JPN-92-067, Forwards Root Cause Failure Analysis of Equalization Line on Reactor Water Cleanup Valve 12RWC-46, & ASME Section XI Repair Program for Repair of RWCU Valve 12RWC-46 - 1/2 Inch Equalization Line Plant Mod M1-92-383
| ML20126A118 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/14/1992 |
| From: | Ralph Beedle POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20126A121 | List: |
| References | |
| JPN-92-067, JPN-92-67, NUDOCS 9212170242 | |
| Download: ML20126A118 (6) | |
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December 14,1992 JPN 92 06/
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Main Station P1 137 Washington, D.C. 20555
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 fic que ntloLHelicLf rom. ASMILSectiottXI
Dear Sir:
This letter provides the Authority's plan to repair the 1/2" bonnet equalization line on Reactor Water Cleanup valve 12RWC 46. A through-wall defect was detected on the i
equalization line to valve bonnet weld during the perfonnance of a hydrostatic test on the Reactor Water Cleanup System. This repair will be performed in accordance with the requirements of ASME Section XI 1980 Edition through Winter 1981 Addenda with the exception of flaw removal. In accordance with 10 CFR 50.55a(a)(3), the Authority requests relief from ASME Section XI paragraphs IWA-5250(a)(2) and IWA-4130(a)(2).
Paragraph IWA-5250(a)(2) requires that the repair be performed in accordance with IWA-4000. Within Section IWA-4000, " Repair Procedures," lWA 4130(a)(2) requires that the repair program contain esseritlal requirements of the repair including the flaw removal method and method of measurement of the cavity createct by removing the flaw. The Authority plans to repair the defect by encapsulation without removing the flaw. Removal of the flaw would require disassembly of the reactor vessel and unnecessarily delay restart from the current outage. Justification for performing the repair without removing the flaw is discussed in Attachment 1.
A preliminary root cause analysis of the defect determined that the most likely cause of the f ailure is thermal fatigue due to differential expansion between the stainless steel equalizing line and the carbon steel valve body. This preliminary root cause evaluation is a qualitative assessment based upon the available data and is provided in. A more exhaustive root cause analysis is being performed. During the next refueling outage, the equalizing line will be removed and a thorough metallurgical evaluation will be performed. The Authority has reviewed the 19 other valves with equalizing lines and the same material configuration does not exist nn other valves.
Therefore these other valves are not subject to the differential themial expansion experienced by 12RWC-46.
The Authority designed a pipe assembly to encapsulate the component that has l-g2Q
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the through wall defect. The design was completed in accordance with the requirements of ASME Section lli 1986 Edition. The design ensures that the repair is capable of meeting all primary and secondary stresses. The design analysis also considered the results of the preliminary root cause analysis.
A fatigue analysis has been performed on the proposed repair to ensure that all components are within the ASME Section lli design requirements for one fuel cycle, alts i
which time the encapsulated pipe assemblies and the enclosed pipes will be removed.
The permanent repair will address all postulated failure mechanisms, The Authority has contingency plans in place to mitigate postulated leakage which may occur during the repair. Elements of the contingency plans include: operable ECCS systems for inventory make-up; gasket materials, clamps, damage control plugs and extra catch containments staged and readily available; and a preplanning meeting to specifically discuss contingency plans. During the repair there will be no reactor vessel water level or pressure changes which could affect the conditions at the repair location.
Additionally, the Authonty will perionn a VT 2 visualinspection of the repair during i
all plant shutdowns in which the containment is deinerted.
The Minor Modification Package, Nuclear Safety Evaluation and the ASME Section XI Repair Program are included in Attachment 3. Welding, non-destruct!ve evaluation requirements, pressure test requirements and other code requirements are discussed in the documentation provided in Attachment 3. The Authorized Nuclear Inservice inspector (ANil) has reviewed and accepts this repair method.
The Authority requests that NRC review and approval of this reliet request be expedited since approvalis required prior to plant startup, il you have any questions, please contact J. A. Gray, Jr.
Vcy truly
- urs, n.G e Ralph E. B.wdle Attachments:
- 1) Justification for Proposed Altemative Repair Method
- 2) Root Cause Failure Analysis: Structural Integrity Associates, Inc., SIR-92 082, December,1992.
- 3) Modification Infom1ation (Minor Modification Package, Nuclear Safety Evaluation and ASME Section XI Repair Program)
cc: Regional Administrator U.S, Nuclear Regulatory Commission 475 Allendato Road King of Prussia, PA 19406 Office of the Resident Inspector U.S. Nuclear Regulatoly Commission P.O. Box 130 L.ycoming, NY 13093 Mr. Brian C. McCabe Project Directorato 11 Division of Reactor Projects 1/(1 U.S. Nuclear Regulatory Commission Mall Stop 14 B2
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ATTACHMENT 1 to JPN-92-067 JUSI]EICAT1QREOREf10EOSED ALTERNAILVE_BEPAIR METHOR The Code of Federal Regulations in 10 CFR 50.55a(a)(3) allows proposed attematives to the requirements of ASME Section XI. Contrary to ASME Section XI, paragraph IWA-5250(a)(2), the proposed repair does not meet all the requirements of IWA 4000.
Specifically, the Authority proposes to leave the pipe defect in place underneath the encapsulated pipe repair. This repair technique does not comply with IWA 4130(a)(2) i i
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wh ch requ res t at the repa r program contain the law remova met o and inet od of measurement of the cavity created by removing the flaw.
10 CFR 50.55a(a)(3) allows alternatives provided that, "(l) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements of this section would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety."
- 1. The proposed allematives would provide an acceptable lovel of quality and safety As discussed in the documentation provided in Attachment 3 including the Nuclear Safety Evaluation, the level of quality and safety has been maintained since the encapsulated pipe assembly restores structuralintegrity to the pipe containing the defect. The encapsulated pipe assembly has been designed to comply with the requirements of ASME Section Ill, as discussed in the Modification Package and the ASME Section XI Repair Prtgram provided in Attachment 3. The encapsulated pipe assembly with the pipe defect will be removed during the next refueling cutage to perform a thorough metallurgical -
cvaluation to complete the root cause evaluation.
- 2. Compliance with the requirements of ASME Section XI would result in hardship or unusual difficuttles without a compensating increase in the level of quality and safety.
Removal of the pipe defect requires that the affected Reactor Water Cleanup (RWCU)
Line be drained. The line with RWCU valve 12RWC-40 containing the flawed pipe ties into the Residual Heat Removal (RHR) Shutdown Cooling suction line which ties into "B" Reactor Recirculation System suction line. These lines cannot be normally isolated from the RPV pressure boundary. The location of this valve is shown in FSAR Figure 4.91 and the Reactor Water Recirculation System piping isometric drawing mark up provided as Attachment IV. Two methods of isolating this pipe were considered; installing a freeze -
seal, and installing a plug into the Recirculation System suction line.
Eteere_ Seal Installation of a freeze seal on the 6" RWCU line is impossible due to the piping configuration. This line rises vertically from the 20" RHR shutdown cooling suction line with valve 12RWC 46 located 8" from the Teo. This is insufficient distance to ensure an
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t adequate freeze seal. In addition, because the RWCU piping is vertical, convective currents would develop within the pipe while attempting to install the freeze seal which would draw the chilled water down into the RHR line. This would prohibit proper freezing of the line.
Elugginglhe_Beckculation.SyS10HLSucil0ILLine A plug could be installed to isolate the recirculation suction line and hence the RWCU line.
The plug must be installed from within the reactor vessel to isolate and drain this line.
This requires disassembly of the reactor vessel, removal of the reactor intemals (steam separator and dryer assernblies), and offload of over 1/3 of the nuclear fuel from the reactor core.
The following is a discussion on the costs associated with the RPV disassembly.
- 1. Disassembly of the Reactor Vessel would add approximately sixteen days to the present outage schedule. The lost revenue for this time period would be approximately G.4 million dollars.
- 2. The increased person rem exposure is estimated at 11.9 REM based on the recently completed RPV re-assembly. This is a significant increase in personnel exposure in comparison to the proposed altemative repair method. Additional exposure is anticipated for the fuel offload and reload.
- 3. Increased personnel costs as maintenance personnel and operations personnel are shifted from outage startup items to perform the following work tasks:
- RPV Disassembly
- VesselIntemals Removal
- Nuclear Fuel Offload
- Reactor Recirculation System Suction Plug installation
- Recirculation "B" Suction Line Draindown
- Reactor Recirculation System Suction Plug Removal
- Nuclear Fuel Reload
- RPV Assembly
. RPV Pressure Test
- System Tumover from Pressure Test Conditions to Startup Conditions.
The estimated man-hour cost, based on the personnel required, is over $130,000.
Based on the above discussion, removal of the flawed pipe would significantly delay the
present outage schedule, increase personnel radiation exposure and result in significant personnel costs to disassemble the reactor vessel to allow draindown of the line.
Pursuant to 10 CFR 50.55a(a)(3), the Authority requests relief from the ASME Section XI requirement to remove the flawed pipe. The flawed pipe will be removed during the next scheduled refueling outage. The proposed repair is in compliance with all design requirements for one operating cycle, and the removal of the flawed pipe this outage would significantly increase costs and personnel radiation exposure as detailed above.
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