ML081750252
| ML081750252 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/31/2008 |
| From: | Boyce T NRC/NRR/ADRO/DORL/LPLII-2 |
| To: | Young D Progress Energy Co |
| SABA F, NRR/DORL/LPLI-2, 415-1453 | |
| References | |
| TAC MD7738 | |
| Download: ML081750252 (7) | |
Text
July 31, 2008 Mr. Dale E. Young, Vice President Crystal River Nuclear Plant (NA1B)
ATTN: Supervisor, Licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 34428-6708
SUBJECT:
CRYSTAL RIVER NUCLEAR PLANT, UNIT 3 - RELIEF REQUEST NO. 07-001-PT RELATED TO THE FOURTH 10-YEAR INTERVAL INSERVICE INSPECTION ON SYSTEM PRESSURE TEST BOUNDARY (TAC NO. MD7738)
Dear Mr. Young:
By letter dated December 21, 2007, Florida Power Corporation (licensee) submitted Relief Request (RR) No. 07-001-PT, related to the fourth 10-year inservice inspection (ISI) interval for the Crystal River Nuclear Plant, Unit 3 (CR-3). In RR 07-001-PT, the licensee proposed to perform a system leakage test conducted at or near the end of each inspection interval of American Society of Mechanical Engineers Code, Class 1 pressure retaining components in reactor coolant pressure boundary vent, drain, and branch lines and small bore connections (1 inch), with both isolation valves closed that would exclude a small segment of Class 1 line from the test boundary.
Based on the information provided in RR No. 07-001-PT, the U.S. Nuclear Regulatory Commission (NRC) staff concludes that the licensees compliance with the applicable ISI code would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Additionally, the NRC staff concludes in the enclosed safety evaluation that the licensees proposed alternative provides reasonable assurance of structural integrity.
Therefore, pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 50.55a (a)(3)(ii), the NRC staff authorizes the ISI program alternative proposed in RR No. 07-001-PT for the fourth 10-year ISI interval for CR-3.
The bases for the NRC staffs conclusions are contained in the enclosed Safety Evaluation. If you have any questions regarding this issue, please contact Farideh Saba at (301) 415-1447 or farideh.saba@nrc.gov.
Sincerely,
/RA by E. Brown for/
Thomas H. Boyce, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
Safety Evaluation cc w/enclosure: See next page
Mr. Dale E. Young, Vice President July 31, 2008 Crystal River Nuclear Plant (NA1B)
ATTN: Supervisor, Licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 34428-6708
SUBJECT:
CRYSTAL RIVER NUCLEAR PLANT, UNIT 3 - RELIEF REQUEST NO. 07-001-PT RELATED TO THE FOURTH 10-YEAR INTERVAL INSERVICE INSPECTION ON SYSTEM PRESSURE TEST BOUNDARY (TAC NO. MD7738)
Dear Mr. Young:
By letter dated December 21, 2007, Florida Power Corporation (licensee) submitted Relief Request (RR) No. 07-001-PT, related to the fourth 10-year inservice inspection (ISI) interval for the Crystal River Nuclear Plant, Unit 3 (CR-3). In RR 07-001-PT, the licensee proposed to perform a system leakage test conducted at or near the end of each inspection interval of American Society of Mechanical Engineers Code, Class 1 pressure retaining components in reactor coolant pressure boundary vent, drain, and branch lines and small bore connections (1 inch), with both isolation valves closed that would exclude a small segment of Class 1 line from the test boundary.
Based on the information provided in RR No. 07-001-PT, the U.S. Nuclear Regulatory Commission (NRC) staff concludes that the licensees compliance with the applicable ISI code would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Additionally, the NRC staff concludes in the enclosed safety evaluation that the licensees proposed alternative provides reasonable assurance of structural integrity.
Therefore, pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 50.55a (a)(3)(ii), the NRC staff authorizes the ISI program alternative proposed in RR No. 07-001-PT for the fourth 10-year ISI interval for CR-3.
The bases for the NRC staffs conclusions are contained in the enclosed Safety Evaluation. If you have any questions regarding this issue, please contact Farideh Saba at (301) 415-1447 or farideh.saba@nrc.gov.
Sincerely,
/RA by E. Brown for/
Thomas H. Boyce, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosure:
Safety Evaluation cc w/enclosures: See next page Distribution:
PUBLIC RidsNrrDorlLpl2-2 RidsOgcRp Branch Reading RidsNrrLACSola RidsAcrsAcnw&mMailCenter AHiser RidsNrrPMFSaba RidsRgn2MailCenter RidsNrrAdes RidsNrrTOrf PPatrick ADAMS ACCESSION NO.: ML081750252 *by memo NRR-106 OFFICE LPL2-2/PM LPL2-2/PM LPL2-2/LA CSCB/BC* OGC(NLO) LPL2-2/BC NAME TOrf FSaba CSola AHiser JAdler TBoyce (EBrown for)
DATE 7/23/08 7/23/08 7/22/08 5/28/08 7/22/08 7/31/08 OFFICIAL RECORD
Florida Power Corporation Crystal River Nuclear Plant, Unit 3 cc:
Mr. R. Alexander Glenn Associate General Counsel (MAC-BT15A) Mr. Jon A. Franke Florida Power Corporation Director Site Operations P.O. Box 14042 Crystal River Nuclear Plant (NA2C)
St. Petersburg, Florida 33733-4042 15760 W. Power Line Street Crystal River, Florida 34428-6708 Mr. Michael J. Annacone Plant General Manager Senior Resident Inspector Crystal River Nuclear Plant (NA2C) Crystal River Unit 3 15760 W. Power Line Street U.S. Nuclear Regulatory Commission Crystal River, Florida 34428-6708 6745 N. Tallahassee Road Crystal River, Florida 34428 Mr. Jim Mallay Framatome ANP Ms. Phyllis Dixon 1911 North Ft. Myer Drive, Suite 705 Manager, Nuclear Assessment Rosslyn, Virginia 22209 Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Mr. William A. Passetti, Chief Crystal River, Florida 34428-6708 Department of Health Bureau of Radiation Control David T. Conley 2020 Capital Circle, SE, Bin #C21 Associate General Counsel II - Legal Dept.
Tallahassee, Florida 32399-1741 Progress Energy Service Company, LLC Post Office Box 1551 Attorney General Raleigh, North Carolina 27602-1551 Department of Legal Affairs The Capitol Mr. Daniel L. Roderick Tallahassee, Florida 32304 Vice President, Nuclear Projects &
Construction Mr. Craig Fugate, Director Crystal River Nuclear Plant (SA2C)
Division of Emergency Preparedness 15760 W. Power Line Street Department of Community Affairs Crystal River, Florida 34428-6708 2740 Centerview Drive Tallahassee, Florida 32399-2100 Mr. David Varner Manager, Support Services - Nuclear Chairman Crystal River Nuclear Plant (SA2C)
Board of County Commissioners 15760 W. Power Line Street Citrus County Crystal River, Florida 34428-6708 110 North Apopka Avenue Inverness, Florida 34450-4245 Mr. Stephen J. Cahill Engineering Manager Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, Florida 34428-6708
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOURTH 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM REQUEST FOR RELIEF NO. 07-001-PT FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER NUCLEAR PLANT, UNIT 3 DOCKET NO. 50-302
1.0 INTRODUCTION
By letter dated December 21, 2007, Florida Power Corporation (licensee) submitted Relief Request (RR) No. 07-001-PT, related to the fourth 10-year inservice inspection (ISI) interval for the Crystal River Nuclear Plant, Unit 3 (CR-3). In RR 07-001-PT, the licensee proposed to perform a system leakage test conducted at or near the end of each inspection interval of American Society of Mechanical Engineers (ASME) Code, Class 1 pressure retaining components in reactor coolant pressure boundary (RCPB) vent, drain, and branch (VTDB) lines and small bore connections (1 inch) with both isolation valves closed that would exclude a small segment of Class 1 line from the test boundary.
The licensees request for relief is based on hardship of making multiple entries into the containment for valve alignment and thus, exposing personnel to high radiation and the risk of failure due to single valve isolation.
2.0 REGULATORY REQUIREMENTS As required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), ISI of ASME Code Class 1, 2, and 3 components must 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).
In addition, according to 10 CFR 50.55a(a)(3)(ii), alternatives to the requirements of paragraph 50.55a(g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC, Commission), if an applicant demonstrates that the specified requirement 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 preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection (ISI) 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 ISI 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. The ISI Code of Record for the fourth 10-year inspection interval for CR-3 is the 2001 Edition through the 2003 Addenda of the ASME Code,Section XI.
3.0 TECHNICAL EVALUATION
System/Component(s) for Which Relief is Requested RCPB VTDB lines and small bore connections (1 inch).
ASME Code Requirements The 2001 Edition through the 2003 Addenda of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, Item Number B15.10, requires all Class 1 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with IWB-5220. This pressure test is to be conducted prior to plant startup following each reactor refueling outage. The pressure retaining boundary for the test conducted at or near the end of each inspection interval shall be extended to all Class 1 pressure retaining components per IWB-5222(b).
Licensees Request for Relief Relief is requested from performing the system leakage test of the pressure retaining components per IWB-5222(b) for portions of the ASME Code, Class 1 RCPB VTDB lines and connections with the inboard isolation valves closed, which would exclude a small segment of Class 1 line between each inboard and outboard isolation valve from test boundary.
Licensees Basis for Requesting Relief The VTDB lines and connections 1-inch nominal pipe size (NPS) and smaller off the RCPB are equipped with manual valves, which provide double valve isolation of the RCPB. The requirement to extend the system leakage test boundary for the leakage test conducted at or near the end of each inspection interval to the outboard valve on these VTDB connections results in a hardship without a compensating increase in the level of quality and safety.
Repositioning the inboard manual valves before and after the test will take considerable time and will result in an unnecessary increase in dose to plant personnel. Manual operation (opening and closing) of the VTDB valves is estimated to expose plant personnel to 0.5 man-roentgen equivalent man (rem) per test.
The 1-inch NPS and smaller VTDB connections are normally closed during plant operation. The outboard valves would only see pressure if the inboard valve is open or leaks by the seat. Seat leakage, although undesirable, is not indicative of a flaw in the pressure boundary. Furthermore, these valves are in close proximity to the main runs of pipe. The non-isolable portion of these VTDB connections is pressurized and VT-2 examined during the test. The VT-2 examination
performed each refueling outage extends to the outboard valve, even though it is not pressurized.
The CR-3 Technical Specifications for RCPB leakage monitoring requires appropriate actions, including plant shutdown if leakage exceeds specified limits. Based on the above criteria, Florida Power Corporation requests authorization to use the proposed alternative in lieu of the ASME Section XI, IWB-5222(b), requirement.
Licensees Proposed Alternative The RCPB VTDB lines and connections 1-inch nominal pipe size and smaller will be visually examined for leakage with the inboard isolation valves in the normally closed position during the system leakage tests [per ASNE Code IWB-5222(b)] at or near the end of each inspection. This visual examination test provides reasonable assurance of structural integrity.
4.0 STAFF EVALUATION The ASME Code,Section XI of Record requires that all Class 1 components within the reactor coolant system (RCS) boundary undergo a system leakage test at the end of each refueling outage and a system leakage test at or near the end of each inspection interval. In RR No. 07-001-PT, the licensee proposed an alternative to the requirement of the test for the RCPB VTDB lines which would cause some line segments to be excluded from the test boundary. The line segments include two manually operated valves separated by a short pipe that is connected to the RCS. The line configuration, as outlined, provides double-isolation of the RCS. Under normal plant operating conditions, the subject line segments would see RCS temperature and pressure only if leakage through the inboard valves occurs. For the licensee to perform the system leakage test at or near the end of inspection interval in accordance with the ASME Code [IWB-5222(b)], it would be necessary to manually open the inboard valves to pressurize the line segments. Pressurization by this method would preclude the RCS double valve isolation and may cause safety concerns for the personnel performing the examination.
Typical line/valve configurations are in close proximity of the RCPB main runs of pipe and thus, would require personnel entry into high radiation areas within the containment. Manual actuation (opening and closing) of these valves is estimated to expose plant personnel to 0.5 man-rem per test. The licensee has proposed to visually examine the isolation valves in the normally closed position for leaks and any evidence of past leakage during system leakage test after each refueling outage. Also, the RCS vent and drain connections will be visually examined with the isolation valves in the normally closed position during the system leakage test at or near the end of the inspection interval. These visual examination tests provide reasonable assurance of structural integrity. Therefore, the NRC staff believes that the licensees proposed alternative will provide reasonable assurance of structural integrity for the RCPB VTDB line segments while maintaining personnel radiation exposure to as low as reasonably achievable. The NRC staff has further determined that compliance to the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
5.0 CONCLUSION
The NRC staff has evaluated the licensees request for relief and determined that compliance to the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety and the licensees proposed alternative would provide reasonable assurance of structural integrity. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the proposed alternative in RR No. 07-001-PT is authorized for the fourth 10-year ISI interval of CR-3. 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.