RNP-RA/10-0056, Response to NRC Request for Additional Information Regarding Review of Refueling Outage 25 Inspections of the Reactor Vessel Nozzle Dissimilar Metal Butt Welds

Response to NRC Request for Additional Information Regarding Review of Refueling Outage 25 Inspections of the Reactor Vessel Nozzle Dissimilar Metal Butt Welds

ML101800267
Person / Time
Site:	Robinson
Issue date:	06/24/2010
From:	Castell C Carolina Power & Light Co, Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/10-0056
Download: ML101800267 (7)

Text

10 CFR 50.55a

& Progress Energy Serial: RNP-RA/10-0056 JUN 2 4 2010 United States Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, Maryland 20852 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING REVIEW OF REFUELING OUTAGE 25 INSPECTIONS OF THE REACTOR VESSEL NOZZLE DISSIMILAR METAL BUTT WELDS Ladies and Gentlemen:

By letter dated March 26, 2010, the NRC requested that Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. (PEC), respond within 90 days to a request for additional information (RAI) regarding review of Refueling Outage 25 inspections of the reactor vessel nozzle dissimilar metal butt welds. Attachment I to this letter provides the RAI response for the H. B.

Robinson Steam Electric Plant, Unit No. 2.

If you have any questions concerning this matter, please contact me at (843) 857-1626.

Sincerely, Curtis A. Castell Supervisor - Licensing/Regulatory Programs CAC/ahv Attachments:

I. Response to NRC Request for Additional Information Regarding Review of Refueling Outage 25 Inspections of the Reactor Vessel Nozzle Dissimilar Metal Butt Welds II. Design of the Inlet and Outlet Nozzles and Safe-End Configurations c: Mr. L. A. Reyes, NRC, Region II Mr. T. J. Orf, NRC, NRR NRC Resident Inspector Progress Energy Carolinas, Inc.

Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550 MoL

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/10-0056 Page 1 of 4 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING REVIEW OF REFUELING OUTAGE 25 INSPECTIONS OF THE REACTOR VESSEL NOZZLE DISSIMILAR METAL BUTT WELDS NRC Question 1 Provide the as-built information (or design if the as-built information is not available) regarding the width (axial length) of the safe-ends and Alloy 82/182 dissimilar metal welds (DMW) in the hot and cold leg nozzles. Provide the as-built (or design) distance between the center line of the safe-end-to-pipe weld and the center line of the DMW for the hot and cold legs.

Response

The design information for the inlet and outlet nozzles and safe-end configurations are provided in Attachment II. The pipe-to-safe-end weld was a field weld made by Ebasco Services using a 5/32 inch EB Type A consumable insert. This configuration would make the centerline of the weld 5/64 inch-beyond the dimension shown for the safe-end in Attachment II.

NRC Question 2 The December 8, 2008 letter, Attachment, page 4, states that the hot leg nozzles were not buttered before welding to the safe-ends. The staff would like to understand the fabrication details of the hot and cold leg nozzles.

a. Describe the fabrication process for the hot and cold leg nozzles, including welding of the DMW, safe-end, and the associated pipe. Discuss the heat treatment of the hot leg and cold leg nozzles including the safe-ends (i.e., were the safe-ends heat treated?). How many hours were the nozzles heat treated and at what temperature.

b. Discuss whether the safe-ends were fabricated with wrought or cast stainless steel.

c. Confirm that the cold leg nozzle was buttered and stress relieved prior to the DMW fabrication.

Response

Item 2a.

For the inlet (cold leg) nozzles, the fabrication process consisted of:

• The nozzle forgings were machined and weld clad on the inside diameter (ID) surface using 309 and 308 filler metal.

" The 221/2 degree weld bevel was machined, weld clad with 182 filler metal (Inconel), and remachined.

• The extensions (safe-ends) were then welded to the nozzles using Inconel.

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/10-0056 Page 2 of 4 Intermediate post weld stress reliefs were applied following the cladding and extension welds. These intermediate treatments are not detailed in the records since they were applied under another contract and there was no recording required by Code. Furnace data shows the "interstage" stress reliefs for the nozzles were half-an-hour holding times.

For the outlet (hot leg) nozzles, the fabrication process consisted of:

" The outlet nozzle forgings were procured, machined, and weld clad on the ID surface using 309 and 308 weld filler metal.

" The 221/2 degree weld bevel was machined, and the extensions (safe-ends) were welded to the nozzles using 182 (Inconel) filler metal. The outlet nozzle 22V2 degree bevels were not clad prior to the DMW to the nozzle.

• Intermediate post weld stress reliefs were applied similar to the inlet nozzles.

• The inlet and outlet nozzles were welded into the vessel shell and the vessel ID cladding was back-welded.

• The heat treatment of the completed vessel, nozzle extensions included, was a final stress relief for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> at 1150 degrees Fahrenheit +/-25 degrees Fahrenheit, for which the time-temperature recordings are available.

Item 2b.

The safe-ends were produced from forgings to SAl 82 Type 316 stainless steel.

Item 2c.

The Westinghouse design drawing E232-267 shows that the inlet nozzle forging was clad on the 22 degree 30 minute bevel prior to the DMW to the safe-end. This is confirmed by the Weld Inspection Record. The heat treatment of the inlet nozzles, after buttering and prior to the DMW, was done on the Malibu Vessel contract and is not evident in the "Furnace Logs."

Intermediate postweld heat treatment was allowed by code without the time-temperature recordings. The "Furnace Data" records show that the "interstage stress reliefs," done following the DMW of the extension to the nozzle, consisted of holding times of half an hour. The piece-specific time and temperature of these intermediate stress reliefs are not reported in the records.

NRC Question 3 Discuss the inspection results of the examinations performed in 2008, including any eddy current surface examinations conducted.

a. Discuss the locations and number of indications.

b. Confirm the axial flaws are located in the safe-end base material, not in the interface between the safe-end and DMW or between the safe-end and cladding. The staff understands that five circumferential flaws are wholly contained in clad material. The staff assumes that only the vessel nozzles are cladded and that DMW and safe-end are not cladded.

United States Nuclear Regulatory Commission Attachment Ito Serial: RNP-RA/1 0-0056 Page 3 of 4

c. Discuss the location of these five circumferential flaws in cladding with respect to the DMW. Based on the nozzle configuration drawings on Page 3 of the attachment to the December 8, 2008 letter, the DMW is in contact with the stainless steel nozzle cladding.

Therefore, discuss whether a circumferential flaw is located in the interface between the DMW and cladding.

d. Confirm that none of indications detected is connected to the inside surface of the pipe.

e. Discuss whether ultrasonic testing inspection of the hot and cold leg was performed from the outside diameter or inside diameter surface of the pipe during refueling outage 25 in the fall of 2008.

Response

Item 3a.

Flaw ASME B&PV Code WCAP-15621-NP Nozzle Detection Flaw Sizing Requirements Evaluation Flaw Evaluation Ultrasonic (UT) Ultrasonic Eddy NDE Ultrasonic [Axial (UT) Current Section XI Section XI Method (UT) Flaws] lCirc Flaws] (ECT) IWB-3500 IWB-3600 A (Hot Leg) 3 Axial Flaws in SS Outlet@ Detected Flaws Sized Not Required Connected Clad1 1 Axial Flaw in Weld 1300 1 Axial Flaw in Weld 7 Axial Flaws in SS Clad' B (Hot Leg) Flaws Flaws Sized Flaws Sized Flaws not ID 4 Axi 2 Axial Flaws in SE Outlet @ 100 Detected Connected 4 Axial Flaws in Weld 3 Cire Flaws in Clad' 4 Axial Flaws in SS C (Hot Leg) Flaws Flaws not ID Clad' 3 Axial Flaws in Weld Outlet @ Detected Flaws Sized Not Required Connected 3 Axial Flaws in Weld I Axial Flaw in SE 2500 1 Axial Flaw in SS Safe-End A (Cold Leg) Flaws not Not Not Required Not Inlet @ 800 Detected Required Required Not Required Not Required B (Cold Leg) Flaws Flaws Sized Flaws Sized Flaws not ID 2 Axial Flaws in Weld 3 1 Axial Flaw in Weld Inlet @ 320, Detected Connected 2 Circ Flaws in Clad' C (Cold Leg) Flaws Flaws Sized Not Required Not 3 Axial Flaws in Clad' Not Required Inlet @ 2000 Detected Required Note 1: Flaws wholly contained in the cladding do not require IWB-3600 evaluation per IWB-3514,1(d)(1).

Note 2: 2 of the 4 axial flaws are located in the SS safe-end and evaluated per Table IWB-3514-2, as are the SS weld flaws.

Note 3: 2 flaws evaluated as 1 per proximity rules of IWA-3000.

Item 3b.

The axial flaws are located in safe-end base material, the interface between the safe-end and DMW, and between the safe-end and cladding. Only the vessel nozzles are cladded and the DMW and safe-end are not cladded.

United States Nuclear Regulatory Commission Attachment Ito Serial: RNP-RA/1 0-0056 Page 4 of 4 Item 3c.

The 5 circumferential flaws are fully contained in the cladding and did not require IWB-3600 evaluation per IWB-3514.1 (d)(1).

Item 3d.

The indications that were detected are not connected to the inside surface of the pipe.

Item 3e.

Ultrasonic testing was performed on the hot and cold leg from the inside diameter surface of the pipe during Refueling Outage 25 (RO-25).

NRC Question 4 The licensee accepted some of the flaws for continued service based on analysis in accordance with IWB-3600 of the ASME Code,Section XI. IWB-2420 of the ASME Code,Section XI requires three successive examinations of the flaws remaining in service.

a. Provide the schedule (approximate month and year and refueling outage number) for the three successive examinations.

b. Discuss inspection plans (nondestructive examination techniques and the components) for the 2010 refueling outage and 2011 refueling outage.

Response

Item 4a.

In accordance with IWB-2420, re-examination of 4 of the 6 reactor vessel nozzles is currently scheduled for RO-27 (October 2011), RO-29 (October 2014) and RO-32 (April 2019). If mitigation occurs prior to successive examinations, successive examinations will not be required.

Item 4b.

Visual examination was performed in RO-26 (2010) as required by 10CFR50.55a for implementation of ASME Code Case N-722 and N-729-1. In RO-27 (2011), visual examinations are scheduled to occur on the hot leg DMW. In addition, volumetric examinations are scheduled to occur on the indications on the 4 nozzles.

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