Text

Response to Request for Additional Information Related to Relief Requests
	ML110320387
Person / Time
Site:	Clinton
Issue date:	02/01/2011
From:	Hansen J; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-11-015, TAC ME4185, TAC ME4183, TAC ME4184, TAC ME4186, TAC ME4187, TAC ME4188, TAC ME4189
	Download: ML110320387 (56)
	v • d • e

Exelon Generation 4300 Winfield Road Warrenville, IL 60555 RS-11-015 February 1, 2011 www.exeloncorp.com U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Nuclear 10 CFR 50.55a

Subject:

Response to Request for Additional Information Related to Relief Requests

References:

1.

Letter from Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Second Inservice Inspection Interval Relief Requests 4216, 4217, 4218, 4219, 4220, 4221, and 4222," dated July 1,2010

2.

Letter from U. S. NRC to M. J. Pacilio (Exelon Generation Company, LLC), "Clinton Power Station, Unit No.1 - Request for Additional Information Related to Second Inservice Inspection Interval Relief Requests 4216,4217,4218,4219,4220,4221, and 4222 (TAC Nos. ME4183, ME4184, ME4185, ME4186, ME4187, ME4188, and ME4189)," dated December 21,2010 In Reference 1, Exelon Generation Company, LLC (EGC) requested approval of relief requests associated with the second 10-year Inservice Inspection (lSI) Program Interval at Clinton Power Station, Unit 1 (CPS).

During its review of Reference 1, the NRC found that additional information was required to support its review as discussed in Reference 2. The requested information is provided in the attachment to this letter.

Exelon Generation 4300 Winfield Road Warrenville, IL 60555 RS-11-015 February 1, 2011 www.exeloncorp.com U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Nuclear 10 CFR 50.55a

Subject:

Response to Request for Additional Information Related to Relief Requests

References:

1.

Letter from Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Second Inservice Inspection Interval Relief Requests 4216, 4217, 4218, 4219, 4220, 4221, and 4222," dated July 1,2010

2.

Letter from U. S. NRC to M. J. Pacilio (Exelon Generation Company, LLC), "Clinton Power Station, Unit No.1 - Request for Additional Information Related to Second Inservice Inspection Interval Relief Requests 4216,4217,4218,4219,4220,4221, and 4222 (TAC Nos. ME4183, ME4184, ME4185, ME4186, ME4187, ME4188, and ME4189)," dated December 21,2010 In Reference 1, Exelon Generation Company, LLC (EGC) requested approval of relief requests associated with the second 10-year Inservice Inspection (lSI) Program Interval at Clinton Power Station, Unit 1 (CPS).

During its review of Reference 1, the NRC found that additional information was required to support its review as discussed in Reference 2. The requested information is provided in the attachment to this letter.

February 1, 2011 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter.

Should you have any questions concerning this letter, or require additional information, please contact Mitchel A. Mathews at (630) 657-2819.

Respectfully,

". aZ%n L ytA<v-~J Jeffrey'

~n Manag~ - Licensing and Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Response to NRC Request for Additional Information February 1, 2011 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter.

Should you have any questions concerning this letter, or require additional information, please contact Mitchel A. Mathews at (630) 657-2819.

Respectfully,

". aZ%n L ytA<v-~J Jeffrey'

~n Manag~ - Licensing and Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Response to NRC Request for Additional Information

Attachment Response to NRC Request for Additional Information Page 1 of 54 Generic Request Pertaining to RR Nos. 4216 through 4222

1. Please confirm that all examinations included in [relief request] RR Nos. 4216 through 4222 have been completed in full. If any have not been completed in full, please note exactly what has been completed.

Exelon Generating Company. LLC (EGC) Response All examinations included in Relief Request (RR) Nos. 4216 through 4222 have been completed to the maximum degree possible as permissible by the limitations/interferences described in the applicable relief requests.

RR No. 4216 - Surface Examination of Residual Heat Removal Pump A Casing Weld

2. The illustrations provided in the submittal do not make clear how the instrument line obstructs the required examination. Please include a visual representation that clearly illustrates the obstruction.

EGC Response Non-destructive examination (NDE) report ROO-011 states that for weld RHR-A-2, the weld length is 100 inches with no inspection of 13 inches (Le., from 58.75 inches to 71.75 inches) due to an instrument lined welded to the piping weld. This weld is the "A" residual heat removal (RHR) pump shell-to-upper-f1ange weld; see Figures 1 and 2 below. The obstruction is caused by a combination of the A RHR pump suction vent line, which is welded to the RHR-A-2 weld with a fillet weld around the circumference of the nozzle, and the 1A seal cooler, which is up against the shell-to-upper-flange weld. This fillet weld and the associated seal cooler are obstructing the RHR flange weld from being inspected with the surface examination method. Figures 1 and 2 below provide a visual representation that clearly illustrates the obstruction.

Attachment Response to NRC Request for Additional Information Page 1 of 54 Generic Request Pertaining to RR Nos. 4216 through 4222

1. Please confirm that all examinations included in [relief request] RR Nos. 4216 through 4222 have been completed in full. If any have not been completed in full, please note exactly what has been completed.

Exelon Generating Company. LLC (EGC) Response All examinations included in Relief Request (RR) Nos. 4216 through 4222 have been completed to the maximum degree possible as permissible by the limitations/interferences described in the applicable relief requests.

RR No. 4216 - Surface Examination of Residual Heat Removal Pump A Casing Weld

2. The illustrations provided in the submittal do not make clear how the instrument line obstructs the required examination. Please include a visual representation that clearly illustrates the obstruction.

EGC Response Non-destructive examination (NDE) report ROO-011 states that for weld RHR-A-2, the weld length is 100 inches with no inspection of 13 inches (Le., from 58.75 inches to 71.75 inches) due to an instrument lined welded to the piping weld. This weld is the "A" residual heat removal (RHR) pump shell-to-upper-f1ange weld; see Figures 1 and 2 below. The obstruction is caused by a combination of the A RHR pump suction vent line, which is welded to the RHR-A-2 weld with a fillet weld around the circumference of the nozzle, and the 1A seal cooler, which is up against the shell-to-upper-flange weld. This fillet weld and the associated seal cooler are obstructing the RHR flange weld from being inspected with the surface examination method. Figures 1 and 2 below provide a visual representation that clearly illustrates the obstruction.

Attachment Response to NRC Request for Additional Information Page 2 of 54 Figure 1: Photograph of Obstructed Portion of the A RHR Pump Casing Weld Attachment Response to NRC Request for Additional Information Page 2 of 54 Figure 1: Photograph of Obstructed Portion of the A RHR Pump Casing Weld

Attachment Response to NRC Request for Additional Information Page 3 of 54 Figure 2: Alternate View of Obstructed "A" RHR Pump Casing Weld

3. It was noted that to increase inspection coverage would require staff at Clinton Power Station, Unit No.1 to "cut out the noted instrument line and weld it back." However, it was not made clear if alternate modes of inspection, such as miniaturized cameras, were used to attempt to obtain greater coverage despite the blocking instrument line.

Please discuss the efforts used to maximize inspection coverage without plant modification and why these efforts did, or did not, lead to increased coverage.

EGC Response A review was performed to determine if other techniques such as visual examination could be performed to inspect the obstructed 13 inches of weld RHR-A-2. No alternate techniques could be used to exam the obstructed area due to the fillet weld and the clearances associated with the seal cooler without performing a plant modification and removing either the seal cooler and/or the suction vent line.

Attachment Response to NRC Request for Additional Information Page 3 of 54 Figure 2: Alternate View of Obstructed "A" RHR Pump Casing Weld

3. It was noted that to increase inspection coverage would require staff at Clinton Power Station, Unit No.1 to "cut out the noted instrument line and weld it back." However, it was not made clear if alternate modes of inspection, such as miniaturized cameras, were used to attempt to obtain greater coverage despite the blocking instrument line.

Please discuss the efforts used to maximize inspection coverage without plant modification and why these efforts did, or did not, lead to increased coverage.

EGC Response A review was performed to determine if other techniques such as visual examination could be performed to inspect the obstructed 13 inches of weld RHR-A-2. No alternate techniques could be used to exam the obstructed area due to the fillet weld and the clearances associated with the seal cooler without performing a plant modification and removing either the seal cooler and/or the suction vent line.

Attachment Response to NRC Request for Additional Information Page 4 of 54 RR No. 4217 - Reactor Pressure Vessel (RPY) Head-to-Flange Weld

4. The submittal did not identify how the variation with configuration/geometry of RPV head flange across the length of the weld produced such varied coverage percentages. Can these coverage differences be contributed entirely to the near field effects? Please specify exactly what variations along the weld length reduced the weld coverage for this weld.

EGC Response The inspection of the reactor top head to flange weld, CH-C-2, was separated into three segments with one segment being examined in 2002, 2006, and 2010. General Electric (GE)-Hitachi was the vendor performing all three of the inspections.

Report 02-002 from 2002 utilized procedure UT-EXLN-300, Version 4, "Procedure for Manual Examination of Reactor Vessel Assembly Welds," which was written to meet the requirements listed in NRC Regulatory Guide (RG) 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," and American Society of Mechanical Engineers (ASME) Section XI-1989 Edition. The inspection was performed and coverage calculated as 53.8% from the 0, 45, and 60 degree angles utilized to inspect from o to 120 degrees as shown in Figure 3 below.

Report 06-062 from 2006 utilized procedure GE-UT -300, Version 10, "Procedure for Manual Examination of Reactor Vessel Assembly Welds in Accordance with POI," which was written to meet ASME Section XI, Appendix VIII requirements. The inspection was performed and coverage calculated as 79.5% from the 60 degree "Near Surface" and "Full Volume" techniques demonstrated for Appendix VIII and shown in Figure 4 below.

Report C1R12-046 from 2010 utilized procedure GE-UT-300 Version 10 written to meet ASME Section XI, Appendix VIII requirements. The inspection was performed and coverage calculated as 840/0 from the 60 degree "Near Surface" and "Full Volume" techniques demonstrated for Appendix VIII and shown in Figure 5 below. The change in coverage from 79.5%) in 2006 to 840/0 in 2010 was due to the difference in defining the process used to calculate the coverage and providing physical measurements of transducer location versus weld location to give better definition to the calculations performed.

After reviewing all three reports there is no configuration or geometry changed that affected the coverage percentages. The differences were due to changes in the examination process employed (Le., using Appendix VIII of the ASME Code techniques) and coverage calculation methodology.

Attachment Response to NRC Request for Additional Information Page 4 of 54 RR No. 4217 - Reactor Pressure Vessel (RPY) Head-to-Flange Weld

4. The submittal did not identify how the variation with configuration/geometry of RPV head flange across the length of the weld produced such varied coverage percentages. Can these coverage differences be contributed entirely to the near field effects? Please specify exactly what variations along the weld length reduced the weld coverage for this weld.

EGC Response The inspection of the reactor top head to flange weld, CH-C-2, was separated into three segments with one segment being examined in 2002, 2006, and 2010. General Electric (GE)-Hitachi was the vendor performing all three of the inspections.

Report 02-002 from 2002 utilized procedure UT-EXLN-300, Version 4, "Procedure for Manual Examination of Reactor Vessel Assembly Welds," which was written to meet the requirements listed in NRC Regulatory Guide (RG) 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," and American Society of Mechanical Engineers (ASME) Section XI-1989 Edition. The inspection was performed and coverage calculated as 53.8% from the 0, 45, and 60 degree angles utilized to inspect from o to 120 degrees as shown in Figure 3 below.

Report 06-062 from 2006 utilized procedure GE-UT -300, Version 10, "Procedure for Manual Examination of Reactor Vessel Assembly Welds in Accordance with POI," which was written to meet ASME Section XI, Appendix VIII requirements. The inspection was performed and coverage calculated as 79.5% from the 60 degree "Near Surface" and "Full Volume" techniques demonstrated for Appendix VIII and shown in Figure 4 below.

Report C1R12-046 from 2010 utilized procedure GE-UT-300 Version 10 written to meet ASME Section XI, Appendix VIII requirements. The inspection was performed and coverage calculated as 840/0 from the 60 degree "Near Surface" and "Full Volume" techniques demonstrated for Appendix VIII and shown in Figure 5 below. The change in coverage from 79.5%) in 2006 to 840/0 in 2010 was due to the difference in defining the process used to calculate the coverage and providing physical measurements of transducer location versus weld location to give better definition to the calculations performed.

After reviewing all three reports there is no configuration or geometry changed that affected the coverage percentages. The differences were due to changes in the examination process employed (Le., using Appendix VIII of the ASME Code techniques) and coverage calculation methodology.

Attachment Response to NRC Request for Additional Information Page 5 of 54 Figure 3: Reactor Pressure Vessel (RPV) Head-to-Flange Weld Attachment Response to NRC Request for Additional Information Page 5 of 54 Figure 3: Reactor Pressure Vessel (RPV) Head-to-Flange Weld

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Attachment Response to NRC Request for Additional Information Page 8 of 54

5. Please explain why the near field effects are problematic in only one-third of the weld length.

EGC Response The near field effect was only noted on report 02-002 as being problematic in the examination from 0 to 120 degrees. This 2002 examination was performed in accordance with the 1989 Edition of ASME Section XI and RG 1.150. The techniques utilized during 2002 did not have a process to demonstrate the ability of the equipment to see near surface flaws. The remaining 240 degrees were examined utilizing equipment and techniques demonstrated in accordance with Appendix VIII of the ASME Code for near surface examinations.

RR No. 4218 - RPV Shell-to-Flange Weld

6. There is no clear identification of how the RPV shell flange configuration/geometry varies through the length of the weld to produce such varied coverage percentages.

Can these coverage differences be contributed entirely to the near field effects?

Please specify exactly what variations reduced the weld coverage for this weld.

EGC Response The inspection of RPV-C5, RPV Vessel to Flange weld was separated into two segments with one half being examined in 2002 and the other in 2010. GE-Hitachi was the vendor performing both of the inspections.

Report 02-003 from 2002 utilized procedure UT-EXLN-300, Version 4, which was written to meet the requirements in accordance with RG 1.150 and ASME Section XI-1989 Edition.

The inspection was performed and coverage calculated to be 52.2% from the 0, 45, and 60 degree angles utilized to inspect from 0 to 180 degrees as shown in Figure 6 below.

Report C1R12-061 from 2010 utilized procedure GE-UT-300 Version 10 which was written to meet ASME Section XI, Appendix VII I requirements. The inspection was performed and coverage calculated to be 750/0 from the 60 degree "Near Surface" and "Full Volume" techniques demonstrated for Appendix VIII and shown in Figure 7 below.

After reviewing both reports there is no configuration or geometry changed that affected the coverage percentages. The differences were due to changes in the examination process employed (Le., using Appendix VIII of the ASME Code techniques) and coverage calculation methodology.