ML080280073
| ML080280073 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/22/2008 |
| From: | Mark Kowal NRC/NRR/ADRO/DORL/LPLI-1 |
| To: | Balduzzi M Entergy Nuclear Operations |
| Boska J, NRR, 301-415-2901 | |
| References | |
| TAC MD6831 | |
| Download: ML080280073 (8) | |
Text
February 22, 2008 Mr. Michael A. Balduzzi Sr. Vice President & COO Regional Operations, NE Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 - RELIEF REQUEST (RR) NO. RR-3-43 FOR TEMPORARY NON-CODE REPAIR OF SERVICE WATER PIPE (TAC NO. MD6831)
Dear Mr. Balduzzi:
By letter dated September 27, 2007, as supplemented by letters dated October, 3, 5 and 10, 2007, Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted a relief request for Nuclear Regulatory Commission (NRC) approval. Specifically, the licensee is proposing a temporary non-code repair to an American Society of Mechanical Engineers Boiler and Pressure Vessel Code Class 3 piping elbow in the service water system at Indian Point Nuclear Generating Unit No. 3. To support the licensees repair schedule, verbal authorization of the subject relief request was granted on October 11, 2007.
Based on the information provided in the licensees submittal, the NRC staff concludes that pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(ii), the licensees proposed temporary non-code repair of the elbow in the service water system is acceptable.
The NRC safety evaluation is provided in the enclosure.
If you have any questions regarding this approval, please contact the Indian Point Project Manager, John Boska, at (301) 415-2901.
Sincerely,
/RA/
Mark G. Kowal, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286
Enclosure:
Safety Evaluation cc w/encl: See next page
ML08
STurk MKowal DATE 1/30/08 1/30/08 1/8/08 2/12/08 2/22/08
Indian Point Nuclear Generating Unit No. 3 cc:
Mr. Michael R. Kansler President & CEO / CNO Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. John T. Herron Sr. Vice President Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Sr. Vice President Engineering & Technical Services Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. Fred R. Dacimo Site Vice President Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Mr. Anthony Vitale - Acting General Manager, Plant Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway P.O. Box 249 Buchanan, NY 10511-0249 Mr. Oscar Limpias Vice President Engineering Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. Joseph P. DeRoy Vice President, Operations Support Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. John A. Ventosa GM, Engineering Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. John F. McCann Director, Nuclear Safety & Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Ms. Charlene D. Faison Manager, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Ernest J. Harkness Director, Oversight Entergy Nuclear Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Mr. Patric W. Conroy Director, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Mr. Robert Walpole Manager, Licensing Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P. O. Box 249 Buchanan, NY 10511-0249 Mr. William C. Dennis Assistant General Counsel Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
Indian Point Nuclear Generating Unit No. 3 cc:
Mr. Paul Tonko President and CEO New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. John P. Spath New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector=s Office Indian Point 3 U. S. Nuclear Regulatory Commission P.O. Box 59 Buchanan, NY 10511 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Raymond L. Albanese Four County Coordinator 200 Bradhurst Avenue Unit 4 Westchester County Hawthorne, NY 10532 Mayor, Village of Buchanan 236 Tate Avenue Buchanan, NY 10511 Mr. William DiProfio PWR SRC Consultant 48 Bear Hill Road Newton, NH 03858 Mr. Garry Randolph PWR SRC Consultant 1750 Ben Franklin Drive, 7E Sarasota, FL 34236 Mr. William T. Russell PWR SRC Consultant 400 Plantation Lane Stevensville, MD 21666-3232 Mr. Jim Riccio Greenpeace 702 H Street, NW Suite 300 Washington, DC 20001 Mr. Phillip Musegaas Riverkeeper, Inc.
828 South Broadway Tarrytown, NY 10591 Mr. Mark Jacobs IPSEC 46 Highland Drive Garrison, NY 10524 Mr. Sherwood Martinelli FUSE USA via email
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF NO. RR-3-43 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
1.0 INTRODUCTION
By letter dated September 27, 2007, as supplemented by letters dated October 3, 5 and 10, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession Numbers ML072910355, ML072890132, ML072970095, and ML072970096, respectively), Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted relief request RR-3-43 for Nuclear Regulatory Commission (NRC) approval. Specifically, the licensee is proposing a temporary non-code repair to an American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 3 piping elbow in the service water system at Indian Point Nuclear Generating Unit No. 3 (IP3). To support the licensees repair schedule, verbal authorization of the subject relief request was granted on October 11, 2007.
2.0 REGULATORY EVALUATION
As specified in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), inservice inspection of nuclear power plant components shall be performed in accordance with the requirements of ASME Code,Section XI, except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Pursuant to 10 CFR 50.55a(a)(3),
alternatives to the requirements of paragraph (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. As stated in 10 CFR 50.55a(g)(5)(iii),
if the licensee has determined that conformance with certain Code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in §50.4, information to support the determinations.
The information provided by the licensee in support of its relief request has been evaluated by the NRC staff, and the bases for disposition are documented below.
3.0 TECHNICAL EVALUATION
3.1 Licensees Relief Request RR-03-43 3.1.1 Components for Which Relief is Requested The affected pipe is the 18-inch diameter service water supply pipe, line number 408, which supplies cooling water to three of the containment building fan cooler units (FCUs). This line is one of two lines which supplies Hudson River water to the five FCUs which are used to remove heat from the containment building atmosphere during normal plant operation and following a design-basis accident.
3.1.2. Applicable ASME Code Edition and Addenda The applicable Code of Record for the current 10 year inservice inspection interval is the ASME Code, Section Xl, 1989 Edition with no Addenda. However, for ASME Code Repair and Replacement activities, Entergy requested and the NRC approved, in a letter dated April 24, 2007 (ML070880358), the use of subsection IWA-4000 of the ASME Code,Section XI, 2001 Edition through the 2003 Addenda.
The affected portion of the service water piping was designed and constructed in accordance with the requirements of the USAS B31.1.0, 1967 Edition of the Power Piping Code.
3.1.3 Applicable Code Requirements IWA-4422.1 requires that defects be removed or reduced to an acceptable size prior to implementing a repair or replacement in accordance with the requirements of IWA-4000. Since the current through-wall defects are beyond the acceptance criteria of IWD-3000 and removal is not practical without system depressurization, the proposed repair method would not be consistent with IWA-4422.1.
3.1.4 Reason for Request On September 18, 2007 a nuclear plant operator conducting a routine plant walkdown noted minor leakage of approximately 5 drops per minute in one of the two cement-lined 18" diameter, 0.375" nominal thickness, service water supply lines for the containment building FCUs. As a result of this leak a volumetric examination of the surrounding area was performed and the results were evaluated (IP-CALC-07-00083) against the requirements of ASME Code Case N-513-1. Although this evaluation confirmed that the affected piping remains within the requirements of Code Case N-513-1, the calculated corrosion rate does not support continued structural integrity through the remainder of the current operating cycle.
A weld repair/replacement fully compliant with the requirements of IWA-4000 is not practical.
The affected piping section would need to be removed from service which would result in three FCUs inoperable. IP3 Technical Specification 3.6.6 does not have a condition statement for that configuration, so it would require IP3 to be shut down.
Entergy has evaluated alternative options for repairing this degraded area including weld overlay using ASME Code Case N-661 or an approach using a welded reinforcing plate. The weld overlay based on Code Case N-661 does not have a high probability of success due to the risk of "burn-through" in small areas where the remaining pipe thickness is insufficient to deposit weld metal. To protect against "burn-through" as shown in Electric Power Research Institute testing, a modified approach for weld overlay may be possible by placing a small intermediate plate over the localized area subject to "burn-through" and then the weld overlay could be applied over that plate. Both the reinforcing plate option and the overlay-with-intermediate-plate option could be designed to adequately restore the required structural margin for the remainder of the current operating cycle. The welded reinforcing plate is the preferred option because less welding will result in lower residual shrinkage stresses. Therefore, the balance of the discussion provided in this relief request is directed at describing the welded reinforcing plate approach.
3.1.5 Proposed Alternative and Basis for Use As discussed above, IWA-4422.1 requires that a defect be removed prior to implementing an IWA-4000 repair. However, this would require the pipe to be isolated and drained, which would require IP3 to be shut down. The preferred alternative proposed under this relief request would install a reinforcing plate over the degraded area to allow the attachment welding to be located in an area with minimal degradation therefore ensuring a structurally sound load path while minimizing the risk of "burn-through" and increased leakage.
The design will also ensure that the configuration of the repair will allow continued monitoring of the region by volumetric examination to ensure that future degradation will not adversely impact the structural capability of the repaired section.
- 1. Materials and Installation:
The material of the component to be repaired is concrete lined carbon steel, ASTM International (ASTM) designation A-234, Grade WPB. The proposed reinforcing material to be installed is ASTM A-234, Grade WPB/A-106 or equivalent carbon steel material with an ASME Code stress allowable of 15,000 pounds per square inch (psi). The welding process to be used in this repair is shielded metal arc welding (SMAW) with a carbon steel, 7018 weld wire. The reinforcing material would either be plate stock rolled to fit the contour of the affected repair area or a section from pipe will be used to fit the contour. The gap between the repair area and the reinforcing material will be controlled by procedure.
The welding will be performed per the requirements of ASME Code,Section XI using qualified welders and the weld procedure will be qualified in accordance with ASME Code,Section IX.
The weld procedure specifies 50 oF pre-heat for welds less than 3/4 inch thickness and no post weld heat treatment required for P-1 materials less than 3/4-inch thick.
- 2. Design Parameters:
The welded plate/weld repair option will be designed and installed consistent with the original USAS B31.1.0, 1967 Edition of the Power Piping Code requirements for a reinforcing plate (paragraph 104.3). A structural evaluation (IP-CALC-07-00209) has been performed to ensure that the resulting stresses in the piping, the plate and the attaching welds do not exceed the allowable stresses of the USAS B31.1.0 Code, 1967 Edition. The repair material will be carbon steel or pipe equivalent to the existing pipe material with allowable stress of S = 15,000 psi.
The ASME Code Case N-513-1 evaluation used the required factors of safety of 2.77 for the normal / upset condition and 1.39 for the emergency / faulted condition.
For purposes of this repair design and monitoring, Entergy will assume that the cement lining is no longer present in the area of the planned repair so that the corrosion rate for unprotected carbon steel will be applied.
- 3. Non Destructive Examinations:
The area to be repaired has been characterized by performing straight beam ultrasonic testing (UT) mapping (Report IP3-UT-07-1 10) of the region to bound the degraded area and to ensure that the welds for repair are located in areas of sound base metal. At least 1/2 inch of the weld for attaching the reinforcing plate to the elbow will be performed in an area of average wall thickness exceeding 0.18 inches to ensure a structurally sound load path around the perimeter of the repair area. Nondestructive examination (NDE) of this area was also performed in March 2007 (Report IP3-UT-07-049) when a through-wall flaw was discovered during startup from refueling outage 3R14. Plant conditions at that time allowed for a weld repair consistent with ASME Code,Section XI, IWA-4422.1, so that a relief request was not needed. Four areas with thickness readings less than 0.110 inches were excavated and weld repaired in accordance with the requirements of ASME Code,Section XI. Corrective action at that time also included developing plans for replacing this elbow at the next refueling outage (3R15, spring 2009).
The pipe wall was repaired to a minimum wall thickness needed to support operation until the next refueling outage, based on nominal corrosion rate assumptions. The typical unprotected metal corrosion rate for service water crevice corrosion observed at Indian Point is 0.024 inches per 2-year cycle (0.012 inches per year). This is based on the wear rates observed and calculated for the evaluation of previous service water piping degradations. However, corrosion rates could be higher in localized areas.
The location of the March 2007 repair with respect to the current area of interest is adjacent to grid location H6 asshown on the UT map in IP3-UT-07-110. A final assessment of why a new through-wall leak developed near the area of the prior repair has not been completed at this time. Further characterization of the degradation in this elbow will be accomplished when the component is replaced.
Prior to shutdown for 3R13 (March 2005) radiography of this elbow performed for the IP3 NRC Generic Letter 89-13 Corrosion Monitoring Program identified an area of interest on the opposite side of the elbow from the current flaw. Localized UT performed during 3R13 identified a 0.25-inch diameter area in the weld with a thickness less than 0.135 inches. An ASME Code, Section Xl repair was implemented prior to startup from that outage. There is no historical UT data resulting from the March 2005 repair for the current area of interest.
NDE inspections for the extent-of-condition review will also be performed as stated in Section E.5 of the licensees submittal. NDE related to the repair and inservice monitoring is discussed in Section E.4 of the licensees submittal.
- 4. Repair Monitoring:
During installation of the reinforcing plate, the welds will be examined, consistent with the requirements described in ASME Code Case N-661. This includes performing a surface examination of the area to be welded, a surface examination (dye penetrant or magnetic particle) after the first weld pass and a final surface examination of the completed weld.
Inservice monitoring of the repair will be accomplished by applying a 1-inch by 1-inch grid over the area which will cover the reinforcing plate and the flat portion of the attaching weld (refer to Figure A of the licensees submittal). The intersection points in the grid will be inspected using straight beam UT. An initial baseline UT will be performed after installing the repair.
Subsequent UTs will then be performed to verify that the structural requirements of the original construction code are maintained through the remainder of the current operating cycle. The UTs will be performed monthly for the first quarter and if no unexpected degradation is identified, UTs will then be performed quarterly for the balance of the duration of this relief request. To determine unexpected degradation, UT of the repair plate, attaching weld and surrounding base material (i.e. elbow) will be performed and an average corrosion rate will be calculated based on the point to point comparison between the current and the previous inspections. If this average corrosion rate exceeds the predicted corrosion rate (i.e. 0.012" per year) it will be considered unexpected. The inspection results discussed above will be evaluated as required by ASME Code Case N-513-1 to ensure that the structural margins required by the code case are maintained through the remainder of the current operating cycle.
If the results of the monitoring program indicate that the structural margins required by the code case will be exceeded prior to the end of the current fuel cycle, Entergy will implement additional repair and/or replacement activities prior to reaching the limits of the code case. These repair and/or replacement activities will be consistent either with (1) the requirements of this relief request or (2) the requirements of the ASME Code, Section Xl, sub-section IWA-4000. NRC approval will be requested prior to the performance of any additional non-code repair.
Also, routine walkdowns will be performed by nuclear plant operators at least daily. This piping is not insulated and is accessible for visual inspection.
- 5. Degradation mechanism:
Based on the location of the defect and based on the UT inspections of the degraded area, Entergy concludes that this defect was likely caused by degradation of the protective concrete lining directly under the degraded area which allowed brackish water from the Hudson River to contact the unprotected carbon steel piping resulting in localized corrosion. The degradation of the concrete lining was likely caused by the high flow velocities and turbulence from the valve located just upstream of the degraded area. Further evaluation of the degradation mechanism will be performed during the next outage as stated in Section F of the licensees submittal, when the elbow can be removed and replaced. Entergy will perform augmented inspections, as required by ASME Code Case N-513-1, for the extent-of-condition evaluation. The inspections will be at five locations selected as most susceptible to the degradation mechanism suspected at this time. Parameters to be considered for selection of the augmented inspection locations will include system operating conditions, proximity of upstream valves, and years of service.
- 6. Applicable Loads:
The repair will be designed to accommodate all appropriate deadweight, pressure, and seismic loads. Since the system is a moderate energy system which operates at a low temperature, differential thermal expansion between the repair plate and the repaired component is not a concern.
3.1.6 Duration of Proposed Alternative:
The duration of the temporary repair is limited until the next scheduled outage exceeding 30 days, but no later than the next refueling outage, currently scheduled for the spring of 2009.
3.2
NRC Staff Evaluation
On September 18, 2007, during a routine plant walkdown with IP3 operating at 100% power, the licensee found a piping elbow leaking in the service water system. The observed leakage was small at approximately 5 drops per minute. The leaking elbow was located in one of the two cement-lined 18-inch diameter service water supply lines for the containment FCUs. The subject piping has a nominal wall thickness of 0.375 inch. The licensee proposed to perform a temporary non-code repair. The licensee stated in its submittal that to perform a repair in accordance with the ASME Code requirements is not practical because the code requires the defects to be removed from the leaking component. However, to remove the defects from the degraded component, the affected piping section would need to be removed from service which would result in three FCUs inoperable. The IP3 technical specification does not allow plant operation in this configuration. Therefore, a plant shutdown is required in order to perform a code repair. The NRC staff finds that requiring the licensee to implement a code repair in this case would impose an undue hardship upon the licensee without a compensating increase in the level of quality and safety, taking into consideration the alternative that the licensee has proposed. Furthermore, the shutdown and restart would unnecessarily cycle plant systems and components which would reduce the safety margin for plant operation.
In discussing the degradation mechanism, the licensee attributed the observed piping degradation to the degradation of the protective concrete lining directly under the degraded area. The loss of the protection from the concrete lining would allow the brackish water from the Hudson River to contact the unprotected carbon steel piping resulting in localized corrosion.
The licensee believes that the degradation of the concrete lining was caused by the high flow velocities and turbulence from the valve located just upstream of the degraded area. The licensee will perform further evaluation of the degradation mechanism during the next outage when the degraded elbow is removed from service.
In lieu of performing a code repair, the licensee proposed a temporary non-code repair as an alternative. The service duration of the proposed non-code repair is limited to the next scheduled outage exceeding 30 days, but no later than the next refueling outage currently scheduled for the spring of 2009. The licensees proposed non-code repair is based on the welded reinforcing plate approach. This approach consists of welding a reinforcing plate with a nominal thickness of 1/2 inch to the degraded elbow, which had a nominal thickness of 0.375 inch. This welded reinforcing plate will contain the leakage and restore the wall thickness of the degraded component. The proposed reinforcing material is ASTM A-234, Grade WPB/A-106 or equivalent carbon steel material with an ASME Code stress allowable of 15000 psi. This material is similar to the material used for the degraded component. The welding process to be used in this repair is SMAW with a carbon steel 7018 weld wire. The licensee stated that the welding will be performed in accordance with the requirements of ASME Code,Section XI, using qualified welders and the weld procedure will be qualified in accordance with the ASME Code,Section IX. For welding of carbon steel materials with thickness less than 3/4 inch thick, the code specifies a pre-heat of 50 degrees F and requires no post weld heat treatment. The NRC staff finds the materials and the welding process to be used in the proposed repair are acceptable because they comply with the applicable code requirements.
The licensee stated that the welded plate and the attachment weld will be designed and installed consistent with the original USAS B31.1.0, 1967 Edition of the Power Piping code requirements for a reinforcing plate. The licensee performed a structural evaluation to ensure the resulting stresses in the piping, the plate and the attachment welds do not exceed the code allowable stresses. In the licensees structural evaluation, the safety factor of 2.77 for the normal/upset condition and 1.39 for the emergency/faulted condition was applied. The licensees structural evaluation also assumed a nominal corrosion rate of 0.012 inch per year for the potential corrosion of the unprotected carbon steel. The NRC staff finds the proposed repair design acceptable because the repair was designed in accordance with the original construction code requirements.
In a response to the NRC staffs request for additional information (RAI) regarding how the nominal corrosion rate was determined and the results of local corrosion rate, the licensee stated that the referenced nominal corrosion rate was developed over several years of evaluating service water piping systems and corrosion degradation at IP2 and IP3. However, the licensee has not developed the value for the local corrosion rate. The NRC staff notes that the local corrosion rate is expected to be higher than the assumed nominal corrosion rate, which may lead to leakage in the local area. However, the NRC staff finds the use of the assumed nominal corrosion rate in the design of the non-code repair is acceptable considering the proposed repair monitoring. The implementation of the proposed repair monitoring will provide reasonable assurance that the structural integrity of the non-code repair is maintained.
The proposed temporary non-code repair consists of attaching a reinforcing plate, with approximate dimensions of 6 inches by 10 inches by 1/2 inch thick, by a half-inch weld to the elbow area with an average wall thickness around the perimeter of the plate exceeding 0.18 inches to ensure a structurally sound load path around the perimeter of the repair area. The licensees original design of the attachment weld is a fillet weld. However, the NRC staff had a concern regarding its inspectability because the configuration of the fillet weld will not allow a proper volumetric examination by ultrasonic technique. As a result of the NRC staffs concern, the licensee increased the width of the attachment weld to at least 1 inch from the lower edge of the plate to provide sufficient surface area for performing a proper UT examination of the attachment weld and the areas surrounding the attachment weld.
The licensees original inservice monitoring of the repair consisted of UT examination (straight beam UT) of the reinforcing plate and the flat portion of the attachment weld on the intersection points of a 1-inch by 1-inch grid. In a response to the NRC staffs concern regarding potential corrosion of the base materials, the licensee enhanced the monitoring program to include a 3-inch band of the base materials, to the extent practical, surrounding the repair. A baseline ultrasonic examination of the areas discussed above will be performed after installation of the repair. In addition, an average corrosion rate will be calculated based on the point to point comparison between the current and previous inspections. The UT examinations will be performed monthly for the first quarter and if no unexpected degradation is identified, UT will then be performed quarterly for the balance of the duration of this relief request. In a response to the NRC staffs request for additional information regarding what constitutes unexpected degradation, the licensee stated that if the average corrosion rate exceeds the predicted corrosion rate (0.012 inch per year) then the inspection frequency will remain monthly, otherwise, the inspection frequency will be performed on a quarterly basis through the remainder of the current operating cycle. The NRC staff also requested details regarding what inspection results would require additional corrective action such as increased frequency of UT monitoring, or additional repair or elbow replacement. The licensee stated that if the results of the monitoring program indicate that the structural margins required by the ASME Code Case N-513-1 will be exceeded prior to the end of the current fuel cycle, additional repair and/or replacement activities prior to reaching the limits of the code case will be implemented. In addition to the periodic UT, the licensee will also conduct a routine daily walkdown inspection of the subject piping. This piping is not insulated and is accessible for visual inspection.
Therefore, any leakage from the subject piping will be detected during the walkdown inspection.
The NRC staff finds the licensees repair design and repair monitoring program as described in its submittal, and discussed above, are acceptable because they follow the requirements of the construction code and will provide reasonable assurance that the structural integrity of the repaired elbow will be maintained during the current cycle of operation.
4.0 CONCLUSION
Based on the above review, the NRC staff concludes that the licensees proposed temporary non-code repair of a degraded piping elbow as stated in relief request RR-03-43 is acceptable.
This is based on the consideration that requiring the licensee to implement an ASME Code repair in this case would impose a hardship on the licensee without a compensating increase in the level of quality and safety. The licensees proposed non-code repair as discussed above will provide reasonable assurance that the structural integrity of the repaired elbow will be maintained during the current cycle of operation. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the use of the proposed non-code repair in lieu of the ASME Code requirements at IP3, with the duration of the temporary repair limited to the next scheduled outage exceeding 30 days, but no later than the next refueling outage currently scheduled for the spring of 2009.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: William H. Koo Date: February 22, 2008