RS-13-234, Relief Request I3R-ll Associated with Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds
| ML13263A372 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 09/19/2013 |
| From: | Gullott D Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-13-234 | |
| Download: ML13263A372 (7) | |
Text
1 4300 Wintle d Road Warrenville. IL 6055 ~
Exelon Generation 630657 2000 Office RS-13-234 10 CFR 50.55a September 19, 2013 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-n NRC Docket Nos. STN 50-456 and STN 50-457
Subject:
Relief Request 13R-ll Associated with Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i), Exelon Generation Company, LLC (EGC), is requesting NRC approval of a proposed alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda for Braidwood Station, Units 1 and 2.
The proposed alternative would permit the use of an alternative method of repair and nondestructive examination for control rod drive mechanism (CRDM) canopy seal welds. The CRDM assemblies were designed and fabricated to the ASME B&PV Code, Section 111,1974 Edition through Summer 1974 Addenda.
During boroscopic inspection of the reactor head assembly during the Braidwood Station, Unit 1 2013 fall refueling outage (Le., Al R17), white residue was observed on the CRDM canopy seal welds for reactor head penetrations 41, 49, 61, 65 and 73 indicating the potential of past reactor coolant system pressure boundary leakage at one or more of these locations. The locations of the leakage is suspected to be the omega seal welded threaded connection on one or more of these CRDM penetrations.
IWA*4000 of Section XI requires that repairs be performed in accordance with the original construction Code of the component or system, or later editions and addenda of the Code. The canopy seal weld is described in Section III and a repair to this weld would require: 1) an excavation of the rejectable indication(s); 2) a surface examination of the excavated area; 3) re-welding and restoration to the original configuration and materials; and 4) a final surface examination. An alternative to the Code repair process exists that provides an acceptable level of quality and safety, consistent with 10 CFR 50.55a(a)(3)(i). The alternative method also significantly reduces the projected occupational radiation dose when compared to the Code required repair method.
U. S. Nuclear Regulatory Commission September 19, 2013 Page 2 The alternative repair involves use of applicable portions of ASME Code Case N-504-4, "Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping,Section XI, Division 1." The Code Case will be used as guidance for repair by weld overlay to provide a new leakage barrier. Note that the repair material used will be nickel based Alloy 52M instead of austenitic stainless steel as specified in the Code Case. In lieu of a liquid penetrant examination (PT), a 5X or better magnified visual inspection is proposed. By eliminating the requirement to excavate the rejectable indication and allowing a magnified visual inspection to be performed in lieu of a PT examination, it is estimated that an occupational radiation dose savings of approximately 0.600 person-Rem will be realized.
EGC requests approval of the proposed alternative (Le., attached Relief Request 13R-11) by September 28, 2013 (Le., prior to Unit 1 entering Mode 4) in order to support the return to service from the current Braidwood Station, Unit 1 refueling outage. The proposed relief request would also be utilized for CRDM canopy seal weld repairs, should the need arise, during future outages within the Third Ten-Year Inservice Inspection Interval. The Third Ten-Year Inservice Inspection Interval for Braidwood Station is currently scheduled to end on July 28,2018for Unit 1 and October 16, 2018 for Unit 2.
There are no regulatory commitments contained within this letter.
Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2804.
Respectfully, David M. Gullott Manager - licenSing Exelon Generation Company, LLC
Attachment:
10 CFR 50.55a Relief Request 13R-11
151 Program Plan Braidwood Station Units 1 & 2, Third Interval 10 CFR 50.55a RELIEF REQUEST 13R-11 Revision 0 (Page 1 of 5)
Request for Relief for Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 Pursuant to 10 CFR 50.55a(a)(3)(i)
- 1. ASME CODE COMPONENT AFFECTED:
Code Class: 1
Reference:
IWA-4000 Examination Category: N/A Item Number: N/A
Description:
Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 Component Number: Reactor CRDM Canopy Seal Welds - Class 1 Appurtenance to the Reactor Vessel
- 2. APPLICABLE CODE EDITION AND ADDENDA:
The inservice inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.
- 3. APPLICABLE CODE REQUIREMENTS:
The CRDM assemblies were designed and fabricated to the ASME BP&V Code, Section 111,1974 Edition through Summer 1974 Addenda.
IWA-4000 of ASME Section XI requires that repairs be performed in accordance with the owner's original construction Code of the component or system, or later editions and addenda of the Code. The canopy seal weld is described in Section III, and a repair to this weld would require the following activities:
- a. Excavation of the rejectable indications,
- b. A surface examination of the excavated areas,
- c. Re-welding and restoration to the original configuration and materials, and
- d. Final surface examination.
lSI Program Plan Braidwood Station Units 1 & 2, Third Interval 10 CFR 50.55a RELIEF REQUEST 13R-11 Revision 0 (Page 2 of 5)
- 4. REASON FOR THE REQUEST:
The principal issues leading to this relief request are the excavation of indications contained within the existing weld, the accompanying dose received during the excavation and examination activities, and the weld material used for the repair or replacement.
Due to the nature of the flaw, the excavation of the leaking portion of the weld would necessitate a cavity that extends completely through wall. A liquid penetrant examination (PT) of this cavity is required to verify the removal of the rejectable flaw or to verify that the flaw is removed or reduced to an acceptable size. This PT examination would deposit the penetrant materials onto the inner surfaces of the component. This material would not be readily removed prior to re-welding due to the inaccessibility of the inside surface. The remaining penetrant material would introduce contaminants to the new weld metal and reduce the quality of the repair weld. The configuration of the canopy assembly would prevent the establishment and maintenance of an adequate back-purge during the welding process and would further reduce the quality of the repair weld.
The CRDM canopy seal welds are located above the Reactor Vessel Closure Head, which is highly congested and subject to high radiation levels. The high radiological dose associated with a CRDM canopy seal weld repair in strict compliance to these ASME Code requirements would be contrary to the intent of the as low as reasonably achievable (ALARA) radiological controls program. In order to reduce the exposure to personnel involved in the welding process, most of the repair activities would be performed remotely using robotic equipment to the extent practical. However, the required excavation and PT examinations would necessitate hands on access to the canopy weld. Based on expected radiation dose levels and time estimates to perform the excavation and PT examination for a single CRDM repair, the estimated total dose for these activities is estimated to be in excess of 0.600 person-Rem. This dose estimate is consistent with industry experience for similar activities.
IWA-4200 requires that the repair material conform to the original Design specification or Section III. In this case, the replacement material would have the same resistance to stress corrosion cracking as the original material. Use of the original material does not guarantee that the repaired component will continue to maintain leakage integrity throughout the intended life of the item.
Applicable portions of ASME Code Case N-504-4, "Alternative Rules for Repair of Class 1,2, and 3 Austenitic Stainless Steel Piping,Section XI, Division 1," will be used as guidance for repair by weld overlay to provide a new leakage barrier.
In lieu of performance of PT examinations of CRDM seal weld repairs or replacement, a 5X or better magnification visual examination will be performed after the welding is completed.
lSI Program Plan Braidwood Station Units 1 & 2, Third Interval 10 CFR 50.55a RELIEF REQUEST 13R-11 Revision 0 (Page 3 of 5)
In addition, Alloy 52/52M nickel-based weld repair material will be used rather than austenitic stainless steel as required by Code Case N-504-4.
Alloy 52/52M nickel-based weld repair material was selected rather than austenitic stainless steel as required by Code Case N-504-4 for the repair because of its resistance to stress corrosion cracking. Consequently, the ferrite requirements of Code Case N-504-4 do not apply. The suitability of the replacement material will be evaluated for each application and determined to be compatible with the existing component and will provide a leakage barrier for the remainder of the intended life of the CRDM.
The alternative method of repair is being requested to facilitate contingency repair efforts during the Unit 1 Fall 2013 refueling outage (Le., A1R17) and future outages within the Third Ten-Year Inservice Inspection Interval. The alternative nondestructive examination method is being requested to facilitate examination of a repair of a CRDM canopy seal weld during the Third Ten-Year Inservice Inspection Interval.
Industry experience with failure analyses performed on leaking canopy seal welds removed from service at other plants has attributed the majority of the cases to transgranular stress corrosion cracking (SCC). The size of the opening where the leakage occurs has been extremely small, normally a few thousandths of an inch. The crack orientations vary, but often radiate outward such that a pinhole appears on the surface, as opposed to a long crack. The SCC results from exposure of a susceptible material to residual stress, which is often concentrated by weld discontinuities, and to a corrosive environment, such as water trapped in the cavity behind the seal weld that is mixed with the air initially in the cavity, resulting in higher oxygen content than is in the bulk primary coolant.
- 5. PROPOSED ALTERNATIVE AND BASIS FOR USE:
Following the guidance of Code Case N-504-4, the CRDM canopy seal weld flaws will not be removed, but an analysis of the repaired weldment will be performed, prior to entering Mode 4, using Paragraph (g) of the Code Case as guidance to assure that the remaining flaw will not propagate unacceptably. The canopy seal weld is not a structural weld, nor a pressure-retaining weld, but provides a seal to prevent reactor coolant leakage if the mechanical joint leaks.
The weld buildup is considered a repair in accordance with IWA-411 O.
Applicability of the original Code of construction or design specification is mandated because the weld is performed on an appurtenance to a pressure-retaining component. The alternative CRDM canopy seal weld repair uses a Gas Tungsten Arc Welding (GTAW) process controlled remotely. Should the need arise, a manual GTAW repair may be utilized.
Code Case N-504-4 has been conditionally accepted by the NRC as documented in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability. ASME
lSI Program Plan Braidwood Station Units 1 & 2, Third Interval 10 CFR 50.55a RELIEF REQUEST 13R-11 Revision 0 (Page 4 of 5)
Section XI, Division 1," Revision 16. EGC has reviewed the condition associated with Code Case N-504-4, and determined that the noted condition does not apply to this application because the subject weld is a seal weld for an appurtenance.
The additional requirements outlined in the Regulatory Guide condition are only applicable to subsequent examinations of Class 1, 2, or 3 austenitic stainless steel pipe weldments with stress corrosion cracking.
A visual examination of the repaired/replaced weld will be performed using methods and personnel qualified to the standards of ASME VT-1 visual examination requirements. The VT-1 visual examination will be performed using a camera/viewing system with 5X or better magnification within several inches of the weld, qualified to ensure identification of significant flaws to assure an adequate margin of safety is maintained. The examination technique will be demonstrated to resolve a 0.001" thick wire against the surface of the weld. The repaired/replaced weld will be examined for quality of workmanship and discontinuities will be evaluated and dispositioned to ensure the adequacy of the new leakage barrier.
The automated GTAW weld repair and alternate VT-1 visual examination methods result in significantly lower radiation exposure because the equipment is remotely operated after setup. A post-maintenance VT-2 visual examination will be performed at normal operating temperature and pressure during the System Leakage Test in lieu of the hydrostatic test specified in Paragraph (h) of Code Case N-504-4.
Repair/Replacement activities, using the process described in this relief request, shall be documented on the required forms (Le., NIS-2, "Form NIS-2 Owner's Report for Repair/Replacement Activity," or NIS-2A, "Form NIS-2A Repair/Replacement Certification Record"). This relief request will be identified on the NIS-2/ NIS-2A forms in lieu of an adopted or invoked ASME Code Case.
The repair documents will be reviewed by the Authorized Nuclear Inspector, and maintained in accordance with the requirements for archiving permanent plant records.
- 6. DURATION OF PROPOSED ALTERNATIVE:
Relief is requested for the remainder of the Third Ten-Year Inspection Interval for Braidwood Station, Units 1 and 2, which is currently scheduled to end on July 28, 2018 for Unit 1 and October 16, 2018 for Unit 2.
- 7. PRECEDENTS:
Similar relief requests have been approved for this alternative. Recent approvals include the following:
- 1) Letter from Gloria Kulesa (U. S. NRC) to Bruce H. Hamilton (Duke Energy Carolinas, LLC), "McGuire Nuclear Station, Unit 1, Relief 08-MN-005, For
lSI Program Plan Braidwood Station Units 1 & 2, Third Interval 10 CFR 50.55a RELIEF REQUEST 13R-11 Revision 0 (Page 5 of 5)
Control Rod Drive Mechanism (CRDM) Canopy Seal Welds (TAC No.
MD9875)," dated October 14, 2009 (ADAMS Accession ML092530620)
- 2) Letter from Russell A. Gibbs (U. S. NRC) to Christopher M. Crane (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2 -
Evaluation of Relief Request 13R-06 for Control Rod Drive Canopy Seal Welds (TAC Nos. MD3863 and MD3864)," dated March 9, 2007 (ADAMS Accession ML070520480)
- 8.
REFERENCES:
- 1) ASME Code Case N-504-4, "Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping,Section XI, Division 1,"
July 14, 2006