ML15037A005
| ML15037A005 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/13/2015 |
| From: | Eric Oesterle Plant Licensing Branch IV |
| To: | Reddemann M Energy Northwest |
| George A | |
| References | |
| TAC MF3562 | |
| Download: ML15037A005 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Mark E. Reddemann Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352-0968 February 13, 2015
SUBJECT:
COLUMBIA GENERATING STATION - REQUEST FOR ALTERNATIVE 31Sl-15 TO THE REQUIREMENTS OF THE ASME CODE (TAC NO. MF3562)
Dear Mr. Reddemann:
By letter dated March 7, 2014, as supplemented by letter dated August 21, 2014, Energy Northwest (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission for the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI requirements at Columbia Generating Station (CGS).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(ii),
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 paragraph headings in 10 CFR 50.55a were changed by Federal Register notice dated November 5, 2014 (79 FR 65776), which became effective on December 5, 2014 (e.g., 10 CFR 50.55a(a)(3)(i) is now 50.55a(z)(1), and 50.55a(a)(3)(ii) is now 50.55a(z)(2)). See the cross-reference tables, which are cited in the notice, in the Agencywide Documents Access and Management System (ADAMS) at Accession No. ML14015A191 and package Accession No. ML14211A050.
The licensee proposed to perform the system leakage test of the reactor pressure vessel (RPV) flange leak-off lines at CGS using the pressure developed when the refueling cavity is filled to the normal refueling water level, as an alternative to the pressure required by ASME Code,Section XI, paragraph IWB-5221. This proposed methodology is identical to that presented in ASME Code Case N-805, "Alternative to Class 1 Extended Boundary End of Interval or Class System Leakage Testing of the Reactor Vessel Head Flange 0-Ring Leak-Detection System."
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that Entergy Northwest has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2), and that the proposed alternative provides reasonable assurance of the structural integrity of the RPV flange leak-off lines. Therefore, the NRC staff authorizes the use of alternative 31Sl-15 at CGS for the duration of the third 10-year inservice inspection interval ending on December 12, 2015.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this request for alternative remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
If you have any questions regarding this matter, Andrea George of my staff may be reached at (301) 415-1081 or via e-mail at andrea.george@nrc.gov.
Docket No. 50-397
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE 31Sl-15 REGARDING TESTING OF THE REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF LINES ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By letter dated March 7, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14077A173), as supplemented by letter dated August 21, 2014 (ADAMS Accession No. ML14245A058), Energy Northwest (the licensee), requested an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI for system leakage testing of the reactor pressure vessel (RPV) flange leak-off lines at Columbia Generating Station (CGS). Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 O CFR) 50.55a(a)(3)(ii), the licensee requested to use the proposed 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.
In request for alternative 31Sl-15, the licensee proposed to perform the system leakage test of the RPV flange leak-off lines using the pressure developed when the refueling cavity is filled to the normal refueling water level in lieu of the pressure required by ASME Code,Section XI, paragraph IWB-5220, "System Leakage Test," subparagraph IWB-5221 (a). The licensee stated that this proposed methodology is identical to that presented in ASME Code Case N-805, "Alternative to Class 1 Extended Boundary End of Interval or Class System Leakage Testing of the Reactor Vessel Head Flange 0-Ring Leak-Detection System." ASME Code Case N-805 is not yet approved by the U.S. Nuclear Regulatory Commission (NRC) for use and is not included in the current revision (Revision 17) of Regulatory Guide 1.147, "lnservice Inspection Code Case Applicability, ASME Section XI, Division 1, dated August 2014 (ADAMS Accession No. ML13339A689).
The paragraph headings in 10 CFR 50.55a were changed by Federal Register notice dated November 5, 2014 (79 FR 65776), which became effective on December 5, 2014 (e.g.,
10 CFR 50.55a(a)(3)(i) is now 50.55a(z)(1 ), and 50.55a(a)(3)(ii) is now 50.55a(z)(2)). See the cross-reference tables, which are cited in the notice, at ADAMS Accession No. ML14015A191 and ADAMS package Accession No. ML14211A050.
Enclosure
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a(g)(4), "lnservice inspection standards requirement for operating plants," require that ASME Code Class 1, 2, and 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 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 in-service examination 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 conditions listed therein.
The regulations in 10 CFR 50.55a(z) state, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) 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 the above, and subject to the following technical evaluation, the NRC staff concludes that regulatory authority to authorize an alternative to the ASME Code requirement, as requested by the licensee, exists. Accordingly, the NRC staff has reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(z)(2).
3.0 TECHNICAL EVALUATION
3.1 The License's Request
Applicable Code Edition and Addenda
The ASME Code of record for the third 10-year inservice inspection (ISi) interval at CGS, which began on December 13, 2005, and is scheduled to end on December 12, 2015, is the ASME Code,Section XI, 2001 Edition through the 2003 Addenda.
Components for which Relief is Being Requested ASME Code Class 1, 1-inch carbon steel RPV flange leak-off piping and fittings originating from reactor vessel nozzle N-17 and 3/4-inch RPV flange leak-off branch piping.
Applicable ASME Code Requirements The ASME Code,Section XI, IWB-2500, Table IWB-2500-1, Code Category B-P, Item Number B 15.10 requires that all Class 1 pressure retaining components be subject to a system leakage test with a visual VT-2 examination each refueling outage. Subparagraph IWB-5221 (a) requires that the system leakage test be conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power.
Reason for Request (as stated by the licensee)
The RPV flange seal leak detection is separated from the reactor pressure boundary by one passive membrane, which is an 0-ring, located on the vessel flange. A second 0-ring is located on the opposite side of the tap in the vessel flange (See Figure 1 [of the licensee's application]). This piping is required during plant operation in order to detect failure of the inner flange seal 0-ring.
Failure of the 0-ring would result in the annunciation of an alarm in the Control Room (See Figure 2 [of the licensee's application]). Failure of the inner 0-ring is the only condition under which this line is pressurized. Therefore, the line is not expected to be pressurized during the system pressure test following a refueling outage.
The configuration of this piping precludes system pressure testing while the vessel head is removed because the configuration of the vessel tap, coupled with the high test pressure requirement, prevents the tap in the flange from being temporarily plugged or connected to other piping. The opening in the flange is smooth walled, making the effectiveness of a temporary seal very limited.
Failure of this [temporary) seal could possibly cause ejection of the device used for plugging or connecting to the reactor vessel.
The configuration also precludes pressure testing with the vessel head installed because the seal prevents complete filling of the piping, which has no vent available. The top head of the vessel contains two grooves that hold the 0-rings.
The 0-rings are held in place by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test was performed with the head on, the inner 0-ring would be pressurized in a direction opposite to what it would see in normal operation. This test pressure would result in a net inward force on the inner 0-ring that would tend to push it into the recessed cavities that house the retainer clips. The thin 0-ring material would very likely be damaged by this inward force.
Purposely failing or not installing the inner 0-ring in order to perform a pressure test would require replacing the new outer and possibly the new inner 0-ring each time the test is conducted. This would result in additional time needed during the outage and additional radiation exposure to personnel associated with the removal and reinstallation of the RPV head.
Licensee's Proposed Alternative In lieu of the pressure requirements of IWB-5221 (a), the licensee proposes to perform a VT-2 examination during the next refueling outage in spring 2015 of the accessible portions of the RPV head flange leak-off piping, while subject to static pressure head with the RPV head removed and the refueling cavity 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 head developed with the leak-off lines filled with water will allow for detection of pressure boundary failures. In its letter dated August 21, 2014, the licensee stated that the normal refueling water level is 24.5 feet above the top of the RPV flange.
3.2
NRC Staff Evaluation
The NRC staff has evaluated the proposed alternative 3181-15 pursuant to 10 CFR 50.55a(z)(2).
In its review, the NRC staff focused on whether compliance with the specified requirements of 1 O CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, and if there is a compensating increase in the level of quality and safety despite the hardship.
Due to the existing design and configurations of the RPV flange and flange leak-off line, system leakage testing of the leak-off line piping at a pressure corresponding to 100 percent rated reactor power, either before or after the RPV head is removed, would be unusually difficult because it would require the licensee to either pursue temporary system modifications or permanent design changes as described in Section 3.1 of this safety evaluation. These modifications could introduce foreign materials into the reactor, either through ejection of a temporary seal or welding and grinding activities in close proximity to the RPV, which would introduce a Foreign Material Exclusion program concern. Personnel carrying out these tasks would incur additional radiation dose, which would be an as low as reasonably achievable (ALARA) program concern. To conduct the required system testing with the RPV head in place, the licensee would have to install new test connections and valves to existing piping to be able to externally pressurize the leak-off piping and provide a vent path. Personnel performing these tasks and conducting testing would be exposed to additional dose, which would be an ALARA program concern. Furthermore, when the RPV head is on, if the subject piping was externally pressurized for the purpose of conducting the ASME Code-required system leakage testing of the RPV flange leak-off detection system piping, the inner 0-ring seal may fail due to being pressurized in the opposite direction than its designed purpose. Requiring the RPV head to be removed and reinstalled to perform the pressure test by purposely failing or not installing the 0-ring would also result in additional radiation exposure to personnel involved with the removal and reinstallation of the RPV head, which would be an ALARA program concern.
Based on the evaluations above, the NRC staff concludes that complying with the ASME Code,Section XI, IWB-5221 (a) requirement for leak testing of the RPV head flange seal leak-off line piping at a pressure corresponding to 100 percent of rated reactor power would result in a hardship and unusual difficulty if imposed upon the licensee. Therefore, the NRC staff determines that the above items constitute a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The licensee proposed to perform a VT-2 examination during the next refueling outage in spring 2015 of the accessible portions of the RPV head flange leak-off piping, while subject to static pressure head with the RPV head removed and the refueling cavity 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 />, in lieu of the current ASME Code requirements in IWB-5221(a).
The static head developed with the leak-off lines filled with water will allow for detection of pressure boundary failures. In a request for additional information (RAI) dated July 11, 2014 (ADAMS Accession No. ML14195A028), the NRC staff requested additional information regarding the static pressure head developed in the subject piping after the refueling cavity is filled to its normal refueling water level. In its RAI response dated August 21, 2014, the licensee stated, in response to RAl-2, that when the refueling cavity is filled to its normal refueling water level the minimum head pressure in the leak-off line is 24.5 feet. Based on the information provided by the licensee and the evaluation above, the NRC staff concludes that the licensee's proposed system leakage test of the RPV leak-off piping will subject the system to the highest pressure that can be obtained without design modifications to existing configurations of both the vessel flange face and the flange leak-off piping.
Any evidence of leakage from an existing flaw in the piping and its associated connections would be detected by the VT-2 examination while the piping is subject to the static head pressure described above. In an RAI dated July 11, 2014, the NRC staff requested information regarding accessibility of the leak-off piping and whether insulation is installed on the piping. In its response dated August 21, 2014, the licensee stated that the leak-off piping is not behind any walls or barriers inside containment, and that access is available so that a VT-2 examination can be performed on the piping up to the ASME Code,Section XI boundary valves.
The licensee also stated that there is an insulation panel covering portions of the subject piping, which would be removed for the examination. Furthermore, in its RAI response, the licensee stated that for any piping located in a high elevation or far away from the VT-2 examiner, the examiner would be able to use standard VT-2 tools to complete the examination and identify any pipe through-wall leakage.
In its application, the licensee provided information regarding RPV flange leak-off piping leakage detection capabilities (i.e., detection of increase in drywell temperature and pressure, detection of an increase in drywell floor drain leakage, containment radiation monitors as well as monitoring of drywell fission products) that provides warning to the control room operator in the unlikely event of a through-wall leak in the RPV flange seal leak-off line piping concurrent with leak or failure of the RPV flange inner seal. The licensee stated that if the proposed alternative (VT-2 examination) was not effective in identifying a through-wall leak, originating from an existing flaw in the subject piping, the plant's existing leakage detection capability would be able to identify the leakage during normal operation and the licensee would implement appropriate corrective actions. Furthermore, in its supplement dated August 21, 2014, the licensee stated that CGS has no history of 0-ring leakage or degradation in RPV head flange leak-off piping. If the inner 0-ring and leak off piping were no longer capable of withstanding system pressure, and/or the RPV flange leak-off leakage rates do not meet technical specification requirements, shutdown of the reactor would be required. Based on the information provided by the licensee and the evaluation above, the NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the subject RPV flange leak-off piping.
On the basis of the above evaluation, the NRC staff concludes that the proposed alternative in 31Sl-15 is acceptable for the remainder of the third 10-year ISi interval at CGS.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative 31Sl-15 provides reasonable assurance of the structural integrity and leak tightness of the RPV head flange leak-off piping, and that complying with the specified ASME Code 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(z)(2) and, therefore, authorizes the use of proposed alternative 31Sl-15 at CGS for the remainder of the third 10-year ISi interval, which began on December 13, 2005, and is scheduled to end on December 12, 2015.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this request for alternative remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: K. Hoffman, NRR Date: February 13, 2015
ML15037A005 OFFICE NRR/DORL/LPL4-1 /PM NRR/DORL/LPL4-1 /PM NAME SGoetz AGeorge DATE 2/12/2015 2/12/2015 Sincerely, IRA/
Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
- via email NRR/DORL/LPL4-1 /LA NRR/DE/EPNB/BC NRR/DORL/LPL4-1 /BC(A)
JBurkhardt*
DAiiey*
EOesterle 2/12/2015 11/21/2014 2/13/2015