ML081600003
| ML081600003 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/01/2008 |
| From: | Melanie Wong NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | Morris R Duke Power Co |
| Stang J, NRR/DORL, 415-13456 | |
| References | |
| TAC MD6271, TAC MD6272, TAC MD6273 | |
| Download: ML081600003 (13) | |
Text
July 1, 2008 Mr. J. R. Morris Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745
SUBJECT:
CATAWBA NUCLEAR STATION, UNIT 2, REQUEST FOR RELIEF 07-CN-002, LIMITED WELD EXAMINATIONS DURING END-OF-CYCLE 14 REFUELING OUTAGE (TAC NOS. MD6271, MD6272 AND MD6273)
Dear Mr. Morris:
By letter dated July 11, 2007, Duke Energy Carolinas, LLC, the licensee, submitted a request for relief, Relief Request No. 07-CN-002, from the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),Section XI, 1998 Edition through the 2000 Addenda, requirement pertaining to limited weld examination coverage at the end of operating cycle 14 during the second 10-year inservice inspection (ISI) interval at Catawba Nuclear Station, Unit 2 (Catawba 2). The third 10-year interval for Catawba 2 started October 15, 2005, and ends August 19, 2016. The licensee already performed the scheduled second 10-year interval ISI on the referenced welds and components resulting in limited volumetric and visual coverages. As a result, the licensee has proposed that no alternate examinations or testing will be performed during the end of operating cycle 14 to compensate for the limited ultrasonic examination coverage.
The enclosed Safety Evaluation contains the U.S. Nuclear Regulatory Commission (NRC) staff's evaluation and conclusions. Based on the information provided in the licensee=s request for relief, the NRC staff has determined that it is impractical for the welds identified to be examined to the extent required by the ASME Code at Catawba 2. The NRC staff also concluded that reasonable assurance of structural integrity is provided by the examinations that were performed by the licensee.
Therefore, relief is granted and requirements are imposed pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(g)(6)(i) for the third 10-year ISI interval at Catawba 2 for referenced welds. Granting relief and imposing requirements are authorized by law and will not endanger life, property, or the common defense and security, and are 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.
J. Morris All other requirements of ASME Code,Section XI, for which relief has not been specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Sincerely,
/RA/
Melanie C. Wong Chief Plant Licensing Branch II-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-414
Enclosure:
Safety Evaluation cc w/encl: See next page
- SE input dated NRR-028 OFFICE LPL2-1/PM LPL2-1/LA NRR/CPNB/BC OGC LPL2-1/BC NAME JStang MO=Brien TChan* MSimon MWong DATE 6/30/08 6/30/08 01/07/08 06/25/08 7/1/08 Catawba Nuclear Station, Units 1 & 2 Page 1 of 2 cc:
Site Vice President Catawba Nulcear Station Senior Resident Inspector Duke Power Company, LLC U.S. Nuclear Regulatory Commission 4800 Concord Road 4830 Concord Road York, SC 29745 York, South Carolina 29745 Associate General Counsel and Managing Manager Attorney Division of Waste Management Duke Energy Carolinas, LLC Bureau of Land and Waste Management 526 South Church Street - EC07H Dept. of Health and Environmental Control Charlotte, North Carolina 28202 2600 Bull Street Columbia, South Carolina 29201-1708 Regulatory Compliance Duke Energy Corporation Manager 4800 Concord Road Nuclear Regulatory Issues York, South Carolina 29745 and Industry Affairs Duke Energy Corporation North Carolina Municipal Power 526 South Church Street Agency Number 1 Mail Stop EC05P 1427 Meadowwood Boulevard Charlotte, North Carolina 28202 P.O. Box 29513 Raleigh, North Carolina 27626 Saluda River Electric P.O. Box 929 County Manager of York County Laurens, South Carolina 29360 York County Courthouse York, South Carolina 29745 Vice President Customer Relations and Sales Piedmont Municipal Power Agency Westinghouse Electric Company 121 Village Drive 6000 Fairview Road Greer, South Carolina 29651 12th Floor Charlotte, North Carolina 28210 Assistant Attorney General North Carolina Department of Justice Owners Group (NCEMC)
P.O. Box 629 Duke Energy Corporation Raleigh, North Carolina 27602 4800 Concord Road York, South Carolina 29745 NCEM REP Program Manager 4713 Mail Service Center Senior Counsel Raleigh, North Carolina 27699-4713 Duke Energy Carolinas, LLC 526 South Church Street - EC07H North Carolina Electric Membership Corp. Charlotte, NC 28202 P.O. Box 27306 Raleigh, North Carolina 27611
Catawba Nuclear Station, Units 1 & 2 Page 2 of 2 cc:
Division of Radiation Protection NC Dept. of Environment, Health, and Natural Resources 3825 Barrett Drive Raleigh, North Carolina 27609-7721 Group Vice President, Nuclear Generation and Chief Nuclear Officer P.O. Box 1006-EC07H Charlotte, NC 28201-1006
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF NO. 07-CN-002 CATAWBA NUCLEAR STATION, UNIT 2 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414
1.0 INTRODUCTION
By letter dated July 11, 2007, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072640194), Duke Energy Carolinas, LCC, the licensee, submitted a request for relief, Relief Request No. 07-CN-002, from the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),Section XI, 1998 Edition through the 2000 Addenda requirement pertaining to limited weld examination coverage at the end of operating cycle 14 during the second 10-year inservice inspection (ISI) interval at Catawba Nuclear Station, Unit 2 (Catawba 2). The third 10-year interval for Catawba 2 started October 15, 2005, and ends August 19, 2016. The licensee already performed the scheduled third 10-year interval ISI on the referenced welds and components resulting in limited volumetric and visual coverages. As a result, the licensee has proposed that no alternate examinations or testing will be performed during the end of operating cycle 14 to compensate for the limited ultrasonic examination coverage.
2.0 REGULATORY REQUIREMENTS Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(g) specifies that inservice inspection (ISI) of nuclear power plant components shall be performed in accordance with the requirements of the ASME Code,Section XI, and applicable addenda, except where specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to Section 50.55a(g)(6)(i). Section 10 CFR 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. Section 50.55a(g)(5)(iii) states that if the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the NRC and submit, as specified in 10 CFR 50.4, information to support the determinations.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, 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 components. The regulations require that Enclosure
inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements of 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 applicable code of record for the second ISI interval for Catawba 2 is the ASME Code,Section XI, 1998 Edition through the 2000 Addenda.
The information provided by the licensee in support of the request has been evaluated by the NRC staff and the bases for disposition are documented below.
3.0 TECHNICAL EVALUATION
3.1 Licensees Evaluation 3.1.1 Components for Which Relief is Requested Relief Request 07-CN-002:
- Weld ID Number 2NI55-11, Safety Injection System Valve 2NI081 to Pipe Circumferential Weld;
- Weld ID Number 2NI55-8, Safety Injection System Pipe to Tee Circumferential Weld; and
- Weld ID Number 2NV12-10, Chemical and Volume Control System Tee-to-4x3 Reducer Circumferential Weld.
Relief Request 07-CN-001 requests relief from the requirements listed below:
Examination Item Component Examination Requirement Category No.
B-J B9.11 2NI55-11 Essentially 100-percent volumetric examination of specified weld examination volume.
B-J B9.11 2NI55-8 Essentially 100-percent volumetric examination of specified weld examination volume.
C-F-1 C5.21 2NV12-10 Essentially 100-percent volumetric examination of specified weld examination volume.
3.1.2 Impracticality/Burden Caused by Code Compliance During the ultrasonic examination of weld ID Number 2NI55-11, 100-percent coverage of the required examination volume could not be obtained due to valve configuration. Ultrasonic examination resulted in 37.50-percent coverage of the required examination volume; this percentage of coverage reported represents the aggregate coverage from all scans performed on the weld and adjacent base material.
The 45-degree shear wave circumferential scans, both clockwise and counter-clockwise covered 50 percent of the weld and base material; the 60-degree shear wave scan from the pipe side perpendicular to the weld covered 50 percent of the weld and base material. A supplemental scan using a 60-degree refracted longitudinal wave search unit covered 50 percent of the examination volume on the valve side from one direction perpendicular to the weld but is not
included in the coverage calculations because 10 CFR 50.55a(b)(2)(xv)(A)(1) mandates scanning from four directions two axially and two circumferentially for welds of dissimilar metal.
There were no recordable indications found during the inspection of this weld. The coverage limitation was caused by the taper on the valve side of the weld which did not allow placement of the search unit on that side of the weld. In order to scan all of the required volume for this weld, the valve would have to be redesigned to allow scanning from both sides of the weld.
During the ultrasonic examination of weld ID Number 2NI55-8, 100-percent coverage of the required examination volume could not be obtained due to the tee configuration. Ultrasonic examination resulted in 62.50-percent coverage of the required examination volume; this percentage of coverage reported represents the aggregate coverage from all scans performed on the weld and adjacent base material.
The 45-degree shear wave circumferential scans, both clockwise and counter-clockwise, covered 100 percent of the weld and base material; the 60-degree shear wave scan from the pipe side perpendicular to the weld covered 50 percent of the weld and base material. A supplemental scan using a 60-degree refracted longitudinal wave search unit covered 50 percent of the examination volume on the tee side from one direction perpendicular to the weld but is not included in the coverage calculations because 10 CFR 50.55a(b)(2)(xv)(A)(1) mandates scanning from four directions two axially and two circumferentially for welds of dissimilar metal. There were no recordable indications found during the inspection of this weld.
The coverage limitation was caused by the radius on the tee side of the weld which prevented placement of the search unit on that side of the weld. In order to scan all of the required volume for this weld, the tee would have to be redesigned to allow scanning from both sides of the weld.
During the ultrasonic examination of Weld ID Number 2NV12-10, 100-percent coverage of the required examination volume could not be obtained due to the tee configuration. Ultrasonic examination resulted in 78.70-percent coverage of the required examination volume; this percentage of coverage reported represents the aggregate coverage from all scans performed on the weld and adjacent base material.
The 45-degree shear wave circumferential scans, both clockwise and counter-clockwise, covered 100 percent of the weld and base material; the 60-degree shear wave scan from the reducer side perpendicular to the weld covered 71.60 percent of the weld and base material in two directions; and a 60-degree shear wave scan from the tee side where access was available covered 43.30 percent of the weld and base material in two directions and 56.74 percent in one direction. A supplemental scan using a 60-degree refracted longitudinal wave search unit covered 50 percent of the examination volume on the tee side from one direction perpendicular to the weld in the obstructed area but is not included in the coverage calculations because 10 CFR 50.55a(b)(2)(xv)(A)(1) which mandates scanning from four directions two axially and two circumferentially for welds of dissimilar metal. There were no recordable indications found during the inspection of this weld.
The coverage limitation was caused by the radius on the tee side of the weld which prevented placement of the search unit on that side. In order to scan all of the required volume for this weld, the tee would have to be redesigned to allow scanning from both sides of the weld.
For Request for Relief 07-CN-002, the examinations were performed using personnel qualified in accordance with ASME Code,Section XI, Appendix VII. The ultrasonic procedures and equipment complied with the requirements of ASME Code,Section XI, Appendix VIII as administered by the Performance Demonstration Initiative (PDI), 1998 Edition through the 2000 Addenda.
3.1.3 Proposed Alternative Examinations or Testing The scheduled 10-year code examinations were performed on the referenced welds and resulted in the noted limited-coverage of the required ultrasonic volume. No alternate examinations or testing are planned for the welds during the current inspection interval which will end on August 19, 2016, for welds 2NI55-11, 2NI55-8 and 2NV12-10.
3.1.4 Justification for Granting Relief The base materials for all of these welds are austenitic-based materials which (a) have high corrosion resistance with low contribution of corrosion products to the coolant, (b) have good mechanical properties and (c) are highly weldable. Very few service-induced problems with stainless steel in pressurized-water reactions (PWR) primary system applications have been observed in operating plants. There has been limited susceptibility to stress corrosion cracking (SCC) due to chloride contamination and cracking in stagnant borated systems. Chemistry limits on chlorides, fluorides and sulfides and dissolved oxygen are controlled by selected licensee commitments (SLC) and other administrative procedures at Catawba 2 to ensure that any favorable conditions for SCC are precluded. Additionally, controls on welding filler material consistent with Regulatory Guide 1.31, Control of Ferrite Content in Stainless Steel Weld Metal Revision 3, also have served to limit the susceptibility of these welds to SCC. No other known degradation mechanisms are applicable to this material at this particular location within the system.
The piping containing weld 2NI55-11 is an ASME Code, Class 1 line with a design temperature of 650 degrees Fahrenheit (°F) and a design pressure of 2500 pounds per square inch absolute (psia). The weld is located inside the Catawba 2 containment. The piping line connects the cold leg accumulator 2C to the reactor coolant system and also ties the residual heat removal (RHR) pump discharge header back to the reactor coolant loop (RCL) cold leg 2C. The primary function of this piping is to serve as part of the flow path that a) supplies emergency core cooling system (ECCS) injection from the safety injection and RHR systems to the RCL during accident conditions and b) provides core decay heat removal during shutdown/startup operations. This line is normally stagnant during normal plant operation.
Weld 2NI55-11 is a circumferential butt weld just upstream of valve 2NI81. The weld is between valve 2NI81 and a section of 10, schedule 140 pipe. The valve body material is SA182 F316 and the attached piping is SA376 TP316.
This piping is not insulated. During each refueling outage, multiple walkdowns of the containment area are performed to determine the presence of external leakage. These walkdowns include a boric acid walkdown while the primary system remains at temperature and pressure. Other walkdowns performed during the outage are system engineer walkdowns, operation walkdowns at 350 pounds per square inch (psi), 1000 psi and normal operation pressure, and the ASME Code,Section XI, IWA-5000 system leakage test. During these
various walkdowns, any leakage from this weld would be recognized by active leakage or boron deposit buildups around the valve and piping.
In addition, any leakage during operation at the weld location would be detected by various other leakage detection systems available to the operator. The primary detection method at this location is the containment floor and equipment sump level monitors where unidentified accumulated water on the containment floor would be monitored and evaluated as sump level changes. Leakage would also be promptly identified through a continuous decline in the cold leg accumulator level.
These walkdowns and leakage detection systems provide a high level of confidence that any leakage would be promptly identified at this welded joint inside containment.
The piping containing weld 2NI55-8 is an ASME Code, Class 1 line with a design temperature of 650 degrees °F and a design pressure of 2500 psia. The weld is located inside the Catawba 2 containment. The piping line connects the cold leg accumulator 2C to the reactor coolant system and also ties the RHR pump discharge header back to the RCL cold leg 2C. The primary function of this piping is to serve as part of the flow path that a) supplies ECCS injection from the safety injection system and RHR systems to the RCL during accident conditions and b) provides core decay heat removal during shutdown/startup operations. This line is normally stagnant during normal plant operation.
Weld 2NI55-8 is a circumferential butt weld on a 6 schedule 160 line between valves 2NI81 and 2NI82. The weld is between a 10 x 10 x 6 seamless, reducing tee and 6 schedule 160 piping.
The tee material is SA403 WP316 and the attached piping is SA376 TP304.
This piping is normally covered by metal reflective insulation. During each refueling outage, multiple walkdowns of the containment area are performed to determine the presence of external leakage. These walkdowns include a boric acid walkdown while the primary system remains at temperature and pressure (Mode 3). Other walkdowns performed during the outage are system engineer walkdowns, operation walkdowns at 350 psi, 1000 psi and normal operation pressure, and the ASME XI, IWA-5000 system leakage test. During these various walkdowns, any leakage from this weld would be recognized by active leakage or boron deposit buildups around the valve and mirror insulation.
In addition, leakage during operation (assuming leakby of the primary isolation valve, 2NI82) at the weld location would be detected by various other leakage detection systems available to the operator. These systems identified with plant technical specification include:
- Containment atmosphere particulate radioactivity (EMF 38) monitor which would detect airborne radiological activity;
- Containment ventilation unit condensate drain tank level monitor which collects and measures as unidentified leakage the moisture removed from the containment atmosphere;
- Containment floor and equipment sump level monitors where identified accumulated water on the containment floor would be monitored and evaluated as sump level changes;
- A reactor coolant system water inventory balance is performed on a regular basis (i.e., at least once every 3 days). The normal operating practice is to perform this computer
based program on a daily frequency and/or whenever the operators suspect any abnormal changes to other leakage detection systems. A plant technical specification requires system leakage from unidentified sources to be maintained below 1 gpm; however, plant operation procedure (PT/1(2)/A/4150/001D, NC system leakage calculation) establishes an administrative limit of 0.15 gpm above which the source of leakage will be investigated. Leakage as a result of a failed weld would show up as unidentified leakage and be subject to the 0.15 gpm administrative limit; and
- Other leakage detection systems available to the operator include (1) volume control tank (VCT) level changes, (2) VCT make-up frequencies, and (3) cold leg accumulator level changes.
Without leakby of the primary isolation valve, leakage would be promptly identified through a continuous decline in the cold leg accumulator level.
Weld ID 2NV12-10 is between a 4 schedule 160 butt welded tee and a 4 x 3 concentric reducer. The weld is downstream of the Centrifugal Charging (NV) Pump 2B on ASME Class 2 piping with a design temperature of 250°F and a design pressure of 2750 psia. The pipe containing this weld is located in the auxiliary building and is pressurized during normal operation. This weld maintains the pressure boundary as part of normal charging flowpath and the ECCS flow path under accident conditions.
The piping is not insulated and is located in Room 241 (NV2B Pump Room) on the 543 ft.
elevation of the auxiliary building. Leakage during normal operation would be seen as active leakage due to low fluid temperature conditions and readily identified on the 543 ft. floor elevation below. The room is accessible during normal operation and is within the scope of the weekly operation walkdowns. Periodic system engineer walkdowns are also performed that include leakage identification on the chemical volume and control (NV) system.
In addition to walkdowns, this weld location is part of the reactor coolant system mass balance performed daily. An operational leak rate test (PT/2/A/4206/006) for the chemical volume and control system is performed with the system pressurized on an eighteen month frequency.
Furthermore, an ASME Code,Section XI, IWA-5000 system leakage test is also performed every ISI period. Any of these tests would identify leakage at this particular weld. These walkdowns and leakage tests provide a high level of confidence that any leakage would be promptly identified at this welded joint in the NV2B Pump room of the Auxiliary Building.
3.2 NRC Staff Evaluation By letter dated July 11, 2007, Duke Energy Carolinas, LLC, requested pursuant to 10 CFR 50.55a(g)(5)(iii) approval of Relief Request 07-CN-002 which involved limited weld examination coverage at Catawba 2. Specifically, during the ultrasonic examination conducted for the first period of the third 10-year inservice inspection interval (Request for Relief 07-CN-002), 100-percent coverage of the required examination volume could not be obtained for welds 2NI55-11, 2NI55-8 and 2NV12-10 because of weld configuration and the geometry of associated components.
NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability - ASME Section XI, Division 1 endorses ASME Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 Welds. Code Case N-460 defines weld examination coverage greater than 90 percent to meet the essentially 100-percent requirement specified in ASME Code,Section XI.
The applicable code of record for the second ISI interval for Catawba 2 is the ASME Code,Section XI, 1998 Edition through the 2000 Addenda.
The NRC staffs review of the data submitted for the subject welds regarding the inspection volumes found that 100-percent coverage of the required examination volume could not be obtained due to the geometry of the welds and associated components. The coverage limitations and inspection results for each component are listed in the following table:
Component Exam Category/Item No. Coverage Limitation Recordable ID Figure No. Indications Aggregate Coverage Obtained (Yes/No) 2NI55-11 B-J/B9.11 Taper on the valve side of No IWB-2500-8(c) the weld prevented 37.50-percent coverage placement of the search unit on that side of the weld.
2NI55-8 B-J/B9.11 Radius on the tee side of the No IWB-2500-8(c) weld prevented placement of 62.50-percent coverage the search unit on that side of the weld.
2NV12-10 C-F-1/C5.21 Radius on the tee side of the No IWC-2500-1 weld prevented placement of 78.70-percent coverage the search unit on that side of the weld.
The licensee also stated that in order to achieve more coverage, in each case, the welds and the associated components would have to be redesigned to allow more scanning area. The NRC staff finds that this alternative would be impractical and would impose undue burden on the licensee.
Based on our review of the information provided, the NRC staff finds that the licensees proposed alternative provides reasonable assurance of structural integrity for the subject welds.
This conclusion is based on the fact that the subject welds have been examined to the extent practical using 45-degree and 60-degree shear waves and 60-degree refracted longitudinal waves achieving an aggregate coverage of 37.50 percent to 78.70 percent with no recordable indications found during the inspections. Therefore, the NRC staff finds that any significant degradation, if present, should have been detected.
3.0 CONCLUSION
S Based on the above reviews, the NRC staff concludes that due to the geometry of the welds and associated components, the Code-required examinations are impractical to perform to the extent required by the Code. Furthermore, the examinations performed by the licensee provide reasonable assurance of continued structural integrity of the subject components. Therefore, request for Relief 07-CN-002 is granted pursuant to 10 CFR 50.55a(g)(6)(i) for the third 10-year ISI interval. This grant of relief is authorized by law and will not endanger life or property or
common defense and security, and is otherwise in the public interest, giving due consideration to the burden upon the licensee and facility 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 authorized, herein, by the NRC staff remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: C. Nove Date: July 1, 2008