55119 Commissioner Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.
Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 November 2005