ML13221A091
| ML13221A091 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 08/13/2013 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Heflin A Union Electric Co |
| Lyon C | |
| References | |
| TAC MF1745 | |
| Download: ML13221A091 (9) | |
Text
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 August 13, 2013 Mr. Adam C. Heflin Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 65251 SUB..CALLAWAY PLANT, UNIT 1 -REQUEST FOR RELIEF 13R-14, PROPOSED ALTERNATIVE REGARDING LEAKAGE TESTING OF REACTOR PRESSURE VESSEL HEAD FLANGE LEAKOFF LINES (TAC NO. MF1745)
Dear Mr. Heflin:
By letter dated May 2, 2013, as supplemented by letters dated May 6 and July 3, 2013, Union Electric Company (dba Ameren Missouri, the licensee) submitted "Proposed Alternative to ASME Section XI Requirements for Leakage Testing of Reactor Pressure Vessel Head Flange Leakoff Lines (Relief Request 13R-14)" for U.S. Nuclear Regulatory Commission (NRC) review and approval.
The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, paragraph IWC-5221, requires that the system leakage test be performed with lines pressurized to the system pressure prior to performing a VT -2 visual examination.
The licensee proposed to perform the system leakage test of the reactor pressure vessel (RPV) head flange leakoff lines at Callaway Plant, Unit 1, with the pressure head that results when the refueling cavity is filled to the normal refueling water level. The licensee requested authorization to use the proposed alternative pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 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. The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii).
Therefore, the staff authorizes use of the proposed alternative at Callaway Plant, Unit 1, until the end of the third 10-year inservice inspection interval, which is currently scheduled to end December 18,2014. The NRC staff provided verbal authorization for relief request 13R-14 during a teleconference with your staff on May 6, 2013.
A. Heflin -2 All other ASME Code,Section XI requirements for which relief was not specifically requested and authorized in the subject proposed alternative remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Sincerely, Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No.
Safety cc w/encl: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 13R-14 REGARDING LEAKAGE TESTING OF REACTOR PRESSURE VESSEL HEAD FLANGE LEAKOFF LINES UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. SO-483 By letter dated May 2,2013, as supplemented by letters dated May 6 and July 3,2013 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 13123A09S, ML13126A304, and ML 13186AOSO, respectively), Union Electric Company (dba Ameren Missouri, the licensee) submitted "Proposed Alternative to ASME Section XI Requirements for Leakage Testing of Reactor Pressure Vessel Head Flange Leakoff Lines (Relief Request 13R-14)" for U.S. Nuclear Regulatory Commission (NRC) review and approval.
The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, paragraph IWC-S221, requires that the system leakage test be performed with lines pressurized to the system pressure prior to performing a VT -2 visual examination.
The licensee proposed to perform the system leakage test of the reactor pressure vessel (RPV) head flange leakoff lines at Callaway Plant, Unit 1, with the pressure head that results when the refueling cavity is filled to the normal refueling water level. The licensee requested authorization to use the proposed alternative pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part SO, paragraph SSa(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. Verbal authorization to use the proposed alternative was given in a conference call between the NRC staff and licensee representatives on May 6, 2013 (ADAMS Accession No. ML13126A320).
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR Part SO, paragraph SSa(g)(4), Inservice 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 In service Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of Enclosure
-2 the components.
The regulations require that inservice examination of components and system pressure tests conducted during the first 1 O-year inspection interval and subsequent 10-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 SO.SSa(b) 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. Paragraph SSa(a)(3) of 10 CFR Part SO states, in part, that alternatives to the requirements of 10 CFR SO.SSa(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 concludes 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 SO.SSa(a)(3)(ii).
3.0 TECHNICAL EVALUATION 3.1 Licensee's Request for Alternative Components for which Relief is Being Requested ASME Code Class 2 RPV head flange O-ring leakoff lines BB-07S-BCB-3/4", BB-07S-BCB-1", BB-07S-BCB-2", BB-076-BCB-3/4", BB-076-BCB-1", BB-076-BCB-2", BB-077-BCB-3/8", and valves BBV0079, BBV0080, BBV0081, and BBHV8032.
ASME Code Requirements The inservice inspection (lSI) Code of record for Callaway Plant, Unit 1, third 10-year lSI interval that started on December 19,2004, and is currently scheduled to end on December 18, 2014, is the 1998 Edition through the 2000 Addenda of the ASME Code,Section XI. Paragraph IWC-2S00 of the ASME Code,Section XI, Table IWC-2S00-1, 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-S221 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 portions of the leak detection system, consisting of the accessible portions of the RPV head flange O-ring leakoff lines, using the VT-2 visual examination method. The test shall be conducted at ambient conditions 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 /> when the piping is subjected to the static pressure head that exists when the reactor cavity is filled.
By letter dated May 2, 2013, the licensee cited the following precedents: STP Nuclear Operating Company (STPNOC), Units 1 and 2, Third Inspection Interval Relief Request RR-ENG-3-10, "Request for Relief from ASME Section XI Code Requirements for Reactor Pressure Vessel Head Flange O-Ring Leakoff Lines Non-Destructive Examination," as approved by the NRC letter dated March 12, 2013 (ML12312A234). Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, "Request for Relief from the American Society of Mechanical Engineers (ASME) Code,Section XI, Reactor Vessel Head Flange Seal Leak Detection Piping-Relief Request No. 49," as approved by the NRC letter dated April 4, 2013 (ML13085A254). Comanche Peak Nuclear Power Plant, Second Inspection Interval Relief Request C-9, "Alternative Pressure Testing Requirements for the Reactor Pressure Vessel Flange Leak-off Piping," as approved by NRC letter dated December 19, 2011 (ML113110092). LaSalle County Station, Third Inspection Interval Relief Request 13R-08, "Request for Relief for Inservice Inspection Impracticality of Pressure Testing the RPV Head Flange Seal Leak Detection System," as approved by NRC letter dated January 30,2008 (ML073610587). Susquehanna Steam Electric Station, Third 10-Year Inservice Inspection (lSI) Interval Program Plan Request for Relief 3RR-07, "Exemption from Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System," as approved by NRC letter dated September 24,2004 (ML042680078).
Licensee's Basis for Requesting Relief (as stated) The [four] flange seal leakoff lines are essentially a leakage collection/detection system and would only function as a Class 2 pressure boundary in the event of failure of the O-rings that separate the lines from reactor coolant system operating pressure.
Any significant leakage due to this condition would be expected to clearly exhibit boric acid accumulation that would be discernible during the proposed alternate VT -2 visual examination that will be performed.
The static head developed with the leak detection line filled with water will allow for the detection of any gross indications in the line. NRC Staff Evaluation The subject lines collect potential leakage past the inner and the outer RPV head flange O-rings and conduct it to a temperature element where the elevated temperature of any leakage is sensed. During normal operation, one line collects potential leakage past the inner O-ring and the valve on the second line, which collects leakage past the outer O-ring, is closed. If there is
-4 significant leakage past the inner O-ring, the valve on the line collecting inner O-ring leakage would be closed, and the valve on the outer O-ring leakage line would be opened, allowing any potential leakage past the outer O-ring to travel to the temperature element. It is only during the condition when the inner O-ring is leaking and the valve on the inner O-ring leakage line is closed will there be any significant pressure in the inner O-ring leakage line. As long as there is no indication of leakage past the outer O-ring, the plant is not required to shut down. The subject piping is heavy walled schedule 160 stainless steel (SA-312, Type 304) with a plant design pressure of 2485 pounds per square inch gauge (psig), and includes four Class 2 valves. The reactor operating pressure is 2235 psig. The leakoff lines are open to the reactor coolant drain tank and are normally pressurized to the pressure in the reactor coolant drain tank, approximately 4 psig. By letter dated May 6,2013, the licensee stated that there has been no identified occurrence of leakage or other failures of the subject leakoff piping. By letter dated July 3, 2013, in response to the NRC staff's request for additional information (RAI) dated May 17, 2013 (ADAMS Accession No. ML13186A050), the licensee stated that the plant has no history of reactor pressure vessel flange O-ring leakage. The licensee identified several methods of pressurizing the subject lines to the system pressure, as required by ASME Code,Section XI, paragraph IWC-5221, prior to performing the required VT-2 visual examination.
These methods included:
modification of the RPV flange to install mechanical threads and a threaded plug into each leakoff line to establish a boundary for leakage testing; pressurizing the lines prior to removing the RPV head at the beginning of the refueling outage; and pressurizing the lines after installation of the RPV head at the end of the refueling outage. Modification of the RPV flange to install a threaded plug into each leakoff line would require a design modification to install mechanical threads into each leakoff line at the location of the reactor vessel flange. Threaded plugs would then have to be installed prior to the pressure test and removed after the test was complete.
The NRC staff concludes 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, and installation and removal of the plugs would present foreign material exclusion issues. Performing a leakage test during shutdown prior to the removal of the head for refueling activities would delay the shutdown, requiring the plant to be maintained in Mode 3 at normal operating temperature and pressure at higher decay heat loads, resulting in an incremental increase in core damage frequency.
The NRC staff concludes that this evolution would be contrary to site efforts to reduce the outage sequence/time to cold shutdown in order to minimize shutdown risk. Applying system pressure to the leakoff lines for the purpose of leakage testing with the RPV head installed after refueling would 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 concludes 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 NRC staff has reviewed these options and concludes that there is a hardship associated with each. The NRC staff concludes that performing the VT -2 visual examination while the subject lines are at ASME Code-required system pressure would present a hardship.
The licensee proposed to conduct a VT-2 visual examination of the leakoff 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 pressure at the RPV flange due to the refueling water level is approximately 10 psig. In response to the NRC staff's RAI, the licensee stated that to ensure the lines are clear of air prior to performance of the VT-2 examination, the surveillance instructions have been revised to include a flush of lines BB-075-BCB and BB-076-BCB, after which a 4-hour hold is required.
Per the surveillance, the flush of the lines is accomplished by attaching a drain hose to the flange downstream of valve BBV0081, and then flushing lines BB-075-BCB and BB-076-BCB each for the greater of either 5 minutes or until no air is seen coming from the drain hose. The NRC staff concludes that the procedure is adequate to produce a water-solid line where any leakage would be that of water that can 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 concludes that the flushing procedure is 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's concern is whether the proposed low test pressure would be sufficient to demonstrate the leak tightness of the leakoff lines. If a leakoff line has a large through-wall flaw, leakage would be evident under either a high or low pressure test condition.
However, a leak from a small, tight crack may not be evident in the 4-hour time when the piping is subjected to the low pressure of the refueling water head. The NRC staff notes that the subject piping is pressurized for several days during each refueling outage when the refueling cavity is filled; for the current refueling outage the VT-2 visual examination was performed after the lines were subject to pressure of the refueling water head for approximately 24 days. Any coolant leakage during either the present or a previous refueling outage would result in boric acid accumulation that would be evident during a VT-2 visual examination.
The NRC staff concludes that the VT-2 visual examination after the leakoff lines have been subjected to the refueling water head at approximately 10 psig pressure provides evidence of leak tightness, and provides evidence that the lines can transport potential O-ring leakage at approximately 4 psig to the temperature element. Therefore, the NRC staff concludes that the proposed procedure is acceptable.
The NRC staff recognizes that an opportunity exists to examine the inner O-ring leakoff line while at reactor operating pressure if there is leakage past the inner O-ring and the valve on the inner O-ring leakoff line is closed. In response to the NRC staff's RAI, the licensee stated that Control Room annunciator response procedure OTA-RK-00020 Addendum 580, "RV Flange Leakoff Temperature High," directs the isolation of the inner O-ring leakoff line upon indication of leakage. Since the valves isolating the leakoff lines are located outside the bioshield, they can be accessed and operated while the plant is at power. The licensee stated that the procedure is being revised to have Operations personnel perform a visual examination for leakage of the accessible portions of the line 10 minutes after line isolation.
Additionally, the licensee stated that an on-demand surveillance is being generated to perform a VT-2 examination on the accessible portion of the line after a minimum hold time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The pressurized portion of piping outside the bioshield is in a radiation field of approximately 5 milli-roentgen equivalent man (mrem)/hour and the time needed to perform the VT-2 examination would be less than five minutes. In its letter dated July 3, 2013, the licensee stated that approximately 22 feet of the pressurized piping inside the bioshield can be seen from the bioshield entrance with the examiner standing in a radiation field of 100 to 500 mrem/hour.
To examine the remaining piping inside the bioshield, personnel would have to enter a radiation field of greater than 16 rem/hour and would put the examiner in very close proximity to the cross-over loop piping. The NRC staff concludes that the high-radiation fields preclude examination of the remaining piping within the bioshield walls. The NRC staff is satisfied that this procedure to examine the inner O-ring leakoff line while it is pressurized by inner O-ring leakage provides reasonable assurance of leak tightness while maintaining ALARA practices and, therefore, is acceptable.
Based on evaluation of past performance, as well as the service conditions, materials of construction, and the precautions taken to prevent buildup of contaminants, the NRC staff concludes that that service-induced degradation is unlikely.
The NRC staff further concludes that if any significant leakage were to occur in the leakoff 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 concludes that the proposed low test pressure, although not as effective as a high test pressure, will provide reasonable assurance of the leak tightness of the subject leakoff lines, and demonstrates that the leakoff lines can perform their intended function.
The NRC staff also concludes 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 concludes that the licensee's proposed alternative, "Proposed Alternative to ASME Section XI Requirements for Leakage Testing of Reactor Pressure Vessel Head Flange Leakoff Lines (Relief Request 13R-14)," provides reasonable assurance of 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 Callaway Plant, Unit 1, until the end of the third 1 O-year lSI interval, which is currently scheduled to end December 18, 2014. All other ASME Code,Section XI requirements for which relief was not specifically requested and authorized in the subject proposed alternative remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor:
J. Wallace, NRRlDE/ESGB Date: August 13, 2013 A. Heflin All other ASME Code.Section XI requirements for which relief was not specifically requested and authorized in the subject proposed alternative remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Docket No.
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