ML12271A366

Relief Request Number RR-2-4 Regarding Fourth 10-Year Interval Inservice Inspection Program

ML12271A366
Person / Time
Site:	Kewaunee
Issue date:	10/04/2012
From:	Istvan Frankl Plant Licensing Branch III
To:	Heacock D Dominion Energy Kewaunee
Feintuch K
References
TAC ME8503
Download: ML12271A366 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 4, 2012 David A. Heacock Dominion Energy Kewaunee, Inc.

Innasbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

KEWAUNEE POWER STATION - ONE-TIME RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE FOR THE FOURTH TEN-YEAR INSERVICE INSPECTION PORGRAM INTERVAL (TAC NO.

ME8503)

Dear Mr. Heacock:

By letter dated May 3, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12125A279), as supplemented by letter dated May 4,2012 (ADAMS Accession No. ML12129A279), Dominion Energy Kewaunee, Inc. (hereinafter referred to as the licensee) submitted request RR-2-4 to the Nuclear Regulatory Commission (NRC) for a temporary deviation from certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI requirements, for the fourth ten-year inservice inspection program interval at the Kewaunee Power Station. This letter documents the authorization granted by verbal means on May 5, 2012, of the repairs accomplished as a result of request RR-2-4.

Paragraph 55a(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50 states that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by NRC, if

0) 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.

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) and, therefore, authorizes the use of the proposed alternative "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR 50.55a, Request No. RR-2-4" at Kewaunee Power Station until May 12,2012, while the plant was in MODE 5 and MODE 4 only. 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.

D. Heacock - 2 If you have any questions, please contact the Project Manager, Karl Feintuch at 301-415-3079 or via e-mail at KarI.Feintuch@nrc.gov Sincerely,

!/JL Istvan Frankl, Acting Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE RR-2-4. TEMPORARY REPAIR OF RESIDUAL HEAT REMOVAL PIPING DOMINION ENERGY KEWAUNEE, INC KEWAUNEE POWER STATION DOCKET NO. 50-305

1.0 INTRODUCTION

By letter dated May 3, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML12125A279), as supplemented by letter dated May 4, 2012 (ADAMS Accession Number ML12129A279), Dominion Energy Kewaunee, Inc. (hereinafter referred to as the licensee) submitted proposed alternative "Inservice Inspection Program Fourth Ten-Year Interval Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Request No. RR-2-4" for temporary alternate repair of residual heat removal (RHR) piping. Specifically, the licensee requests relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, to allow installation of two mechanical clamps on the 3/4-inch diameter pipe segment at valve RHR-600, a portion of the containment pressure boundary, and to relieve the examination extent and frequency requirements. The licensee proposes this alternative during the spring 2012 refueling outage at the Kewaunee Power Station (KPS) to create the conditions necessary to accomplish a Code-compliant repair of the leaking pipe segment. The licensee states that conformance with the ASME Code requirements would constitute a hardship as the result of radiological dose and has requested authorization of their proposed alternative by the Nuclear Regulatory Commission (NRC) staff pursuant to 10 CFR 50 55a(a)(3)(ii), hardship without a compensating increase in the level of quality or safety.

The licensee discovered a leak in the socket weld at valve RHR-600 during the KPS refueling outage in spring 2012. By letter dated April 29, 2012, the licensee submitted the proposed alternative "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR 50.55a, Request No. RR-2-3" (ADAMS Accession Number ML12120A004) for use of a mechanical clamp at the socket weld to stop the leakage and allow performance of a Code-compliant repair. The proposed alternative was subsequently reviewed by the NRC staff and a verbal authorization for its use was given on April 30, 2012 (ADAMS Accession Number ML12214A134). Execution of the authorized repair procedure caused a second leak. As a result, the initial proposed alternative was withdrawn (ADAMS Accession Number ML12125A300) and replaced with the Enclosure

-2 present proposed alternative (ADAMS Accession Number ML 1212SA279) that envelops the initial proposed alternative for the mechanical clamp and extends the repair activity to include installation of a second mechanical clamp to stop the leak that resulted from performance of the first repair. For purposes of administrative clarity, both proposed mechanical clamps are simultaneously examined in the present safety evaluation.

Verbal authorization of the subject proposed alternative, RR-2-4, was given in a telephone conversation between the NRC staff and licensee representatives on May S, 2012 (ADAMS Accession Number ML12212A046). The second mechanical clamping device and structural restraint were installed, and KPS transitioned to MODE 4 on May 6, 2012. Both mechanical clamps and the structural restraint were subsequently removed from the subject RHR segment and a Code-compliant repair was successfully performed on May 6, 2012.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR, Part SO, paragraph SSa{g){4), Inservice Inspection Reql,lirements, ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except for the design and access provisions and the pre-service examination requiremElnts, set forth in the ASME Code,Section XI, "Rules for Inservice 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 first Ten-year inspection interval and subsequent Ten-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 SO states 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 the above, the NRC staff finds that the regulatory authority exists to authorize an alternative to the AS ME Code requirements. Accordingly, the NRC staff has reviewed and evaluated the licensee's request pursuant to 10 CFR SO.SSa(a}(3)(ji).

3.0 TECHNICAL EVALUATION

3.1 Licensee's Request for Alternative 3.1.1 Applicable Code Edition and Addenda The inservice inspection (lSI) Code of Record for Kewaunee Power Station fourth Ten-year lSI interval, which began on June 16, 2004 and is scheduled to end on June 1S, 2014, is the 1998 Edition through the 2000 Addenda of Section XI of the ASME Code.

- 3 3.1.2 Components Affected ASME Code,Section XI, Code Class 2, RHR system, 314-inch sockolet and 314-inch schedule 40 pipe to valve RHR-600.

3.1.3 Code Requirements ASME Boiler and Pressure Vessel Code,Section XI, 1998 Edition through 2000 Addenda, IWA-4133 states that mechanical clamping devices used to replace piping pressure boundary shall meet the requirements of ASME Code,Section XI, Appendix IX. Article IX-1000(c){2}

states that clamping devices shall not be used on portions of a piping system that form the containment boundary. Article IX-6000 requires a plan for monitoring defect growth in the area immediately adjacent to the clamping device and requires the area immediately adjacent to the clamping device to be volumetrically examined.

3.1.4 Licensee's Reason for Request KPS was in a refueling outage with the reactor refueled and the reactor vessel reassembled and the plant in MODE 5, Cold Shutdown with the RHR system in operation, when a leak was discovered at a socket weld in a 3/4-inch line that is common to both RHR loops, rendering both RHR loops inoperable. During activities to install a leak-limiting device over this leak, a welder inadvertently created a small through-wall perforation in the 3/4-inch line. Both leaks exist in a line that is common to both RHR loops and neither leak can be repaired without removing both RHR loops from service. KPS Technical Specification 3.4.7, "RCS [reactor coolant system]

Loops - MODE 5, Loops Filled," requires one RHR loop to be operable and in operation; and either one additional RHR loop shall be operable, or the secondary side water level of at least one steam generator shall be greater than or equal to 5 percent. In order to remove the RHR loops from service and perform the repair. the plant needed to transition from MODE 5 to MODE 4 which is not permitted when the RHR system is not operable.

3.1.5 Licensee's Proposed Alternative and Basis for Use The licensee is proposing to perform a temporary alternate repair of the RHR piping by maintaining the installed leak-limiting device (mechanical clamp) over the socket weld leak and installing a second mechanical clamp and its associated structural restraint, that serves as an added measure of safety to prevent a catastrophic separation of the 3/4-inch line, on the through-wall perforation that occurred during installation of the first mechanical clamp. These activities will ensure that structural integrity and leak tightness of the RHR system are maintained prior to proceeding from MODE 5 to MODE 4. This temporary alternate repair will remain in place until the section of pipe containing the leaks is isolated. Once in MODE 4, isolating and repairing the portion of the RHR system with the leaks will take approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, thereby eliminating the need for the leak-limiting devices.

The licensee states that there are currently three known precedents (ADAMS Accession Numbers 9609180177 (Legacy Library). ML052090182, and ML052070047).

- 4 3.2 NRC staff Evaluation The licensee has identified one Code-compliant method to repair the section of pipe containing the leaks. This method involves changing modes from the plant's current condition, MODE 5, to MODE 6 and offloading the core. This process would permit draining the reactor coolant system sufficiently to allow the replacement of the leaking weld and perforated pipe. This approach is Code-compliant but involves heavy lifting, moving nuclear fuel, and the licensee estimates that this process would result in a personnel radiation dose of 8 REM. The NRC staff finds that the licensee's estimates of radiation dose are reasonable. Based on as low as reasonably achievable considerations, as well as the inherent risks involved in heavy lifting and moving nuclear fuel, the NRC staff finds that use of this Code-compliant repair method would constitute a hardship. The NRC staff is unaware of other Code-compliant options for performing the necessary repair.

As an alternative, the licensee proposes to change modes from the plant's current condition, MODE 5, to MODE 4 and isolate the affected section of RHR piping in order to perform the repair. No heavy lifts or other hazardous evolutions would be required, and the dose estimation for this method is 350 milli-REM. The NRC staff notes that RHR-600 is credited with maintaining the containment boundary and ASME Code,Section XI, Appendix IX, IX-1000(c)(2) prohibits the use of a clamping device on portions of a system that form the containment boundary. Therefore, the plant cannot install the temporary modification in order to change to MODE 4 and perform Code-compliant repairs without prior NRC approval.

The licensee proposes to install the first mechanical clamp, consisting of a leak repair enclosure incorporating a strong back design, at the degraded socket weld to support its structural integrity.

The mechanical clamp is designed to meet the requirements of Article IX of the ASME Code,Section XI, with exceptions: use of the mechanical clamp on the containment boundary (Article IX-1000); and defect growth monitoring plan and volumetric examination requirements (Article IX-6000).

To support the structural integrity of the degraded weld, the licensee states that the first mechanical clamp is designed and fabricated to the requirements of ASME Code,Section XI.

The NRC staff has reviewed the licensee's Temporary Modification Package 2012-11 (Revision 4), Enclosure 1 of the submittal. In the design of the mechanical clamp, the licensee assumed the socket weld has a 360 degree, 100 percent through-wall flaw. In order to transmit the longitudinal loads normally transmitted by the piping, the licensee proposes to install a fillet weld between the clamp and the 3f4-inch pipe to ensure that the 3f4-inch pipe will not eject from the degraded socket weld. The NRC staff finds the assumed circumferential crack size to be a bounding assumption. Since the structural clamp is designed to support the full pressure loading of the weld joint and associated 3f4-inch pipe, the NRC staff finds the licensee's design will provide reasonable assurance of structural integrity.

To support the leakage integrity of the degraded weld, the licensee proposes to inject a sealant into the mechanical clamp enclosure to seal the leak. The licensee states that the sealant has a low concentration of halogens (e.g., chlorides), which could degrade the stainless steel base material. The licensee will monitor the injection pressure and volume to ensure that sufficient volume is injected to seal the leak without injecting too much sealant and contaminating the RHR system. The NRC staff finds that the low chloride concentration of the sealant along with the short duration of application will limit the potential for stress corrosion cracking of the

- 5 stainless steel piping, and is therefore acceptable in this limited application. The NRC staff also finds that the licensee's controls on the sealant injection volume and pressure provide reasonable assurance that the leak is sealed and contamination of the RHR system with sealant does not occur.

AS ME Code,Section XI, Appendix IX, Article IX-6000 requires that a plan for monitoring defect growth in the area immediately adjacent to the clamping device be prepared. The licensee proposed not to perform this action. The NRC staff finds the monitoring plan is not necessary in this limited application as any potential degradation mechanism should not affect the structural integrity of the system and clamping device in the short time period that the device will be installed. In addition, the licensee has committed to perform visual examination of the degraded socket weld area every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the structural and leakage integrity of the temporary repair.

Article IX-6000(a) requires that the area immediately adjacent to the clamping device be examined using a volumetric method. Although the licensee initially proposed not to perform the volumetric examination, the welder inadvertently created a small through-wall perforation in the 314-inch pipe between the sockolet and the RHR-600 valve during the aotivities to install the sockolet weld leak-limiting device. To verify that the newly created perforation was not exacerbated by service related degradation (e.g., erosion or thinning), the licensee evaluated the accessible pipe surrounding this through-wall perforation using straight beam ultrasonic examination (UT) techniques and determined it to be at or near nominal thickness (0.113-inches). The NRC staff finds that this examination has met the requirements of IX-6000(a) to volumetrically examine the area immediately adjacent to the first clamping device.

Both the sockolet weld leak and the pipe perforation exist in the 3/4-inch line that is common to both RHR loops. In addition to the original mechanical clamp, the licensee proposed to install a mechanical clamp and a structural restraint bar at the location of the through-wall perforation.

Similar to the initial mechanical clamp, this second mechanical clamp consists of a leak repair enclosure incorporating a strong back design to support its structural integrity, with injection of sealant to provide for leak tightness. The structural restraint bar is designed to replace the fillet weld of the initial repair and support the full longitudinal load transmitted by the 3/4-inch line if a 360 degree, 100 percent through-wall flaw circumferential crack were to exist either in the degraded socket weld or in the pipe. The NRC staff notes that the second mechanical clamp with structural restraint bar fulfills the same criteria as that for the first mechanical clamp: a) the clamp is designed in accordance with ASME Code,Section XI, requirements; b) the clamp will be injected with sealant to curtail leakage; c) the clamp will be installed for a very short time; d) the licensee has committed to inspecting the clamps for leakage at a frequency of not less than one inspection in every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; and e) the structural restraint bar will carry the longitudinal loads normally transmitted by the piping. The NRC staff has reviewed the design of the second mechanical clamp and finds that there is no access for volumetric examination.

Thus, Article IX-6000(a) does not require volumetric examination of the area immediately surrounding the clamping device. Based on these considerations, the NRC staff finds that the use of the two mechanical clamps with injected sealant and the structural restraint bar provide reasonable assurance of structural integrity and leak tightness.

In order to complete the Code-compliant repair in MODE 4, the subject socket weld must remain dry during the welding activity. In response to the NRC staff's request for additional information (RAI, ADAMS Accession Number ML12129A279), the licensee has established a detailed plan

-6 to identify and monitor leakage of the isolating valves that could prevent completion of the Code-corn pliant repair of the affected weld/piping while in MODE 4. The licensee will monitor leakage through the isolating valves prior to beginning work to ensure the water level in the 10-inch RHR line will be at least 3 inches below the socket for welding activities. The system will not be breached until the system is drained and a stable water level is verified using UT, with monitoring performed at nominal 15 minute intervals for a minimum of one hour. The licensee states that mock-up training of the repair weld with the welders assigned to perform the weld has shown that final welding should take less than 30 minutes. Once the system is breached, water level will be monitored at nominal 15-minute intervals for the period of time that the system is breached. If unexpected leakage is encountered after the system is breached up to the time for insertion of the new pipe for final welding, then water from the 10-inch line will be removed by pumping it out using a tube inserted through the socket. The NRC staff is satisfied that the actions proposed by the licensee will adequately ensure that the repair operation will not be interrupted by leakage through the isolation valves prior to completion of the Code compliant repair.

In case of unexpected leakage that would interfere with welding after the system is breached, the licensee has fabricated and staged two plugging devices at the repair location that can be used to stop the unexpected leakage and restore RHR availability, and permit the plant to safely transition to MODE 6, followed by defueling and a permanent ASME Code r¢pair. In response to the NRC staff's RAI, the licensee states that the temporary plugging devices are a safety measure to address the development of boundary leakage after the system nas been breached; the plugs are not designed to facilitate welding while in MODE 4. The NRC staff finds that the licensee's leakage identification and monitoring plan, and contingencies to temporarily plug the pipe after has been breached, provide adequate assurance that MODE 6 can be safely attained if unforeseen circumstances arise. Thus the NRC staff finds these measures acceptable. .

The NRC staff finds that the licensee's proposed temporary repair, conSisting of two temporary mechanical clamps that are designed and fabricated to the requirements of ASME Code,Section XI, with a structural restraint bar to prevent ejection if the socket weld or pipe had a 360 degree, 100 percent through wall crack, provides adequate expectation of structural integrity, and the injection sealant provides adequate expectation of leak tightness for the time envisioned for the plant to transition from MODE 5 to MODE 4 and accomplish the Code compliant repair. In addition, the NRC staff finds that the licensee has adequately considered potential difficulties involving execution of the proposed alternative and planned alternative actions necessary to safely transition the plant to MODE 6 and accomplish a Code-compliant repair should unforeseen difficulties arise. Therefore, the NRC staff finds that it is acceptable to apply the mechanical clamps to the subject RHR piping, a part of the containment boundary, in order to transition to MODE 4. The NRC staff also finds that complying with the specified ASME Code requirements would result in a hardship or unusual difficulty.

4.0 CONCLUSION

As set forth above, the NRC staff finds that the proposed alternative will provide reasonable assurance that the structural integrity and leak tightness of the degraded socket weld and perforated pipe will be maintained during MODE 5 and the expected time to transition to MODE 4 and accomplish the Code-compliant repair, 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

-7 adequately addressed all of the regulatory requirements setforth in 10 CFR SO.SSa(a)(3)(ii) and, therefore, authorizes the use of the proposed alternative "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR SO.SSa, Request No. RR-2-4" at Kewaunee Power Station until May 12, 2012, while the plant is in MODE S and MODE 4 only.

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.

S.O REFERENCES

1. Letter dated May 3,2012, "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR SO.SSa, Request No. RR-2-4," ADAMS Accession Number ML 1212SA279

2. Letter dated May 4,2012, "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR SO.SSa Request No. RR-2-4, Response to Request for Additional Information,"

ADAMS Accession Number ML12129A279

3. Letter dated April 29, 2012, "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR SO.SSa, Request No. RR-2-3," ADAMS Accession Number ML12120A004

4. Verbal Authorization of "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR SO.SSa Request No. RR-2-3," dated April 30, 2012, ADAMS Accession Number ML12214A134 S. Letter dated May 3,2012, "Inservice Inspection Program Fourth Ten-Year Interval Withdrawal of 10 CFR SO.SSa Request No. RR-2-3," ADAMS Accession Number ML 1212SA300

6. Verbal Authorization of "Inservice Inspection Program Fourth Ten-Year Interval 10 CFR SO.SSa, Request No. RR-2-4," dated May S, 2012, ADAMS Accession Number ML12212A046

7. Letter dated September 13, 1996, "Approval of a Relief Request from the Requirements of 10 CFR SO.SSa Related to the Repair of a Sockolet Weld - Kewaunee Nuclear Power Plant (TAC No. M96273)," ADAMS Legacy Library Accession Number 9609180177

8. Letter dated August 1S, 200S, "Turkey Point Nuclear Plant, Unit 4 - Safety Evaluation for Relief Request Regarding Mechanical Clamping Device on Pressure Boundary Piping (TAC No. MC7338)," ADAMS Accession Number MLOS2090182

9. Letter dated August 9, 200S, "James A. Fitzpatrick Nuclear Power Plant - Relief Request for Temporary Non-Code Repair of a Shutdown Cooling Pipe (TAC No. MC7S44),"

ADAMS Accession Number MLOS2070047 Principal Contributor: Jay Wallace, NRRlDE/EPNB Date: October 4, 2012

D. Heacock - 2 If you have any questions, please contact the Project Manager, Karl Feintuch at 301-415-3079 or via e-mail at KarI.Feintuch@nrc.gov Sincerely, IRA!

Istvan Frankl, Acting Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosure:

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