ML080580577
| ML080580577 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 03/26/2008 |
| From: | Melanie Wong NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | Brandi Hamilton Duke Power Co |
| stang J, NRR/DLPM, 415-1345 | |
| References | |
| TAC MD6274 | |
| Download: ML080580577 (14) | |
Text
March 26, 2008 Mr. Bruce H. Hamilton Vice President McGuire Nuclear Station Duke Power Company LLC 12700 Hagers Ferry Road Huntersville, NC 28078
SUBJECT:
MCGUIRE NUCLEAR STATION, UNIT 1, REQUEST FOR RELIEF 06-MN-001, FOR THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN REGARDING REPAIR OF CHEMICAL VOLUME AND CONTROL SYSTEM VALVE 1NV-240 (TAC NO. MD6274)
Dear Mr. Hamilton:
By letter dated July 24, 2007, as supplement December 3, 2007, Duke Power Company LLC (the licensee), submitted Relief Request No. 07-MN-001, for its Third 10-Year Interval Inservice Inspection (ISI) Program Plan for McGuire Nuclear Station, Unit 1. The licensee proposed alternatives to the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), 1998 with the 2000 addenda, for valve 1NV-240 in the chemical volume and control system. The licensee submitted the relief request as a result of through-wall flaw discovered in the valve.
The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and, based on the information provided, concludes that compliance with the specified ASME Code requirements for valve 1NV-240 is impractical, and the proposed alternative provides an acceptable level of quality and safety. Therefore, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(a)(3)(i), the NRC staff authorizes the use of the proposed alternative for the third 10-year ISI interval at McGuire Nuclear Station, Unit 1. The enclosed Safety Evaluation contains the NRC staff's evaluation and conclusions.
Sincerely,
/RA/
Melanie C. Wong, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-369
Enclosure:
Safety Evaluation cc w/encl: See next page
Mr. Bruce H. Hamilton Vice President McGuire Nuclear Station Duke Power Company LLC 12700 Hagers Ferry Road Huntersville, NC 28078
SUBJECT:
MCGUIRE NUCLEAR STATION, UNIT 1, REQUEST FOR RELIEF 06-MN-001, FOR THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN REGARDING REPAIR OF CHEMICAL VOLUME AND CONTROL SYSTEM VALVE 1NV-240 (TAC NO. MD6274)
Dear Mr. Hamilton:
By letter dated July 24, 2007, as supplement December 3, 2007, Duke Power Company LLC (the licensee), submitted Relief Request No. 07-MN-001, for its Third 10-Year Interval Inservice Inspection (ISI) Program Plan for McGuire Nuclear Station, Unit 1. The licensee proposed alternatives to the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), 1998 with the 2000 addenda, for valve 1NV-240 in the chemical volume and control system. The licensee submitted the relief request as a result of through-wall flaw discovered in the valve.
The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and, based on the information provided, concludes that compliance with the specified ASME Code requirements for valve 1NV-240 is impractical, and the proposed alternative provides an acceptable level of quality and safety. Therefore, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(a)(3)(i), the NRC staff authorizes the use of the proposed alternative for the third 10-year ISI interval at McGuire Nuclear Station, Unit 1. The enclosed Safety Evaluation contains the NRC staff's evaluation and conclusions.
Sincerely,
/RA/
Melanie C. Wong, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-369
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
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- memo dated January 29, 2008 OFFICIAL RECORD COPY Date NRR/LPL2-1/PM NRR/LPL2-1/LA EMCB/SC*
OGC NLO w/comments NRR/LPL2-1/BC NAME JStang:sp MO=Brien TChan LBS MWong DATE 03/ 26 /2008 03/ 26 /2008 01/29/08 3/13/2008 03/ 26 /2008
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THIRD 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF NO. 07-MN-001 DUKE POWER COMPANY, LLC MCGUIRE NUCLEAR STATION, UNIT 1 DOCKET NO. 50-369
1.0 INTRODUCTION
By letter dated July 24, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072140851), as supplemented December 3, 2007 (ADAMS Accession No. ML073460046), Duke Power Company LLC, (the licensee) submitted Request for Relief 07-MN-001 requested use of a proposed alternative to the requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, for McGuire Nuclear Station, Unit 1 (McGuire 1)1998 edition, through the 2000 addenda as stipulated in paragraph IWC-3122.2 on the basis that compliance with the specified requirements would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The proposed alternative would delay a Code repair of degraded valve 1NV-240 to either during the end of cycle 19 refueling outage or a forced outage of sufficient duration during current operating cycle 19. Valve 1NV-240 has a through-wall flaw in the valve body and is leaking (1 drop per minute). The valve is part of the charging pump discharge line in the chemical and volume control (NV) system. The valve provides piping system isolation for maintenance, but it does not provide any active safety-related function.
In a subsequent inspection, the licensee did not identify any other through-wall leakage in the accessible NV system piping from charging pump suction to the containment isolation valves at both units.
2.0 REGULATORY REQUIREMENTS Inservice inspection (ISI) of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the Code and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(g), except where specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(g)(6)(i) states that relief from the requirements of paragraph (g) may be granted by the NRC, if granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if
the requirements were imposed on the facility. Section 50.55a(a)(3) states that alternatives to the requirements of the 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.
Pursuant to 10 CFR 50.55a(g)(4), 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 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 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) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ASME Code of record for the McGuire 1 third 10-year interval inservice inspection program, which began on December 1, 2001, and will end on December 1, 2011, is the 1998 edition with the 2000 addenda of Section XI of the ASME Code.
3.0 EVALUATION 3.1 Component for which Relief is Requested The affected component is a Class 2, 3-inch, manually operated, flex-wedge, gate valve manufactured by Walworth (valve number 1NV-240). The valve body is made of SA351, CF8M cast stainless steel material. The valve is a part of the chemical volume and control (NV) system which has a design pressure and temperature of 2,735 pounds per square inch gage (psig) and 189 degrees F, respectively. The valve is situated within the common charging header down stream from the reciprocating charging pump. Both the valve and pump are located outside of containment in the auxiliary building.
3.2 ASME Code Requirements from which Relief is Requested The ASME Code,Section XI, requirements are specified in subparagraph IWC-3122.2, "Acceptance by Repair/Replacement Activity".
3.3 Proposed Alternative and Basis On the basis of the ASME Code,Section XI, subparagraph IWC-3122.3, "Acceptance by Analytical Evaluation", the licensee proposes to temporarily accept the as-found through-wall condition to allow continued service through the current unit run cycle instead of repairing the valve as required in the ASME Code,Section XI, subparagraph IWC-3122.2. This proposed alternative is based on the licensee performing the following actions.
- 1. Operations staff [of the McGuire Nuclear Station Unit 1 ] addressed the potential extent of condition by conducting a thorough inspection of NV piping from pump suction to the containment penetrations for the normal charging header, the reactor coolant (NC) system seal supply headers, and the NV cold
leg injection headers of both trains in both units. All accessible components, including cast valves were inspected. No through-wall leakage, other than valve 1NV-240, was identified during these inspections.
- 2. Operations personnel [of the McGuire Nuclear Station Unit 1] shall observe and measure the valve flaw leakage rate and record a value in a retrievable format once every shift to ensure early detection of an increased leak rate and to ensure the assumptions used in the component operability evaluation remain valid.
- 3. NDE [nondestructive examination] personnel [of the licensee ] shall conduct a "best effort" ultrasonic volumetric examination of the flaw location every 90 days until the valve is repaired.
- 4. An ASME Code repair shall be performed during the next scheduled refueling outage, 1EOC19 RFO (Unit 1 end of cycle 19 refueling outage), which is currently scheduled to begin in September 2008. If a condition leads to a forced outage of sufficient duration before 1EOC19 RFO, the repair will be performed during this forced outage.
The licensee performed fracture mechanics analyses as part of "Operability Evaluation - Valve 1NV-240" which contains the technical basis for (a) considering the valve operable but degraded/non-conforming to ASME Code,Section XI requirements and (b) the requested relief from Code requirements.
The licensee stated that performing an immediate Code repair/replacement activity to correct flaws that have such a minor leak rate (< 0.01 gpm) would create a hardship based on the following overriding concern: the potential risks associated with unit cycling and emergent equipment issues incurred during shutdown and startup evolutions. No compensating increase in the level of quality and safety would be gained by immediate repair of the flaws. Engineering calculations and judgment provide the basis to state that the NV system valve body is very robust and capable of performing its design function through the end of the current fuel cycle.
3.4 Duration of the Proposed Relief The licensee stated that Relief Request 07-MN-001 will be used until Code repair/replacement activities are performed on the valve body either during the refueling outage at the end of cycle 19 (1EOCI9 RFO) or during a forced outage of sufficient duration before 1EOC19 RFO.
3.5 Nuclear Regulatory Commission (NRC) Staff=s Evaluation As stated above, the ASME Code,Section XI, subparagraph IWC-3122.2 requires immediate repair of degraded valve 1NV-240. However, the licensee proposed to delay the repair to either during the next refueling outage, 1EOC19 RFO, or during a forced outage of sufficient duration.
The proposed alternative is based on an operability evaluation which includes the flaw characterization, root cause analysis, and flaw evaluation. The proposed alternative also includes the licensees commitment to perform leakage monitoring and periodic inspections of valve 1NV-240. The NRC staffs evaluation of the proposed alternative is discussed below.
Flaw Characterization and Root Cause Analysis Valve 1NV-240 has a design pressure and temperature of 2890 psig and 189 degrees F.
The system design pressure and temperature are 2735 psig and 189 degrees F. The minimum wall thickness is 0.638 inches in the body neck region. This valve provides isolation for outage maintenance, but does not provide any active safety-related functions. Valve 1NV-240 is located on the common charging header downstream of the charging flow control valve (1NV-238). It remains normally open to provide charging flow and enable operation of reactor coolant pump (NCP) seal injection backpressure control valve (1NV-241). Because 1NV-240 is outside containment, not part of the containment isolation system, and is not a Class 1 pressure boundary, this valve is not part of the Technical Specification defined reactor coolant pressure boundary. Thus, the requirements of Technical Specification 3.4.13 regarding reactor coolant pressure boundary leakage are not applicable.
Valve 1NV-240 normally operates with a nominal internal pressure of 2500 psi. Operating temperature is normally below 120 degrees F as it is upstream of the regenerative heat exchanger. The valve has likely been in service for 30 years or longer. The leakage is from two small pinhole defects, located on the valve body neck. The pinholes are located in close proximity in an area where the valve was previously repaired by welding. The pinholes appear to be located at small depressions. The flaw may be associated with a small thumbnail-shaped defect interconnecting the two primary leak sites. Subsequent examinations determined that the leak is located near the center of a 2.5-inch diameter circular-shaped weld repair. Per the Certified Mill Test Report (CMTR) for 1NV-240, the weld filler material used for the previous repair was E316-16. This weld repair area was one of numerous documented weld repair areas previously performed on the valve body.
The flaw in the valve appears to be curved and discontinuous where it is breaking the outside diameter surface of the valve body. The licensee stated that this flaw is not consistent with a mechanically initiated crack. Furthermore, the process application is not prone to cyclic pressure pulsations/cycles, and the body neck stresses are very low relative to code allowables.
The valve stress analyses documented that under 4g seismic acceleration, the valve stresses were about 20 percent of the code allowable stresses. This acceleration force readily envelopes the safe shutdown earthquake. This valve is located on the charging header, which is not subject to high vibration. Thus, the licensee concluded that the likelihood of fatigue cracking is remote.
Although the presence of a weld repair would increase the localized residual stresses and potentially increase sensitization in the weld heat affected zone (HAZ), the licensee thinks that stress corrosion cracking (SCC) also seems improbable due to the low susceptibility of this material in a deoxygenated, low-temperature borated water process. Additionally, the materials are welded and cast austenitic stainless steel, both of which contain a small percentage of delta-ferrite in their microstructure; this structure is inherently more resistant to stress corrosion cracking than wrought materials.
The licensee stated that based on the shape and surface morphology of the defect, as well as being located near the center of a weld-repaired area, the most likely cause of the leak is a weld flaw. The weld flaw may also have been influenced by the presence of a pre-existing casting flaw. Possible weld and/or casting flaws include shrinkage cracks, hot tearing, porosity, and/or
entrapped slag/inclusions - alone or in combination. Based on review of the CMTR, the weld procedure and filler material used for the previous valve repair were appropriate.
Due to the geometry of the valve body, complexity of the cast microstructure, roughness of the casting surface and other factors, detection of such flaws can be difficult, if not impossible.
Although the previous repair may have been leak tight, through repeated pressure and temperature cycles over time, and slow removal of any entrapped slag or oxidation products within the defect, a tortuous leak path was eventually created. This type of flaw is consistent with the very low observed leak rate, and would also be unlikely to develop into a rapid increase in leak rate.
The NRC staff provided a generic evaluation of the degradation of cast stainless steel in a letter to Douglas J. Walters of Nuclear Energy Institute,
Subject:
License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components, May 19, 2000 (ADAMS Accession Number ML003717179). The study was related to cast austentic stainless steel (CASS) degradation due to thermal aging embrittlement resulting from high temperature conditions (from about 572 to 752 degrees F). However, in the attachment to the May 19, 2000, letter, the NRC specifically evaluated the integrity of less than 4-inch diameter valve bodies made from CASS. The NRCs evaluation concluded that CASS valve bodies can withstand very large through-wall cracks. The industry data show that most degradation in small diameter valves is caused by erosion and general corrosion, not SCC. NRC Report, NUREG/CR-6923, Expert Panel Report on Proactive Materials Degradation Assessment, February 2007, also provides data on thermal aging embrittlement on CASS. The thermal aging embrittlement does not apply to Valve 1NV-240 which is operated at a low temperature of 189 degrees F. Based on the industry data in the two above reports, the likelihood of SCC in valve 1NV-240 is remote.
In addition, the flaw in valve 1NV-240 would not be caused by SCC because flaw growth rate of SCC is aggressive whereas the flaw growth rate of 1NV-240 seems to be low as shown in the low leak rate. The NRC staff agrees with the licensee that a rapid increase in flaw size or leak rate is unlikely during one operating cycle because the valve is fabricated with cast stainless steel material which is resistant to brittle fracture under the plant operating conditions.
Non-Destructive Examination The licensee used the following NDE examination methods to locate and characterize the flaw.
An eddy current examination revealed nine areas on the valve body where the base metal had been repaired by welding. A computed radiography examination with the source placed in two different positions revealed no large voids. Also, "best effort" ultrasonic scans using four (4) beam angles from four directions (two axial and two circumferential) were made of each valve body weld repair area without revealing any planar flaws.
As stated above, the licensee has committed to conduct a "best effort" ultrasonic volumetric examination of the flaw location every 90 days until the valve is repaired. The NRC staff notes that ultrasonic testing (UT) has not yet been qualified to examine cast austenitic stainless steel.
Therefore, the flaw size and associated growth in the valve could not be detected with any certainty or accuracy. The NRC staff questioned the adequacy of the proposed best effort ultrasonic examination and whether other volumetric examinations should be considered.
With regard to alternative volumetric examinations, the licensee stated that radiography does not provide a suitable alternative for this volumetric examination. The sensitivity for radiography of
this material thickness would be expected to be about 2-2T (detectable flaw image limited to approximately 2% of the valve wall thickness). However, given the geometry of the valve (double wall radiography with the valve internals in place), radiography is not adequate to detect a planar flaw prior to it propagating to the acceptance criteria.
As for the UT used in examining the valve, the licensee calibrated the transducer for 45-, 60-and 70-degree beam angles. As no cast stainless calibration block was available, an SA-240 stainless steel plate with a 2-millimeter diameter side drilled hole at a depth of 1.0 inches was used to calibrate for the 45-, 60-, and 70-degree beam angles. The creeping wave probe was calibrated using a 1/16-inch diameter side drilled hole at a depth of 0.2 inches. Reference sensitivity was established using the applicable hole signal set at 80% full screen height.
Scanning was performed at +12dB over reference sensitivity for the 45-degree probe and +6dB over reference sensitivity for 60- and 70-degree probes. The scanning gain for the creeping wave probe was set to reference sensitivity due to excessive front surface noise at higher gain levels. The UT examiners were qualified under the 1998 edition through the 2000 addenda of the ASME Code, Section Xl, Appendix VIII, Supplement 10, for dissimilar metal weld examinations.
The licensee concluded that there is reasonable assurance that the UT techniques used are capable of detecting planar flaws once they grow into the outer two-thirds wall of the cast stainless material. During UT of valve 1NV-240, the 45-degree beam enabled monitoring of the inside diameter surface while all ultrasonic displays were relatively noise free. The 4 MHz straight beam search unit showed similar low noise levels. These characteristics allowed the examiner to conclude that a relatively small uniform grain size is present. Since detection of a flaw in cast stainless steel material is dependent on variations in grain structure, a uniform grain size enhances the ability to detect flaws. With the 45-degree scan it was possible to monitor the inside surface noise level at 10% full screen height assuring that some sound energy was reaching the inner surface. There was little internal noise, indicating grain sizes approximately one wave length or larger were not present.
The licensee stated further that the UT used was consistent with the manual UT used and approved in Catawba Relief Request 04-CN-001, Docket Nos. 50-413 and 50-414. The UT referenced in the relief request used large 1 MHz search units applicable to 34-inch and 36-inch diameter pipe with a thickness of approximately 2.5 inches. The focal distances chosen for the large diameter pipe were based on an area of interest extending from 1.6 inches in the pipe to within 0.1 inch of the outside diameter. The same UT techniques applied to valve 1NV-240 used smaller 2 MHz transducers to fit the smaller diameter of the valve and employed focal distances of 1.0 inch to within 0.1 inch of the valve outside diameter.
The licensee clarified that UT is capable of detecting planar flaws once they grow beyond the inner 1/3 material wall thickness and have a measured length of 1.5 inches or greater. The licensees NDE Level III personnel conducted a thorough research of industry experience before performing the examinations. The references cited below provide the technical basis for examining the outer 2/3 thickness of the material and the flaw length measurement criterion:
- 1. EPRI Report TR-107481, "Status of the Ultrasonic Examination of Reactor Coolant Loop Cast Stainless Steel Materials", March 1998.
- 2. Draft White Paper: Current Inspection Capabilities for Cast Austenitic Stainless Steel Piping, Chockie Group International, Inc., November 2005.
- 3. Safety Evaluation for Relief Request 04-CN-001, TAC Nos. MC 2209 and MC 2210 (ADAMS Accession No. ML051640058).
This SE also cites NUREG/CR-6594, "Evaluation of Ultrasonic Inspection Techniques for Coarse-Grained Materials", October 1998.
The licensee stated that detection capability improves as the flaw increases in through-wall extent because the sound beam travels through less of the coarse grained material. This has been validated by studies at the Electric Power Research Institute Nondestructive Examination Center and Pacific Northwest National Laboratory. Length sizing of the flaw is dependent on variations in grain structure within the cast material and the surface condition on the outside diameter. Ultrasonic characteristics observed during UT inspections suggested that this valve body grain structure is relatively fine and uniform. However, the surface is in the as-cast condition making constant pressure on the search unit difficult. Accurate length sizing of flaws using manual ultrasonic angle beam techniques requires a surface finish much finer than the as-cast condition. Experiments conducted on deep cracks in centrifugally cast stainless steel reactor coolant loop piping specimens show that the best performing search units were only capable of measuring 40% of crack lengths at depths of 62%, 72%, and 82% through-wall.
Because of these factors, the licensee will length-size any crack-like flaws conservatively from peak amplitude down to the baseline, and in no case less than 1.5 inches which is twice the search unit's width.
The licensee noted that an ultrasonic examination performed every 90 days provides reasonable assurance of detecting a planar flaw in the outer 2/3 wall thickness of the cast stainless steel valve before the flaw exceeds the fracture mechanics acceptance criteria.
The 90-day inspection interval was selected based on information provided in ASME Code Case N-513. The licensees flaw evaluation showed that a flaw would not grow beyond the acceptance limit in less than 90 days. The flaw evaluation showed that with an assumed large flaw and a large number of cycles, crack growth is essentially negligible over an 18-month fuel cycle.
Additionally, the licensee will conduct twice-daily visual observations of the leaking valve to identify changes prior to their compromising the safe operation of the plant.
The NRC staffs position is that the existing UT technology is insufficient to satisfy the UT performance demonstration requirements of ASME Code,Section XI, Appendix VIII, with respect to the examination of cast stainless steel material; therefore, the NRC staff has reservations regarding the reliability of the licensees best effort UT examination. First, the licensee used a wrought austenitic stainless steel calibration block which has different microstructure than the cast stainless steel, and second, the licensee used a 2 MHz search unit to inspect the valve.
Use of a lower frequency in the range of 0.5 MHz may provide better results and less scattering as recommended in NRC Report NUREG/CR-6933, Assessment of Crack Detection in Heavy-Walled Case Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods. However, the licensee has increased the frequency of the best effort examination to
every 90 days. Frequent UT examinations provide opportunities for the licensee to compare the results from one examination to another to verify the conditions of the valve, even if the UT is not qualified. The licensees best effort examination will be consistent with Appendix III of ASME Code,Section XI, which is required by the ASME Code for the examination of cast stainless steel. In addition, the licensee will conduct two daily visual examinations of valve 1NV-240. The NRC staff finds that the licensees NDE strategy is acceptable because the frequency and redundancy of the proposed NDE adequately compensates for the prescriptive Appendix III UT.
Leakage Monitoring The licensee stated that the current licensing basis allows 0.35 gpm leakage into the auxiliary building. The current Unit 1 define Emergency Core Cooling System (ECCS) system total leakage, inclusive of 1NV-240 through wall leakage is much less than 0.05 gpm. Thus, the allowed loss-of-coolant accident (LOCA) dose analyses limits are not currently challenged.
With regard to leakage monitoring, the licensee will check the valve leakage rate once per shift per the Semi-Daily Surveillance procedure (PT/1/A/4600/003A). The leakage rate is quantified by visual observation and recorded in drops per minute. Leakage has remained less than 1 drop per minute for several months. The instantaneous leakage rate can be remotely determined by the control room reactor operators from a live video feed. Other routine programmatic monitoring activities would allow detection of any significant change in 1NV-240 leakage rate as follows: (1) auxiliary building rounds in the general area, (2) periodic inspections of 1NV-240 under the fluid leak management program, (3) reactor operators routinely monitor volume control tank level trends, and (4) reactor coolant system leakage calculations.
The licensee also notes that 1NV-240 and associated piping are not insulated and are readily accessible for visual examination under the fluid leak management program and by operations personnel during normal rounds. The NRC staff finds that without insulation on valve 1NV-240, the visual examination should provide more accurate and readily leak rate reading than if the valve is insulated.
Based on the above evaluation, the NRC staff finds that the licensee has implemented adequate procedures to monitor the valve leak rate because of the five programmatic monitoring activities and specific acceptance leak rate criterion discussed above.
Flaw Evaluation The licensee performed two flaw evaluations to determine the allowable and critical crack size.
The first flaw evaluation was performed after the flaw was discovered considering the entire valve body to be inspectable. Hence, the total thickness of the valve was used in determining the allowable flaw size. Later, the licensee performed a second flaw evaluation to address the compound flaw cases to match the final conclusions regarding the UT capability. Other changes regarding cycles assumed, pressure ranges, operating vs. design temperature, bending stresses, and initial flaw sizes for crack growth were also incorporated in the second flaw evaluation to be more conservative or more realistic, and are considered improvements to the first flaw evaluation.
There is not an ASME Code Case to evaluate flaws in a Class 2 valve body. Therefore, the licensee used the ASME Code,Section XI, Appendix C and Code Case N-513-1 (Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section Xl, Division 1) as a guide to perform the flaw evaluations. The NRC staff has endorsed Code Case N-513-1 in Regulatory Guide 1.147, Revision 14. The licensee modeled the valve body neck as a pipe and using a flaw depth to wall thickness ratio of unity that is similar to the approach of ASME Code Case N-513-1. Allowable flaw sizes were determined using limit load criteria specified in Article C-5000 of the ASME Code,Section XI, in both the axial and circumferential directions. Flaw growth evaluation considering fatigue as a possible mechanism is performed using the methodology in ASME Code, Section Xl, Appendix C for stainless steel components.
The licensee used ASME safety factors against failure for Service Levels A/B, and included a safety factor of 2.77 and 3.0 for the circumferential and axial, respectively. Service Levels A/B are designated for the normal operating/upset conditions. The safety factors for Service Levels A/B are higher than for Levels C/D, emergency/faulted conditions.
The results from the first flaw evaluation show that the allowable flaw sizes are either 3.9 inches long 100% through-wall axial flaw or 5.4 inches long 100% through-wall circumferential flaw.
In the second flaw evaluation, the limiting flaw sizes were determined using not merely a single axial or circumferential 100% through-wall crack, but a compound crack. That is, for various assumed depths of 360-degree (full circumferential) inside diameter initiated cracks (corresponding to the deepest undetectable full circumferential inside diameter cracks),
allowable lengths of a 100% through-wall circumferential and axial cracks were determined. The licensee also evaluated crack growth to determine the largest existing axial flaw that would not grow to exceed the allowable size within 100 cycles of 0 to 2735 psig pressure loading.
The fatigue crack growth rate for austenitic steels exposed to water environments was used.
The licensee calculated various allowable and critical flaw sizes in terms of assumed UT capability in detecting flaws through 50% to 100% valve wall thickness. For example, for the case of UT penetrating 100% of the valve wall thickness, the allowable axial and circumferential flaw sizes are 3.95 inches and [greater than] 6 inches, respectively. The critical axial and circumferential flaw sizes are 13.5 inches and [greater than] 9 inches, respectively. For the case of UT penetrating 66% of the valve wall thickness, the licensee calculated an allowable axial flaw size of 1.64 inches and an allowable circumferential flaw of greater than 6 inches. The critical axial flaw size for this case is 7.21 inches and the circumferential flaw size can be greater than 9 inches.
The current flaw size at the valve surface is only a pinhole dimension, and the largest undetectable size (possibly just below the surface) is 1.5 inches. This is well below the 6-inch allowable size or 9-inch critical size in the circumferential direction. Because the largest flaws should not grow more than a negligible amount in the postulated service conditions, the current conditions (assuming circumferential-radial flaw orientation) are stable, and will remain stable, with wide margins of error.
The NRC staffs position is that the existing UT technology is insufficient to satisfy the UT performance demonstration requirements of ASME Code,Section XI, Appendix VIII with respect to the examination of cast stainless steel material. Notwithstanding, the NRC staff believes that the licensees two fracture mechanics analyses provide useful information regarding
approximate allowable and critical flaw sizes. The NRC staff does not believe the existing flaw would grow to exceed the allowable flaw size during the current operating cycle 19.
In addition, the licensee will perform visual examinations of the valve twice daily. The licensee will perform best effort UT every 90 days. The licensee has installed a video camera to monitor the valve. These procedures in concert with the flaw evaluations will provide reasonable assurance that the existing flaw will not grow to unacceptable size because any significant leakage, which would be caused by significant flaw growth, will be detected early. The application of relief request 07-MN-001 is limited to one operating cycle, cycle 19. The NRC staff does not believe that in one cycle of operation, the degraded valve will degrade to the point of affecting public safety significantly because the licensee has increased monitoring of the subject valve. Therefore, the NRC staff finds that relief request 07-MN-001 provides reasonable assurance of structural integrity.
Criteria for Immediate Repair As stated above, the licensee proposed to delay the Code repair of valve 1NV-240 to either during the next refueling outage or a forced outage of sufficient duration in the current operating cycle. However, the NRC staff questioned under what circumstances must the valve be repaired immediately. In the December 3, 2007, submittal, the licensee responded that there are two separate monitoring criteria associated with valve 1NV-240, which if exceeded would result in unit shutdown and immediate valve replacement. The monitoring criteria consist of an allowed operational leak rate and an allowed flaw size.
Valve 1NV-240 is part of the chemical volume and control system, which is part of the emergency core cooling system (ECCS) piping boundary. The radiological dose analysis limits external leakage from the ECCS and containment spray systems to 0.25 gallons per minute (gpm). Technical Specification 5.5.3 further requires that leakage from these systems be minimized. A significantly lower operational leakage threshold for 1NV-240 is specified in the semi-daily surveillance procedure (PT/1/A/4600/003A). If 1NV-240 leakage approaches a steady stream (<<0.25 gpm), procedural actions are in-place to attempt to isolate 1NV-240 (PT/1/A/4600/003A, Enclosure 13.3, and OP/1/A/6200/01B, Enclosure 4.12). If a catastrophic valve structural failure were to occur, abnormal station procedures provide mitigative steps to isolate 1NV-240 and achieve safe unit shutdown (AP/1/A/5500/010, NC System Leakage Within Capacity of Both NV Pumps).
The licensee will perform periodic UTs every 90 days to ensure the flaw size does not exceed the fracture mechanics analyses acceptance criteria. The Operability Evaluation shows that there is reasonable assurance that the ultrasonic examination techniques used are capable of detecting planar flaws once they grow beyond the inner 1/3 material wall thickness and have a measured length of 1.5 inches or greater. The flaw analysis in the Operability Evaluation states that for an axial crack (most limiting), up to an inside diameter wall thinning of 33%
(corresponding to being able to detect flaws in only the outer 2/3 of the wall), the allowable length of a 100% through-wall flaw is 1.64 inches. Forced valve replacement will occur if the periodic UT characterizes a flaw size greater than 1.64 inches in the axial direction or greater than 6 inches in the circumferential direction. The licensee stated that using an inside diameter of 4.313 inches and wall thickness of 0.8125 inch, the inner circumference of the valve would be 13.55 inches and the outer circumference would be 18.6 inches.
Additionally, the licensee has added 1NV-240 to the Unit 1 Forced Outage List. Modification ME1 01307 under Work Order 1757969 has been approved to replace 1NV-240 during either a Unit 1 forced outage or the next refueling outage (1 EOC19).
Based on the above evaluation, the NRC staff finds that the licensee has implemented sufficient procedures and acceptance criteria to perform repairs should valve 1NV-240 develop unacceptable leakage.
4.0 CONCLUSION
The NRC staff has reviewed the licensee's submittal and, based on the information provided, the NRC staff concludes that the licensee has demonstrated reasonable assurance of the structural integrity of valve 1NV-240 and compliance with the specified ASME Code requirements will result in hardship without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the use of Relief Request 07-MN-001 to delay the Code repair of charging pump discharge line valve 1NV-240 to either during the end of cycle 19 refueling outage or a forced outage of sufficient duration during current operating cycle 19 at McGuire 1.
All other requirements of ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: J. Tsao Date: March 26, 2008
McGuire Nuclear Station, Units 1 & 2 cc:
Vice President McGuire Nuclear Station Duke Power Company, LLC 12700 Hagers Ferry Road Huntersville, NC 28078 Associate General Counsel and Managing Attorney Duke Energy Carolinas, LLC 526 South Church Street - EC07H Charlotte, North Carolina 28202 County Manager of Mecklenburg County 720 E. Fourth St.
Charlotte, NC 28202 Regulatory Compliance Manager Duke Energy Corporation McGuire Nuclear Site 12700 Hagers Ferry Road Huntersville, NC 28078 Senior Resident Inspector U.S. Nuclear Regulatory Commission 12700 Hagers Ferry Road Huntersville, NC 28078 Mecklenburg County Department of Environmental Protection 700 N. Tryon St Charlotte, NC 28202 Vice President Customer Relations and Sales Westinghouse Electric Company 6000 Fairview Road, 12th Floor Charlotte, NC 28210 NCEM REP Program Manager 4713 Mail Service Center Raleigh, NC 27699-4713 Assistant Attorney General NC Department of Justice P.O. Box 629 Raleigh, NC 27602 Manager Nuclear Regulatory Issues &
Industry Affairs Duke Energy Corporation 526 S. Church St.
Mail Stop EC05P Charlotte, NC 28202 Division of Radiation Protection NC Dept of Environment, Health & Natural Resources 3825 Barrett Dr.
Raleigh, NC 27609-7721 Owners Group (NCEMC)
Duke Energy Corporation 4800 Concord Road York, SC 29745 Group Vice President, Nuclear Generation
& Chief Nuclear Officer P.O. Box 1006-EC07H Charlotte, NC 28201-1006 Senior Counsel Duke Energy Carolinas, LLC 526 South Church Street - EC07H Charlotte, NC 28202