1CAN101301, Request for Relief from Volumetric Examination Frequency Requirements, Request for Relief ANO1-ISI-023
| ML13281A478 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/08/2013 |
| From: | Pyle S Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN101301 | |
| Download: ML13281A478 (8) | |
Text
1CAN101301 October 8, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Request for Relief from Volumetric Examination Frequency Requirements Request for Relief ANO1-ISI-023 Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51
Dear Sir or Madam:
Pursuant to 10 CFR 50.55a(a)(3)(ii), Entergy Operations, Inc. hereby requests NRC approval of the attached Inservice Inspection (ISI) Request for Relief for Arkansas Nuclear One, Unit 1 (ANO-1). This request is for the current fourth 10-year ISI interval. The interval began on March 31, 2008, and will end on May 30, 2017.
The request is associated with volumetric examination frequency requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Code Case N-770-1 as conditioned in the Final Rule 10 CFR 50.55a(g)(6)(ii)(F)(3), dated June 21, 2011. Specifically, the request is to extend the frequency of the volumetric examination of the core flood dissimilar metal piping-to-nozzle welds from 84 months to 96 months. The justification for this request is provided in the attached request for relief.
This submittal contains no regulatory commitments.
If you have any questions or require additional information, please contact me.
Sincerely, Original signed by Stephenie L. Pyle SLP/rwc
Attachment:
Request for Relief - ANO1-ISI-023 Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Licensing Arkansas Nuclear One
1CAN101301 Page 2 of 2 cc:
Mr. Steven A. Reynolds Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852
ATTACHMENT TO 1CAN101301 Request for Relief - ANO1-ISI-023
Attachment to 1CAN101301 Page 1 of 5 REQUEST FOR RELIEF ANO1-ISI-023 Components / Numbers:
Core Flood Nozzle to Safe End Dissimilar Metal (DM) Welds01-025 and 01-026 Code Classes:
American Society of Mechanical Engineers (ASME) Code Class 1
References:
ASME Section XI 2001 Edition w/ 2003 Addenda, Table IWB-2500-1 ANO Unit 1 Risk Informed Inservice Inspection (ISI) Program (based in part on ASME Code Case N-716) 10 CFR 50.55a ASME Code Case N-770-1, Table 1, Inspection Item B Examination Category:
R-A Item Number(s)
R1.11/15
==
Description:==
Pressure Retaining Welds in Piping - Inspection Program B Unit / Inspection Interval Applicability:
Arkansas Nuclear One, Unit 1 (ANO-1) / Fourth 10-Year Interval I.
CODE REQUIREMENTS ASME Section XI, Table IWB-2500-1, Examination Category B-J, Pressure Retaining Welds in Piping - Inspection Program B and Code Case N-716:
- 1) Code Item B9.11, as designated by the risk-informed process of ASME Code Case N-716 to R-A, Item R1.11/15, requires a volumetric examination of circumferential piping welds nominal pipe size 4 or larger, as depicted in Figures IWB-2500-8 and Risk-Informed Inservice Inspection Evaluation Procedure, EPRI Report No. TR-112657; Topical Report, Revision B-A, June 1999. Surface examination is no longer required with the implementation of ASME Code Case N-716.
ASME Code Case N-770-1 requires subsequent volumetric examination of all Inspection Item B welds, as defined in Table 1 of the code case, at a frequency of every second inspection period not to exceed seven years.
Attachment to 1CAN101301 Page 2 of 5 II.
RELIEF REQUEST Pursuant to 10 CFR 50.55a(a)(3)(ii), Entergy Operations, Inc. (Entergy) requests relief from performing the required volumetric examinations of the components identified above at the frequency prescribed in ASME Code Case N-770-1.
III.
BASIS FOR RELIEF Ultrasonic examination of the core flood DM piping-to-nozzle welds listed above was performed during the refueling outage (1R21) in November 2008. The examination was performed remotely from the inside diameter of the nozzles due to interferences restricting access to the outside diameter of these welds. ASME Appendix VIII, Supplement 10 (Performance Demonstration Initiative) procedures, equipment, and personnel were utilized to perform the examinations.
100% of the code required volume was examined with no recordable indications detected. The performance of these examinations satisfied the then current ISI program requirements and met the subsequent baseline examination requirements of code case N-770-1, Paragraph 2200.
Referencing Inspection Item B on Table 1 of ASME Code Case N-770-1, subsequent volumetric examination of the core flood piping-to-nozzle welds is required to be performed every other period but not to exceed seven years (84 months). The examinations performed in 2008 were first period, fourth 10-year interval examinations; therefore, the subsequent examinations are due again in the third period, fourth 10-year interval but no later than November 2015.
The ANO-1 10-year reactor vessel inspection required by ASME Section XI, Table IWB-2500-1 is scheduled to be performed during refueling outage 1R26, currently scheduled for September 2016. While the reactor vessel internals are removed, Entergy plans to perform a reactor vessel internals examination as required by MRP-227, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines as part of the requirements related to license renewal.
Since examining the core flood welds remotely requires the removal of the reactor vessel internals from the reactor vessel in order to access the inner surface of the welds, Entergy plans to align the subsequent ASME Code Case N-770-1 core flood weld examinations with the 10-year reactor vessel examination and the MRP-227 required reactor vessel internals examination.
Aligning these examinations greatly reduces costs and dose that would be required to remove the internals in consecutive outages. The core barrel was last removed during the 1R21 refueling outage for the reactor vessel ISI examination. The dose received for just the removal and replacing the core barrel was approximately 330 mrem. Entergy believes that compliance with the specified requirements of ASME Code Case N-770-1 for this specific application results in a hardship without a compensating increase in the level of quality or safety.
Entergy is requesting that the frequency required by ASME Code Case N-770-1, Table 1 of 84 months (7 years), be extended to 96 months (8.0 years) for this
Attachment to 1CAN101301 Page 3 of 5 examination. This examination will still occur in the third period, fourth 10-year interval.
Additional Consideration for Basis for Relief During refueling outage 1R21 (Fall 2008) a modification was performed to the piping-to-reactor core flood nozzle DM weld joints to isolate the susceptible A600 material from reactor coolant. A weld onlay was designed to be installed remotely from the inside surface of the pipe weld to cover the DM weld and butter and extend to the base material on the stainless steel pipe and carbon steel /
stainless steel clad core flood nozzles.
The core flood DM weld joints consist of nominal 14-inch diameter stainless steel safe-end attached to a carbon steel / stainless steel clad nozzle that attaches to the reactor vessel. Original material is as follows:
Core flood nozzle forging A-508 CL.2 Nozzle cladding ER308L Stainless steel safe-end SA-336 GR. F8M DM weld and butter layer Alloy 82/182 Prior to beginning the weld modification, the weld joints to both core flood nozzles A (weld 01-025) and B (weld 01-026) were examined to ISI requirements with qualified Appendix VIII ultrasonic examination procedures and personnel. These examinations were performed with robotic equipment deployed from inside the reactor vessel. The ultrasonic scans were performed from the inside surface of the pipe / nozzle weld on November 8, 2008, and accepted with no indications noted.
After the acceptable ISI examinations were complete, plugs were placed in the core flood nozzles from the reactor vessel side. The core flood piping was drained and severed upstream of the DM welds. Remote examination processes were deployed to perform physical, visual and surface examinations of the 01-025 and 01-026 welds and adjacent material on both sides of the welds.
Internal surface cleaning and preparation was performed to support the application of the weld onlay. Prior to performing any welding preparations, the inside surface was examined with remote liquid penetrant to verify no indications of Pressurized Water Stress Corrosion Cracking (PWSCC) or other unacceptable indications were present at the DM weld or butter region.
The onlay weld was applied to an area exceeding four inches on each side of the DM weld centerline. The first weld layer consisted of an ER308L stainless steel buffer layer applied to the safe-end and the cladding on the nozzle side, being careful not to contact the DM weld or butter. In the area of the DM weld / butter and on each side for approximately 0.25 inch, a bridge layer weld of Alloy 82 was installed. This first weld layer was examined physically, visually and with liquid penetrant and accepted prior to the next weld layer.
Attachment to 1CAN101301 Page 4 of 5 The next weld layer of the onlay was installed with Alloy 52M filler material. The weld was examined physically, visually and with liquid penetrant and accepted prior to the final weld layer.
The final layer of the onlay was installed with Alloy 52M filler material. The weld was examined physically, visually and with liquid penetrant for final acceptance.
After the final onlays for 01-025 and 01-026 were accepted visually and with liquid penetrant examinations, then an EPRI-demonstrated volumetric ultrasonic examination was performed and accepted to the requirements of ASME Section III for cladding. These examinations were completed on November 27, 2008.
After restoration of the core flood piping, the plugs were removed from the nozzles inside the reactor vessel. The ISI ultrasonic examinations were repeated on welds01-025 and 01-026 with the onlays installed as a new baseline examination. This Appendix VIII-qualified examination was performed and was acceptable with no indications noted.
The purpose of the onlay modification was to verify PWSCC was not present on either of the core flood nozzle DM welds and to subsequently isolate any susceptible material from the primary water environment. The multiple inspections applied to the onlay welding were to provide assurance of the barrier integrity.
Additionally, during the 2013 - 1R24 refueling outage, a plant industrial accident required the plant to be in an extended shutdown. The outage duration from the March 24, 2013, shutdown until the August 7, 2013, startup was approximately 4.5 months. This added shutdown time reduces the service time of the components accordingly.
Entergy is not requesting NRC acceptance of the onlay described above as a qualified mitigation in accordance with Code Case N-770-1 and 10 CFR 50.55a(g)(6)(ii)(F)(2). Entergy is only providing the above information regarding the onlays that were installed in 2008 as supplemental information that provides some compensation for any risk associated with extending the inspection interval this one time.
IV.
PROPOSED ALTERNATIVE EXAMINATIONS No alternative testing is proposed at this time.
Attachment to 1CAN101301 Page 5 of 5 V.
CONCLUSION 10 CFR 50.55a(g)(6)(i) states:
The Commission will evaluate determinations under paragraph (g)(5) of this section that Code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines 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.
Entergy believes that extending the examination frequency by twelve months to align with an outage that has a planned full core offload and 10 year reactor vessel inspection is similar to and consistent with the rules of ASME Section XI, Subsection IWA-2430(c)(2). Therefore, Entergy requests the proposed relief be authorized pursuant to10 CFR 50.55a(a)(3)(ii).