ML14252A076
| ML14252A076 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/23/2014 |
| From: | Robert Pascarelli Plant Licensing Branch II |
| To: | Pierce C Southern Nuclear Operating Co |
| Williams S | |
| References | |
| TAC MF4206, TAC MF4207 | |
| Download: ML14252A076 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 23, 2014
SUBJECT:
JOSEPH M. FARLEY, UNITS 1 AND 2, (FNP-ISI-ALT-16, VERSION 1)
ALTERNATIVE TO INSERVICE INSPECTION FOR REACTOR VESSEL FLANGE LEAKOFF LINES (TAG NOS. MF4206 AND MF4207)
Dear Mr. Pierce,
By letter dated May 28, 2014, as supplemented on August 7;-2014, Southern Nuclear Operating Company, Inc., (SNC) requested approval to use an alternative to the American Society of Mechanical Engineers (ASME)Section XI code requirement of subarticle IWC-5220 for leakage testing of the Class 2 reactor vessel flange leakoff lines for Joseph M. Farley Nuclear Plant, (FNP) Units 1 and 2, applicable to the fourth 1 0-year inservice inspection interval.
The application was submitted pursuant to Sections 50.55a(a)(3)(ii) of Title 10 of the Code of Federal Regulations* (1 0 CFR), the licensee requested to use an alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff has reviewed the subject request, and concludes that SNC has adequately addressed all of the regulatory requirements and that the proposed alternative provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes the licensee's proposed alternative in accordance with 10 CFR 50.55a (a)(3)(ii). The NRC staff's safety evaluation is enclosed.
If you have any questions, please contact the Project Manager, Shawn Williams, at 301-415-1009 or via e-mail at Shawn.Williams@nrc.gov.
Docket Nos. 50-348, 50-364
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE TO ASME CODE REQUIREMENTS FOURTH TEN-YEAR INSERVICE INSPECTION PROGRAM INTERVAL SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348. 50-364
1.0 INTRODUCTION
By letter dated May 28, 2014, as supplemented by letter dated August 7, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14148A492 and ML14219A381 respectively), Southern Nuclear Operating Company, Inc. (the licensee),
submitted proposed alternative FNP-ISI-ALT-16, Version 1 for U.S. Nuclear Regulatory Commission (NRC) review and authorization. The licensee proposes to perform the system leakage test of the reactor pressure vessel (RPV) flange leak-off lines at Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, using the pressure developed when the refueling cavity is filled to the normal refueling water level in lieu of the pressure required by American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, paragraph IWC-5221. The licensee requested authorization to use the proposed alternative pursuant to Title 10 of the Code of Federal Regulations Part 50 (1 0 CFR 50) Paragraph 55a(a)(3)(ii) on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), lnservice Inspection Requirements, ASME Code Class 1, 2, and 3 components (including supports) must 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 lnservice 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 firsl1 0-year inspection interval and subsequent 1 0-year inspection 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 inspection interval, subject to the limitations and modifications listed therein.
Enclosure Paragraph 55a(a)(3) of 10 CFR 50 states, in part, that alternatives to the requirements of 10 CFR 50.55a(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.
Based on analysis of the regulatory requirements, the NRC staff finds that the regulatory authority exists to authorize the licensee's proposed alternative on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff has reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(a)(3)(ii).
3.0 TECHNICAL EVALUATION
3.1 The License's Request Components for which Relief is Being Requested Code Class 2 Reactor Pressure Vessel Flange Leak-off Piping NPS 3/8", 3/4" and 1" ASME Code Requirements The Code of Record for the Farley, Units 1 and 2, Fourth lnservice Inspection Interval, which began December 1, 2007, and is scheduled to end on November 30, 2017, is the ASME Code,Section XI, 2001 Edition through the 2003 Addenda.
Paragraph IWC-2500 of the ASME Code,Section XI, Table IWC-2500-1, Category C-H, Item Number C7.1 0, requires that a system leakage test with a VT -2 visual examination of Class 2 pressure retaining components be performed each inspection period. Paragraph IWC-5221 requires that the system leakage test be conducted at the pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability.
Licensee's Proposed Alternative The licensee proposes to examine the Class 2 portion of the leak detection system consisting of the accessible portions of the RPV head flange o-ring leak-off piping once each inspection period. The leak-off piping shall be examined using the VT-2 visual examination method and will be performed by certified VT -2 examiners. The test shall be conducted at ambient conditions after the refueling cavity has been flooded to its minimum water level for refueling operations of 23 feet above the top of the RPV flange for at least four (4) hours. A static pressure of approximately 10 pounds per square inch, gauge (psig) is expected at a 23 foot depth in borated water.
3.2
NRC Staff Evaluation
The subject lines are part of the reactor coolant leak detection system. The design function of the leak detection system is to alert the operators of minor leakage that may develop during normal operation. The piping does not perform a safety function and is not required to operate during or after a seismic event.
The RPV flange and head at Farley, Units 1 and 2, are sealed by two hollow, metallic o-rings.
Seal leakage detection is provided by two leak-off connections: one between the two o-rings and one outboard of the outer o-ring. These connections provide a path for potential leakage through the leak-off piping to a thermal detector where the elevated temperature of any leakage is sensed and an alarm in the control room is tripped. The leak-off piping is only pressurized to reactor operating pressure when the inner o-ring is leaking and the valve on the inner o-ring leakage line is closed.
The subject leak-off lines are 3/8" tubing (SA-213, Grade 304) and 3/4" and 1" pipe (schedule 160 stainless steel, SA-376, Type 304 or 316). Both the tubing and pipe have a plant design pressure of 2485 psig at 650 °F. The reactor operating pressure is 2235 psig. The common header for the leak-off lines is routed to the reactor coolant drain tank. In the licensee's August 7, 2014, response to the NRC staffs request for additional information, the licensee states that the plant has only experienced one incident of RPV flange inner o-ring leakage in its previous years of operation and no leakage from the leak-off line.
The licensee has identified several methods of pressurizing the subject lines to the system pressure required by ASME Code,Section XI, paragraph IWC-5221, prior to performing the required VT-2 visual examination. These methods include: modification of tt1e RPV flange to install mechanical threads and installation.of a threaded plug into each leak-off line to establish a boundary fo( leakage testing to allow testing with the RPV head off; pressurizing the lines with the RPV head in place which still require design modifications to allow pressurization; and intentionally failing both the inner and outer o-rings. None of these options were considered to be viable by the licensee.
To perform the leakage test with the RPV head off would require a design change and modification of the.RPV flange to allow installation of mechanical threads into each leak-off line at the RPV flange to allow installation of a threaded plug into each leak-off line. The plugs would then have to be installed prior to the pressure test and removed after the test was complete. The NRC staff finds that performing the modification, as well as installation and removal of the plugs for each leakage test, would result in significant radiological dose, which would be contrary to As Low As Reasonably Achievable (ALARA) considerations. Furthermore, installation and removal of the plugs could present foreign material exclusion issues.
To perform the system leakage test with the head in place would also require a design change and modification to replace valves, install a plug flange for the outer leak-off and install vent piping. Applying system pressure to the leak-off lines for the purpose of system leakage testing with the RPV head installed would also require pressurizing the lines with a hydrostatic test pump in the direction opposite to the intended design function of the o-rings. The NRC staff finds that such pressurization could unseat the installed o-rings, likely resulting in the need to replace the o-rings which would require depressurizing and removal of the reactor vessel head.
The licensee states that removal and reinstallation of the head to replace the o-rings would be accompanied by an additional 3.8 roentgen equivalent man (REM) radiological dose. The NRC staff finds that this evolution would also be contrary to ALARA considerations.
To perform the system leakage test by intentionally failing the inner and outer a-rings would require a second removal and reinstallation of the RPV head during outages in which the test was performed. As stated above, removal and reinstallation of the RPV head to replace the a-rings would require an additional 3.8 REM radiological dose. To perform this activity each period of the 1 0-year interval would require an added 11.4 REM.
The NRC staff has reviewed these options and finds that there is a hardship associated with each. The NRC staff concludes that performing the VT-2 visual exa.mination while the subject lines are at ASME Code-required system pressure would present a hardship without a compensating increase in the level of quality and safety.
The licensee is proposing to conduct a VT-2 visual examination of the leak-off lines after the refueling cavity has been filled to its normal refueling water level for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The static pressure at the RPV flange due to the refueling water level is approximately 10 psig. The licensee states that in order to purge the lines of air prior to performing the VT-2 visual examination, water will be allowed to flow from the piping for approximately five minutes, to ensure the piping is water solid prior to the beginning of the 4-hour pressurization hold time.
The NRC staff finds that the procedure is adequate to produce a water-solid line where any leakage would be detected during a VT-2 visual examination. The NRC staff also notes that the flushing procedure will clear the lines of contaminants that could promote stress corrosion cracking. Therefore, the NRC staff finds the flushing procedure acceptable.
The NRC staff notes that the system leakage test requirements of the ASME Code, IWC-5220 are focused on demonstrating leak tightness rather than structural integrity. The NRC staff notes that the subject piping is pressurized for several days during each refueling outage when the refueling cavity is filled. Any coolant leakage during either the present or a previous refueling outage would result in ~oric acid accumulation that would be evident during a VT-2 visual examination. The NRC staff finds that the VT-2 visual examination after the leak-off lines have been subjected to the refueling water head at approximately 1 0 psig pressure provides evidence of leak tightness, and provides evidence that the lines can transport potential a-ring leakage to the thermal detector. Therefore, the NRC staff finds that the proposed alternative is acceptable.
The NRC staff finds, based on evaluation of past performance, as well as the service conditions, and materials of construction, that service induced degradation is unlikely. The NRC staff further finds that if any significant leakage were to occur in the leak-off line during the time of pressurization during each refueling outage, boric acid accumulation would be discernible during a subsequent visual examination. The NRC staff therefore finds that the proposed low test pressure will provide reasonable assurance of the leak tightness of the subject leak-off lines, and will demonstrate that the leak-off lines can perform their intended function. The NRC staff also finds that requiring compliance with the system leakage test pressure requirements would result in a hardship without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative, FNP-lSl-ALT-16, Version 1, provides reasonable assurance of structural integrity and leak tightness, and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) and therefore authorizes use of the proposed alternative at Joseph M.
Farley Nuclear Plant, Units 1 and 2, during the fourth 1 0-year In service Inspection Interval which began December 1, 2007, and is scheduled to end on November 30, 2017.
All other requirements of the ASME Code for which relief has not been specifically requested and authorized remain applicable, including a third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: Keith M. Hoffman, NRRIDE/EPNB
- via e-mail OFFICE LPLII-1/PM LPLII-1/LA DE/EPNB/BC LPLII-1/BC LPLII-1/PM NAME SWilliams SFlg_ueroa DAiley*
A Pascarelli SWilliams DATE 09/22/14 09/18/14 09/08/14 09/23/14 09/23/14