ML070170502
| ML070170502 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/29/2007 |
| From: | Russell Gibbs NRC/NRR/ADRO/DORL/LPLIII-2 |
| To: | Crane C Exelon Generation Co |
| kuntz, Robert , NRR/DORL, 415-3733 | |
| References | |
| TAC MD1262, TAC MD1263 | |
| Download: ML070170502 (8) | |
Text
March 29, 2007 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 - EVALUATION OF INSERVICE INSPECTION PROGRAM RELIEF REQUEST 12R-47 PERTAINING TO PRESSURE TESTING OF PORTIONS OF THE PROCESS SAMPLING SYSTEM PIPING (TAC NOS. MD1262 AND MD1263)
Dear Mr. Crane:
By letter to the Nuclear Regulatory Commission (NRC) dated April 14, 2006, Exelon Generation Company, LLC (the licensee) submitted relief request I2R-47 for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)Section XI, to the 1989 Edition for Braidwood Station, Units 1 and 2 (Braidwood) for the second 10-year interval. Relief was requested from performing the ASME Code-required pressure test of portions of the process sampling system piping. The relief requested an alternative test consistent with the testing requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50 Appendix J, ?Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.
The NRC staff concludes, based on the enclosed safety evaluation, that pursuant to 10 CFR 50.55a(a)(3)(i), Relief Request I2R-47 for Braidwood is authorized on the basis that the proposed alternative provides an acceptable level of quality and safety. The relief request is authorized for the second 10-year interval for Braidwood.
Sincerely,
/RA/
Russell Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457
Enclosure:
Safety Evaluation cc w/encl: See next page
Mr. Christopher M. Crane March 29, 2007 President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 - EVALUATION OF INSERVICE INSPECTION PROGRAM RELIEF REQUEST 12R-47 PERTAINING TO PRESSURE TESTING OF PORTIONS OF THE PROCESS SAMPLING SYSTEM PIPING (TAC NOS. MD1262 AND MD1263)
Dear Mr. Crane:
By letter to the Nuclear Regulatory Commission (NRC) dated April 14, 2006, Exelon Generation Company, LLC (the licensee) submitted relief request I2R-47 for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)Section XI, to the 1989 Edition for Braidwood Station, Units 1 and 2 (Braidwood) for the second 10-year interval. Relief was requested from performing the ASME Code-required pressure test of portions of the process sampling system piping. The relief requested an alternative test consistent with the testing requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50 Appendix J, ?Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.
The NRC staff concludes, based on the enclosed safety evaluation, that pursuant to 10 CFR 50.55a(a)(3)(i), Relief Request I2R-47 for Braidwood is authorized on the basis that the proposed alternative provides an acceptable level of quality and safety. The relief request is authorized for the second 10-year interval for Braidwood.
Sincerely,
/RA/
Russell Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PUBLIC LPL3-2 R/F RidsOgcRp TBloomer, OGC RidsNrrDorlLpl3-2 RidsAcrsAcnwMailCenter RidsNrrPMRKuntz RidsNrrDciCfeb RidsNrrLAEWhitt RidsRgn3MailCenter RidsNrrDorlDpr Accession Number: ML070170502
- NLO with comments NRR-028 OFFICE LPL3-2/PM LPL3-2/LA CFEB/BC OGC LPL3-2/BC NAME RKuntz:mw EWhitt KGruss DRoth RGibbs DATE 3/28/2007 3/28/2007 3/28/2007 3 /23/2007 3/29/2007 OFFICIAL RECORD COPY
Braidwood Station Units 1 and 2 cc:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 Document Control Desk - Licensing Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Mr. Dwain W. Alexander, Project Manager Westinghouse Electric Corporation Energy Systems Business Unit Post Office Box 355 Pittsburgh, PA 15230 Ms. Bridget Little Rorem Appleseed Coordinator 117 N. Linden Street Essex, IL 60935 Howard A. Learner Environmental Law and Policy Center of the Midwest 35 East Wacker Dr., Suite 1300 Chicago, IL 60601-2110 U.S. Nuclear Regulatory Commission Braidwood Resident Inspectors Office 35100 S. Rt. 53, Suite 79 Braceville, IL 60407 Ms. Lorraine Creek RR 1, Box 182 Manteno, IL 60950 Illinois Emergency Management Agency Division of Disaster Assistance &
Preparedness 110 East Adams Street Springfield, IL 62701-1109 County Executive Will County Office Building 302 N. Chicago Street Joliet, IL 60432 Attorney General 500 S. Second Street Springfield, IL 62701 Plant Manager - Braidwood Station Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Site Vice President - Braidwood Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Senior Vice President - Operations Support Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Chairman, Ogle County Board Post Office Box 357 Oregon, IL 61061 Manager Regulatory Assurance - Braidwood Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Assistant General Counsel Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348
Braidwood Station Units 1 and 2 Senior Vice President - Midwest Operations Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Vice President - Regulatory and Legal Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Manager Licensing - Braidwood, Byron and LaSalle Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED RELIEF TO REQUEST I2R-47 PERTAINING TO PRESSURE TESTING OF PORTIONS OF THE PROCESS SAMPLING SYSTEM PIPING EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456, AND STN 50-457
1.0 INTRODUCTION
By letter to the Nuclear Regulatory Commission (NRC) dated April 14, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061090531), Exelon Generation Company, LLC (Exelon, the licensee), submitted Relief Request I2R-47, related to the Inservice Inspection (ISI) Program requirements pertaining to pressure testing portions of the Process Sampling (PS) system piping associated with the Post Accident Hydrogen Monitoring System for Braidwood Station, Units 1 and 2 (Braidwood) for the second 10-year ISI interval. In Relief Request I2R-47, the licensee requested relief from performing the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, pressure test of the subject piping by proposing an alternative test consistent with the testing requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J,
?Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.
2.0 REGULATORY REQUIREMENTS Paragraph 50.55a(g) requires that ISI of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph 50.55a(g) may be used, when authorized by the NRC staff, if the proposed alternatives would provide an acceptable 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 preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. 10 CFR 50.55a(g)(4)(i) requires that ISI of components and system pressure tests conducted during the initial 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. The ISI Code of record for the second 10-year inspection interval for Braidwood is the 1989 Edition of the ASME Code,Section XI.
3.0 TECHNICAL EVALUATION
3.1 System/Component for Which Relief is Requested Post Accident Hydrogen Monitoring System Piping, PS System 3.2 ASME Code Requirements The 1989 Edition of the ASME Code,Section XI, Table IWC-2500-1, Items C7.30/C7.40 (piping) and C7.70/C7.80 (valves) requires that specified piping be pressure tested and examined using the VT-2 examination method at a frequency of each inspection period and each inspection interval, respectively. The portion of the PS system containing the affected piping is not required to operate under normal plant operating conditions; therefore, as required by IWA-5210 and IWC-5221, only a system functional test is required.
IWC-5210(b) states that the contained fluid in the system shall serve as the pressurizing medium and where air is used, the test procedure shall permit the detection and location of through-wall leakages in components of the system tested.
3.3 Licensees Request for Relief Relief is requested from performing the ASME Code-required system pressure test for the Post Accident Hydrogen Monitoring System piping of the PS system that is located outside the containment during each inspection period and at the end of the inspection interval.
3.4 Licensees Basis for Requesting Relief The PS system piping referenced in Relief Request I2-R47 serves as the supply flow path from the containment to the hydrogen monitors and the return flow path from the hydrogen monitors to the containment using 1/4 inch and 1/2 inch tubing. The system medium is air. The system is composed of two separate trains for each unit with an approximate length of 930 feet. The tubing material is austenitic stainless steel and the system design pressure is 60 pounds per square inch gauge (psig). The nominal operating pressure is 10 psig maximum. A typical ASME Code-required system pressure test would consist of pressurizing the volume to 10 psig and performing a soap bubble or snoop test on all welds and piping. The containment penetration piping of the Post Accident Hydrogen Monitoring System is routinely tested following an outage in accordance with the requirements of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, to a pressure equivalent to the peak accident pressure which happens to be higher than the nominal operating pressure of the system. The licensee has proposed to extend the test boundary of the penetration piping to the remaining portion of Class 2 piping outside the containment using the acceptance criteria of Appendix J. This proposed method will provide equivalent leak detection to the soap bubble solution along with the VT-2 visual examination of the subject piping.
Any leakage in excess of the acceptance criteria used for the containment penetration will be investigated for the source of leakage which will be documented and resolved through the licensees corrective action program.
3.5 Licensees Proposed Alternative In lieu of performing the ASME Code-required system pressure test for the Post Accident Hydrogen Monitoring System piping of the PS system outside the containment during each inspection period and at the end of the inspection interval in accordance with the requirements of IWC-5210(b) for Braidwood, the licensee proposed an alternative to use the provisions of the Type C test in accordance with the requirements of 10 CFR 50, Appendix J. The licensee will use the acceptance criteria for the containment penetration piping for the leakage test in accordance with 10 CFR 50, Appendix J. For any leakage in excess of the acceptance criteria, the licensee will investigate the source of leakage and generate an Issue Report and resolve it in accordance with the Exelon Corrective Action Program.
4.0 NRC STAFF EVALUATION The ASME Code of Record requires a system functional test for the Post Accident Hydrogen Monitoring System piping outside the containment using air as the pressurizing medium and a soap bubble test of all welds and piping. The frequency of the test is once during each inspection period and at the end of an inspection interval. Each Braidwood unit has approximately 930 feet of stainless steel, 1/4 inch and 1/2 inch diameter piping, with socket weld joints that are subject to the pressure test. During normal plant operation the system is not required to operate. When the system is operational, the pressure in the piping is approximately 10 psig on the discharge side of the pump. Hence, the piping system is normally subject to low stresses and no known degradation mechanism. The containment penetration piping for the Post Accident Hydrogen Monitoring System is leak-rate tested following every outage in accordance with Appendix J of 10 CFR 50. But, the balance of this piping system outside the containment is required to be pressure tested once every 40 months in accordance with the Section XI of the 1989 Edition of the ASME Code. The licensee has proposed an alternative to the system pressure test by extending the test boundary of the Type C leakage test of 10 CFR 50, Appendix J, to cover the remainder of the system outside the containment.
The acceptance criteria for the leakage test per 10 CFR 50, Appendix J, would be less than or equal to 10 standard cubic feet per hour applied independently to the supply and return piping for each hydrogen monitor train. If the leak-rate exceeds the acceptance criteria, the licensee will investigate the source of leakage and generate an Issue Report and resolve it in accordance with the Exelon Corrective Action Program.
The NRC staff believes that the Type C leakage test of 10 CFR 50, Appendix J, conducted at the peak accident pressure of approximately 45 psig will provide a leak detection method equivalent to the ASME Code pressure test at 10 psig with application of soap bubble solution along with the VT-2 visual examination. Therefore, the proposed testing provides an acceptable level of quality and safety.
5.0 CONCLUSION
The ASME Code requires system pressure test for the Post Accident Hydrogen Monitoring System piping of the PS system outside the containment during each inspection period and at the end of the inspection interval in accordance with the requirements of IWC-5210(b) of the ASME Code,Section XI, 1989 Edition. The licensee proposed an alternative to use the provisions of the Type C leakage test in accordance with the requirements of 10 CFR 50, Appendix J. The NRC staff finds that the proposed alternative provieds an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative in Relief Request 12R-47 is authorized for the second 10-year ISI interval for Braidwood. All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: P. Patnaik Date: March 29, 2007