ML14034A176
| ML14034A176 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/24/2014 |
| From: | Jessie Quichocho Plant Licensing Branch II |
| To: | Nazar M Florida Power & Light Co |
| Klett A DORL/LPL2-1 301-415-0489 | |
| References | |
| TAC MF2996 | |
| Download: ML14034A176 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power & Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 February 24, 2014
SUBJECT:
TURKEY POINT NUCLEAR GENERATING UNIT NO. 3-SAFETY EVALUATION FOR RELIEF REQUEST NO. 13 FOR FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL-ALTERNATIVE REPAIR OF CONTAINMENT SPRAY SYSTEM VALVE (TAC NO. MF2996)
Dear Mr. Nazar:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 23, 2013, as supplemented by letter dated January 9, 2014, Florida Power & Light Company (the licensee) submitted Relief Request No. 13 for the Turkey Point Nuclear Generating Unit No. 3. Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR), Paragraph 50.55a(a)(3)(ii), the licensee proposed alternatives to the requirements of 10 CFR 50.55a(g)(4) on the basis that complying with the specific requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Specifically, the licensee proposed alternatives to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, subparagraph IWC-3122.2 requirements for repairing or replacing components with flaws that exceed the acceptance standards of Table IWC-341 0-1.
The relief request applies to the containment spray pump manual suction isolation valve, Valve 3-844A.
The NRC staff reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee adequately addressed all regulatory requirements in 10 CFR 50.55a(a)(3)(ii). Accordingly, the NRC staff authorizes Relief Request No. 13 (i.e., to allow the valve to remain in service with increased monitoring and to defer repair/replacement) until the next scheduled refueling outage expected to begin in March 2014, the next forced outage of sufficient duration requiring entry into Mode 5, or the predicted flaw size exceeds acceptance criteria, whichever occurs first.
All other ASME Code,Section XI requirements for which relief was not specifically requested and authorized in Relief Request No. 13 remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
If you have any questions regarding this issue, please contact the project manager, Ms. Audrey Klett, at (301) 415-0489 or by e-mail at Audrey.Kiett@nrc.gov.
Docket No. 50-250
Enclosure:
Safety Evaluation cc w/encl.: Distribution via Listserv Sincerely, Jessie F. Quichocho, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 13 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL FLORIDA POWER & LIGHT COMPANY TURKEY POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-250
1.0 INTRODUCTION
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 23, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13318A010), as supplemented by letter dated January 9, 2014 (ADAMS Accession No. ML14030A183), Florida Power & Light Company (the licensee) submitted Relief Request No. 13 for the Turkey Point Nuclear Generating Unit No. 3. Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 50.55a(a)(3)(ii), the licensee proposed alternatives to the requirements of 10 CFR 50.55a(g)(4) on the basis that complying with the specific requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Specifically, the licensee proposed alternatives to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, subparagraph IWC-3122.2 requirements for repairing or replacing the containment spray pump manual suction isolation valve, Valve 3-844A, which has a flaw that exceeds the acceptance standards of Table IWC-341 0-1. The relief request pertains to the fourth 1 0-year inservice inspection (lSI) interval. The licensee may invoke the provision of the ASME Code,Section XI, paragraph IWA-2430, which allows the licensee to extend the fourth 1 0-year lSI interval by 1 year.
By electronic mail dated December 4, 2013 (ADAMS Accession No. ML13338A696), the NRC sent the licensee a request for additional information regarding the relief request. By letter dated January 9, 2014, the licensee responded to this request.
2.0 REGULATORY EVALUATION
The licensee proposed alternatives to 10 CFR 50.55a(g)(4)- specifically to the ASME Code,Section XI, subparagraph IWC-3122.2-pursuant to 10 CFR 50.55a(a)(3)(ii).
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the Enclosure limitations of design, geometry, and materials of construction of the components. Pursuant to 10 CFR 50.55a(g)(4)(i) and 10 CFR 50.55a(g)(4)(ii), inservice examination of components and system pressure tests conducted during the first 1 0-year inspection interval and subsequent 1 0-year inspection intervals must comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein.
Pursuant to 10 CFR 50.55(a)(3)(ii), alternatives to the requirements of 10 CFR 50.55a(g) may be used when authorized by the Director of the NRC Office of Nuclear Reactor Regulation if 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 its regulatory and technical evaluations in this safety evaluation, the NRC staff finds that the regulatory authority exists to authorize the licensee's proposed alternative on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(a)(3)(ii).
The code of record for the fourth 1 0-year lSI interval is the 1998 Edition with Addenda through 2000 of the ASME Code,Section XI, subject to the limitations and modifications in 10 CFR 50.55a(b).
3.0 TECHNICAL EVALUATION
3.1 Licensee's Proposed Alternative The component associated with the relief request is Valve 3-844A, which is the suction isolation valve for the "A" containment spray pump from the common containment spray pump suction supply header. The valve passively maintains the containment spray system suction piping integrity and provides maintenance isolation for the "A" containment spray pump. The valve is normally maintained in the back-seated, locked-open position during Modes 1 through 4. The valve and pump are located outside containment in the containment spray pump room of the auxiliary building. The valve is a Quality Group 8, ASME Class 2, 8-inch manually operated gate valve manufactured by Anchor-Darling with Series 150 welding ends. The valve is constructed of grade CF-8 cast stainless steel material in accordance with the American Society for Testing and Materials (ASTM or ASTM International) standard, ASTM A351, "Standard Specification for Castings, Austenitic, for Pressure-Containing Parts."
The code of record for the fourth 10-Year lSI interval is the ASME Code,Section XI, 1998 Edition through the 2000 Addenda, subject to the conditions in 10 CFR 50.55a(b). The ASME Code,Section XI, subparagraph IWC-3122.2, "Acceptance by Repair/Replacement Activity," states that a component with flaws that exceed the acceptance standards of Table IWC-3410-1, is unacceptable for continued service until the additional examination requirements of subsubarticle IWC-2430 are satisfied and the component is corrected by a repair or replacement activity to the extent necessary to meet the acceptance standards of article IWC-3000.
In September 2013, the licensee identified a through-wall flaw with boric acid residues near the top of the packing gland on the bonnet of Valve 3-844A that exceeds the acceptance criteria of the ASME Code, Table IWC-341 0-1. The licensee noted that although subparagraph IWC-3122.3 of the ASME Code allows for acceptance by analytical evaluation as described by subarticle IWC-3600 of the ASME Code, it does not provide acceptance criteria for austenitic components. The licensee noted that subarticle IWB-3640, "Evaluation Procedures and Acceptance Criteria for Austenitic Piping," states that the evaluation procedures and acceptance criteria shall be the responsibility of the Owner and shall be subject to approval of the regulatory authority.
NRC Inspection Manual Part 9900 Technical Guidance, "Operability Determinations &
Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality of Safety" (ADAMS Accession No. ML073531346), Appendix C, "Specific Operability Issues," Item C.11, "Flaw Evaluation," specifies that if ASME Class 2 or Class 3 components do not meet ASME Code acceptance standards, NRC-accepted ASME Code Case requirements, or an NRC-approved alternative, then the licensee must determine whether the degraded or nonconforming condition results in a Technical Specification (TS) system, structure, or component being inoperable. Item C.11 also states that whenever a flaw does not satisfy ASME Code or construction code acceptance standards or the requirements of an NRC-accepted ASME code case, a relief request needs to be submitted in a timely manner after completing the operability determination process documentation.
In its letter dated January 9, 2014, the licensee stated that Valve 3-844A could not be isolated from the upstream common supply header of the containment spray system without taking both trains of the containment spray system out of service. With both trains inoperable, TS Limiting Condition Operation 3.6.2.1, Action b would require restoring at least one train to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or being in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Based on its prompt operability determination, the licensee concluded that the valve continues to be capable of performing its required safety functions and is not susceptible to sudden or catastrophic failure. The licensee stated that performing an ASME Code repair/replacement activity immediately to correct the flaw would create a hardship with no compensating increase in the level of quality and safety gained by immediate repair of the flaw because of the potential risks associated with unit shutdown, thermal stress cycling of plant components, and emergent equipment issues that could incur during shutdown and startup evolutions.
As an alternative to performing the flaw correction by a repair or replacement activity described in subparagraph IWC-3122.2 of the ASME Code, the licensee proposed to temporarily accept the through-wall flaw in Valve 3-844A to allow continued service operation until the next scheduled refueling outage or outage of sufficient duration requiring entry into Mode 5. The licensee noted that although the provisions of Code Case N-513-3, "Evaluation Criteria for Temporary Acceptance of Flaws in Class 2 & 3 Piping,Section XI, Division 1," do not apply to valves, it has followed the analytical evaluation and rules for temporary acceptance of flaws in piping of the Code Case for Valve 3-844A.
The licensee predicted negligible growth of the flaw for the remainder of the present operating cycle (i.e., Cycle 27). The licensee stated that it will monitor the flaw by daily walkdowns and monthly penetrant testing. As discussed in its letter dated January 9, 2014, the licensee projected the final flaw size at the time of repair to be 11/16-inch length by 1/16-inch width, which is the current size of the flaw. The licensee further stated that if the monthly measurement increases from the present by 1/16-inch in either direction (allowing 1/16-inch for measurement uncertainty), then it will re-examine the growth rate to verify the structural analysis conclusions and predicted growth rate.
The licensee requested to apply the proposed alternative until it performs ASME Code repair/replacement activities on the valve body during the next scheduled refueling outage (or forced outage of sufficient duration requiring entry into Mode 5) or when the predicted flaw size exceeds acceptance criteria, whichever occurs earlier. The licensee stated that its next refueling outage begins in March 2014.
The licensee stated that this relief request is similar to the relief granted to the McGuire Nuclear Station, Unit 1, on March 26, 2008 (ADAMS Accession No. ML080580577).
3.2 NRC Staff's Evaluation The ASME Code,Section XI, IWC-3122.2 requires the licensee repair or replace the degraded valve (i.e., Valve 3-844A) when acceptance criteria are not met. The licensee requested an alternative to the required ASME Code repair/replacement of the subject valve that uses a flaw evaluation and monitoring of the flaw to demonstrate the structural integrity of the degraded valve until the licensee can repair/replace the valve in accordance with the ASME Code,Section XI during the next outage. The NRC staff evaluated the acceptability of the licensee's flaw evaluation and proposed monitoring of the flaw.
The licensee followed Code Case N-513-3 and the ASME Code,Section XI, Appendix H, "Evaluation of Flaws in Ferritic Piping," for its flaw evaluation. The licensee included the applied loads for pressure, deadweight, thermal, and seismic conditions. The licensee noted that the applied loads are relatively minor because of the low pressure and temperature conditions, the valve being adjacent to the pump suction nozzle anchor point, and the low seismic accelerations. The licensee calculated an allowable flaw size of 5.63 inches and 23.4 inches in the circumferential and axial directions, respectively. The licensee also evaluated the flaw growth based on potential effects of environmentally assisted cracking, limited number of operating cycles, and low resultant stresses. The licensee concluded that flaw growth is not expected for the valve during the remainder of the current operating cycle.
The NRC staff notes that although the licensee has followed the flaw evaluation method in Code Case N-513-3 and Appendix H to the ASME Code,Section XI, both methods apply to pipes rather than valves and especially not to the valve bonnet where the flaw is located. The licensee used an idealized model assuming the flaw is located in a pipe instead of in the valve bonnet. The NRC staff finds that the calculated allowable flaw sizes are an approximation and a reference. However, the NRC staff notes that the measured flaw size in the field is significantly less than the calculated allowable flaw sizes. The NRC staff further notes that the applied loads at the flaw are not significant because the flaw is located at the valve bonnet, which usually does not experience significant loading. The NRC staff finds that the through-wall flaw should be stable and that the valve will not fail catastrophically under design loading or accident conditions.
The NRC staff finds that even if the flaw propagates, the licensee-proposed daily walkdowns and monthly penetrant testing will ensure that the flaw will not exceed the following acceptance criterion. The licensee proposed to limit the flaw growth to a 1/16-inch growth rate per month (i.e., a 1/16-inch change in any monthly measurement). In its letter dated January 9, 2014, the licensee stated that if the flaw grows 1 /16-inch in length in any month, it will reanalyze the flaw to maintain evaluation validity and will increase the inspection frequency of the walkdowns from daily to once per shift. There are three shifts per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The licensee also proposed that if the leak rate becomes greater than 25 drops per hour, then it will initiate corrective actions to repair/replace the valve. The licensee derived an acceptance criterion of 25 drops per hour based on the Turkey Point Updated Final Safety Analysis Report (UFSAR), Section 6.2, "Safety Injection System." Section 6.2 of the UFSAR states that the majority of valves, except those valves that perform a control function, is provided with backseats that are capable of limiting leakage to less than 1.0 cubic centimeter per hour (cc/hour) per inch of stem diameter, assuming no credit is taken for valve packing. The licensee assumed that the subject valve's backseat is capable of limiting the leakage to less than 1.0 cc/hour. The licensee determined that the leak rate that is considered significant for this valve is 25 drops per hour, which is based on the valve stem diameter being 1.25 inches (i.e.,
1.25 inches multiplied by (x) 1cc/hour x 0.0002642 gallons/cc x 1 hour/60 minutes= S.SE-6 gallons/minute = 25 drops/hour).
The NRC staff finds that 25 drops per hour is a stringent and conservative leak rate criterion to initiate the repair/replacement of the valve and is therefore acceptable. The NRC staff further finds that walkdowns performed once per shift (i.e., once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) is acceptable if the flaw length increases by 1/16-inch per month. The NRC staff determines that the administrative limits on the flaw growth rate and leak rate with associated corrective actions will provide reasonable assurance of the structural integrity of the valve.
The licensee stated that Valve 3-844A is located in the containment spray pump room, which is off the north-south hallway of the Unit 3 Auxiliary Building. This area is normally accessible for plant personnel to perform visual inspections during the daily walkdowns, measure the leakage, and perform monthly flaw size measurements. The licensee further stated that Valve 3-844A (including the flaw location) is not covered with insulation during normal operation. The NRC staff finds that the licensee will be able to perform adequate monitoring because the valve is fully accessible.
The licensee disassembled Valve 3-844A for inspection in October 2007. At the time, the licensee inspected and measured the packing gland internal area with the valve stem removed.
The licensee noted that the carbon spacer in the packing was in good condition. The licensee did not notice long-term corrosion on the inside of the packing gland area, nor did it notice a large area of degradation (i.e., missing metal). The licensee stated that the through-wall flaw appears to be a casting defect based upon its void-like shape at the outside diameter surface.
The licensee explained that given the corrosion resistance of casting material to borated water, the through-wall leak path likely developed because of gradual washout or removal of high temperature oxides from a casting defect rather than corrosion of sound metal. The NRC staff finds the licensee's conclusion that the flaw was caused by a casting defect inconclusive unless a destructive examination can be performed on the flaw and determines a casting defect is the degradation mechanism. However, the licensee's inspection of the valve in 2007 provides evidence that significant long-term corrosion is not expected for this valve.
The licensee performed an extent of condition visual examination on manual valves at five of the most susceptible and accessible locations. Three susceptible locations are on identical valves in the same service. The remaining two valves examined are 6-inch stainless steel manual valves in similar service (same fluid and service conditions). The licensee did not identify any evidence of unacceptable defects (i.e., no relevant indications, leakage, or dry boric acid residues) at the valves examined. The NRC staff finds that the licensee performed an adequate extent of condition inspection.
The NRC staff finds that requiring a plant shutdown to repair a relatively insignificant flaw in the subject valve in accordance with the ASME Code,Section XI, subparagraph IWC-3122.2 is a hardship. The NRC staff further finds that considering the corresponding adverse impact on the plant, the ASME Code repair or replacement of the valve would not increase in the level of quality and safety. The NRC staff finds that based on flaw evaluation, proposed monitoring, and stringent acceptance criteria, the licensee has demonstrated structural integrity of the valve (i.e.,
the through-wall flaw is stable, and the valve will not fail catastrophically under design loading or accident conditions) for the duration of the relief request.
4.0 REGULATORY COMMITMENTS In its letter dated October 23, 2013, the licensee included a regulatory commitment that if the monthly measurement of the flaw increases from the present by 1/16-inch in either direction (allowing 1/16-inch for measurement uncertainty), then the growth rate will be reexamined to verify the structural analysis conclusions and predicted growth rate. The licensee incorporated this language in its proposed alternative, which the NRC authorizes. Therefore, the licensee's reexamination of the growth rate, should it be required in accordance with this relief request, is an obligation that is not changeable in accordance with the licensee's commitment management program.
5.0 CONCLUSION
As set forth in the aforementioned evaluation, the NRC staff determines that the licensee's proposed alternative provides reasonable assurance of structural integrity of Valve 3-844A. The NRC staff finds that complying with the specified ASME Code requirement to repair Valve 3-844A would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) and is in compliance with the requirements of the ASME Code,Section XI for which relief was not requested. Therefore, the NRC authorizes the use of Relief Request No. 13 dated January 9, 2014, which specifies an acceptable leak rate criterion for initiating valve repair/replacement, until the next scheduled refueling outage that is scheduled to begin in March 2014, forced outage of sufficient duration requiring entry into Mode 5, or when the predicted flaw size exceeds acceptance criteria, whichever occurs first.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: John Tsao Da~: ~ebruary 24, 2014
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