RS-04-109, Units 1 & 2 and Three Mile Island, Unit 1 - Initial Response to NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized.

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Units 1 & 2 and Three Mile Island, Unit 1 - Initial Response to NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized.
ML042170341
Person / Time
Site: Byron, Three Mile Island, Braidwood  Constellation icon.png
Issue date: 07/27/2004
From: Jury K
AmerGen Energy Co, Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-04-20180, BL-04-001, RS-04-109
Download: ML042170341 (65)
v • d • e


Text

7 - -.- - _t I - Z-r AmerGen b Exekmn.

www.exelonCOTP.COM An Exelon Company AmerGen Energy Company, LLC Nuclear 4300 Winfield Road Exelon Generation Warrenville, IL 60555 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.54 (f)

RS-04-109 5928-04-20180 July 27, 2004 United States Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, Maryland 20852 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 Three Mile Island Nuclear Station, Unit No. 1 Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Initial response to NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized-Water Reactors"

Reference:

NRC Bulletin 2004-01, "Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized-Water Reactors," dated May 28, 2004 The purpose of this letter is to provide the Exelon Generation Company, LLC and the AmerGen Energy Company, LLC initial, sixty-day, response to the referenced NRC bulletin. The responses to NRC Bulletin 2004-01 questions 1(a), 1(b), 1(c), and 1(d) detailing the description and fabrication of the pressurizer connections; the current and proposed inspection program for the components; and the basis for assuring reactor coolant pressure boundary integrity for pressurizer penetrations and steam space connections are provided in the attachments to this letter.

,At

U. S. Nuclear Regulatory Commission July 27, 2004 Page 2 Should you have any questions concerning this letter, please contact David J.

Chrzanowski at (630) 657-2816.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 2720L4 A /.X Keith R. Jury Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC AmerGen Energy Company, LLC Attachments: Attachment 1, Initial Response to NRC Bulletin 2004-01, Braidwood Station, Units 1 and 2 Attachment 2, Initial Response to NRC Bulletin 2004-01, Byron Station, Units 1 and 2 Attachment 3, Initial Response to NRC Bulletin 2004-01, Three Mile Island Nuclear Station, Unit 1

aI- -J Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Page 1 of 18

Attachment I Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units I and 2 Requested Information 1 (a)

A description of the pressurizer penetrations and steam space piping connections at your plant. At a minimum, this description should include materials of construction (e.g.,

stainless steel piping and/or weld metal, Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.), joint design (e.g., partial penetration welds, full penetration welds, bolted connections, etc.), and, in the case of weldedjoints, whether or not the weld was stress-relieved prior to being put into service. Additional information relevant with respect to determining the susceptibility of your plant's pressurizer penetrations and steam space piping connections to PWSCC should also be included.

Response

Braidwood Station Unit 1 and Unit 2 are 4-loop pressurized water reactors with the nuclear steam supply system designed by Westinghouse Electric Company, LLC (Westinghouse). The Braidwood Station units have one Westinghouse Model 84 pressurizer in each unit. Braidwood Station, Unit 1 began commercial operation on July 29, 1988. Braidwood Station, Unit 2 began commercial operation on October 17, 1988.

The pressurizers were fabricated by Westinghouse at their Pensacola, Florida facility.

Each of the pressurizers has six piping connections: five steam space connections and one surge line connection. The five steam space connections are: one spray line connection, one relief valve line connection, and three safety valve line connections.

The pressurizer manways (one per pressurizer) are not discussed in this response since, they do not contain any Alloy 600/82/182. Table'1 lists-the pressurizer steam space connections and corresponding identifier.

Table I Braidwood Station Pressurizer Steam Space Connection Listing Unit Steam Space Connection Identifier 1 safety valve line connection 1PZR-01-SE-02 1 safety valve line connection 1PZR-01-SE-03 1 safety valve line connection 1PZR-01-SE-04 1 relief valve line connection 1PZR-01-SE-05 I spray line connection 1PZR-01 -SE-06 2 safety valve line connection 2PZR-01-SE-02 2 safety valve line connection 2PZR-01-SE-03 2 safety valve line connection 2PZR-01-SE-04 2 relief valve line connection 2PZR-01-SE-05 2 spray line connection 2PZR-01-SE-06 All six of the pressurizer connections have an Alloy 82/182 weld connection from the low alloy steel pressurizer nozzles to the stainless steel safe-end attachments. These six locations are the only Alloy 600/82/182 based connections on the Braidwood Station pressurizers. Piping connections downstream of these safe-end connections do not have any Alloy 600/82/182 based components in any application. Also, the eight, 3/4 inch instrument line connections and the 3/4 inch sample line connection do not incorporate Page 2 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 any Alloy 600/82/182 based components in their connections to the pressurizer or in any of their downstream connections. The Braidwood Station pressurizers have heater penetrations that use stainless steel sleeves connected to the pressurizer by stainless steel welds.

Pressurizer Steam Space Connections - Braidwood Station. Unit 1 Connection 1PZR-01-SE-02 1PZR-01-SE-02 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "A" pressurizer safety relief valve (1RY8001 A) to the pressurizer upper head nozzle (1PZR-01-N4A). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld IPZR-01-SE-02, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment. A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 1PZR-01-SE-03 1PZR-01-SE-03 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the 'Bi"pressurizer safety relief valve (1RY8001 B) to the pressurizer upper head nozzle (1PZR-01-N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-182, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

Page 3 of 18

7 Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 During fabrication, weld 1PZR-01-SE-03, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were five repairs performed on this weld: 2.25" long and 3/8" deep from the inside diameter (ID) between radiography location marks 0 to 1; 1.5" long and 11/16" deep from the outside diameter (OD) between radiography location marks 5 to 6; 0.75" long and 11/16" deep from the OD between radiography location marks 6 to 7; 1" long and 11/16" deep from the OD also between radiography location marks 6 to 7; and 1.25" long and 1/4" to 3/8" deep from the OD between radiography location marks 11 to 0.

Subsequent to the repairs, a final, acceptable, radiography examination (RT) was performed.

Connection IPZR-01-SE-04 1PZR-01 -SE-04 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "C" pressurizer safety relief valve (I RY8001 C) to the pressurizer upper head nozzle (1PZR-01 -N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade'F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 1PZR-01-SE-04, identified in shop records as weld 4C on nozzle A, did not receive any post-weld heat treatment. A review of the fabrication records indicated that there were five spot repairs performed on this weld. The records did not specify the length for any of the repair areas which were located between radiography location marks: 2 to 3, 4 to 5, 5 to 6, 6 to 7, and 7 to 8. All repairs were listed as being 0.2" deep from the OD.

Subsequent to the repairs, a final, acceptable, RT was performed.

Connection 1PZR-01-SE-05 1PZR-01 -SE-05 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (1PZR-01-N2).

Page 4 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182,.Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Up stream of the nozzle to safe-end weld is a weld that joins the four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the SA-403, Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule 160 piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455B and 1RY455C (SA-403, Grade WP304 valve bodies).

During fabrication, weld 1PZR-01-SE-05, identified in shop records as weld 2 on nozzle E, did not receive any post-weld heat treatmnent. A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 1PZR-01-SE-06 '

1PZR-01 -SE-06 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (.1 RY455A and 1RY456) to the pressurizer upper head nozzle (1PZR-01 -N3). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304, 450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455A and 1RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld 1PZR-01-SE-06, identified in shop records as weld 3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Page 5 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Pressurizer Steam Space Connections - Braidwood Station, Unit 2 Connection 2PZR-01-SE-02 2PZR-01-SE-02 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "A" pressurizer safety relief valve (2RY8001 A) to the pressurizer upper head nozzle (2PZR-01-N4A).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160,' SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 2PZR-01-SE-02, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment.

Additional Information for 2PZR-01-SE-02 A review of the fabrication records indicated that there were four repairs performed on this weld. The records did not specify the length for any of the repair areas which were located between radiography location marks: 6 to 7, 10 to 11, 11 to 12, and 0 to 1. All repairs were listed as being approximately 0.5" deep from the OD.

Subsequent to the repairs, a final, acceptable, RT was performed.

Connection 2PZR-01-SE-03 2PZR-01-SE-03 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "B" pressurizer safety relief valve (2RY8001 B) to the pressurizer upper head nozzle (2PZR-01-N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Page 6 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 2PZR-01-SE-03, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there was one repair performed on this weld. The repair area is listed as being 0.5" long and approximately 0.625" deep from the ID between radiography location mark 5 to 6.

Subsequent to the repair, a final, acceptable, RT was performed.

Connection 2PZR-01-SE-04 2PZR-01-SE-04 is a six inch nominal pipe diameter butt welded connection joiniing the pipe that connects the UC" pressurizer safety relief valve (2RY8001 C) to the pressurizer upper head nozzle (2PZR-01-N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 2PZR-01-SE-04, identified in shop records as weld 4C on nozzle A, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 2PZR-01-SE-05 2PZR-01-SE-05 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (2PZR-01-N2).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal Page 7of 18

V Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Up stream of the nozzle to safe-end weld is a weld that joins tlee four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the SA-403, Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455B and 2RY455C (SA-403, Grade WP304 valve bodies).

During fabrication,'weld 2PZR-01-SE-05, identified in shop records as weld 2 on nozzle E, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 2PZR-01-SE-06 2PZR-01-SE-06 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (2RY455A and 2RY456) to the pressurizer upper head nozzle (2PZR-01-N3). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304,450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455A and 2RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld 2PZR-01-SE-06, identified in shop records as weld 3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Page 8 of 18

Attachment 1 InitiaI Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Requested Information 1 (b)

A description of the inspection program for Alloy 82/182/600 pressurizer penetrations and steam space piping connections that has been implemented at your plant. The description should include when the inspections were performed; the areas, penetrations and steam space piping connections inspected; the extent (percentage) of coverage achieved for each location which was inspected; the inspection methods used; the process used to resolve any inspection findings; the quality of the documentation of the inspections (e.g., written report, video record, photographs); and, the basis for concluding that your plant satisfies applicable regulatory requirements related to the integrity of pressurizerpenetrations and steam space piping connections. If leaking pressurizerpenetrations or steam space piping connections were found, indicate what followup NDE was performed to characterize flaws in the leaking penetrations.

Response

The Alloy 82/182/600 pressurizer penetration and steam space connections at Braidwood Station are limited to the ten (five per unit) low alloy steel nozzle to stainless steel, Alloy 82, welded safe-end connections. All the pressurizer welds subject to the actions of this bulletin (identified in Table 2) have had volumetric and surface examinations performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components,' Code Category B-F,1 Code Item number B5.40.2 The first inservice inspection (ISI) Interval examinations were performed in accordance with the 1983 Edition, Summer 1983 addenda, of the ASME Code Section Xl. Examinations performed in the second ISI interval were performed in accordance with the 1989 Edition, no addenda, of the ASME Code Section Xl. The examination areas for these welds were those identified in the ASME Code Section Xl, Figure IWB-2500-8, 'Similar and Dissimilar Metal Welds in Components and Piping.'

This examination area is the same for the ASME Code Section Xl Editions 1983 through 1989.

Both Braidwood Station units are now in the 2nd Interval with the IS program, as stated above, based on the 1989 Edition of the ASME Code Section Xl. On February 20, 2001, the NRC approved the use of a risk informed methodology for the selection and examination of ASME Code Class 1 and Code Class 2 piping welds; however, none of the subject pressurizer safe-end welds have been re-examined under the risk informed ISI program. In addition, there have been no volumetric re-examinations performed on the pressurizer welds subject to the actions of this bulletin since the implementation of Supplement 1 to Appendix VII 4 of the ASME Code Section XI since its implementation in November of 2002.

1 "Pressure Retaining Dissimilar Metal Welds" 2 Pressurizer, Nozzle-to-Safe-end Butt Welds, Nominal Pipe Size 4 Inch or Larger" 3 Qualification Requirements for Dissimilar Metal Piping Welds" 4'Performance Demonstration for Ultrasonic Examination Systems" Page 9 of 18

Attachment 1 Inital Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 All examinations, surface and volumetric, were recorded on hard copy data sheets. No video or photographs were used to supplement these examinations. There have been no recordable indications on these ten welds that required disposition. All surface examinations covered 100% of the required examination area. The ultrasonic examinations, although partially obstructed in some cases, achieved greater than 90% of the required examination volume coverage. I In addition to the nondestructive examinations listed below, all steam space pressurizer Alloy 82 welds were visually examined, at a minimum, each refueling outage in accordance with the pressure test requirements of the ASME Code Section Xl, Category B-P. These examinations have, in the past, been performed with the insulation in place with a four-hour hold time at normal operating pressure and temperature.

There has not been any indication of leaking pressurizer penetrations or steam space piping connections at Braidwood Station and therefore, there has not been a need for dispositions, additional nondestructive examinations (NDE), examination expansions, or flaw characterizations.

Page 10 of 18

Attachment 1 ,.

Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Table 2 Braidwood Station Alloy 82/182 Pressurizer Welds Examination History Weld Exam Date Exam Method Exam Technique Results Disposition/

Identifier . Data Sheet Volumetric (Vol) 450 longitudinal wave-transducer No Recordable Indication(s) (NRI) N/A 1PZR-01-SE-02 October 1995 (450 L-wave) ._95BR1_UT-052 Surface (S) Dye Penetrant (PT) NRI N/A 95BR1PT-026 N/A Vol 450 L-wave and 700 shear wave NRI 94BR1-UT-061 and IPZR-01-SE-03 March 1994 transducer (S-wave) Ni94BRI-UT-062' S PT Three arc strike Indications Acceptable

._ 94BR1-PT-021 N/A Vol 450 L-wave and 700 S-wave NRI 94BR1-UT-061 and 1PZR-01-SE-04 March 1994 . 94BR1-UT-062 S PT NRI N/A Acceptable Vol 450 L-wave 360° root geometry Indication 89BR1-UT-110 and October 1989 89BR1-UT-111 1PZR-01-SE-05 S PT NRI 89BR1-PT-056 Vol 450 L-wave NRI N/A 988R1-UTD-040 September 1998 N/A S S PTPT NRIN/

N~i98BR1-PT-039 Acceptable Vol 450 L-wave 3600 root geometry Indication 89BR1-UT-108 and October 1989 . 89BR1-UT-109 1PZR-01-SE-06 0 S PT .NRI TNi898R1-PT-057 N/A Vol 450 L-wave NRI N/A September 1998 98BR1-UTD-045 S PT -NRI N/A S PT N 98BR1-PT-040 Vol 450 L-wave NRI 95BR2-PT-064 2PZR-01-SE-02 April 1996 N/A S PT NRI 958R1-UT-156 Page 11 of 18

'I Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Table 2 Braidwood Station Alloy 821182 Pressurizer Welds Examination History (continued)

Weld Exam Date Exam Method Exam Technique -. Results Disposition/

Identifier . Data Sheet Vol 450 L-wave NRI 94BRN-UT-060 2PZR-01-SE-03 October 1994 N1A 2PZR-A1-SE-04 October 1994 9 NBR2-UT-/0A PNRI 94BR2-PT-038 Indication was dispositioned as Vol S 55° L-wave Pt NRI from the alloy Ultrasonic Indications non-relevantN/A and acceptable April 1990 821182 cladding Interface 90BR2-UT-100 90BR2-UT-101and s PT NRIN/A 2PZR-01-SE-05 NRI 90R2-PT-043 Pipe Inner diameter geometry Acceptable Vol 450 L-wave Indications 99BR2-UTD-034 April 1999

. PT NRI N/A 99BR2-PT-049 Indication was dispositioned as Vol55, -wveUltrasonic indications from the alloy non-relevant and acceptable Vol 55 11wave82/182 cladding Interface -90BR2-UT-102 and April199090BR2-UT-103 s PT NRIN/A 2PZR-01-SE-06 S TNI90BR2-PTO 048 Pipe inner diameter geometry Acceptable Vol 4501 L-wave' Indications 99BR2-UTD-035 April 1999 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

S PT_ _ _ _ _ _ _ _ _ _ _ _ __ _ _ __ _ _ __

NRI

_ _ __ _ _ _ _ _ _ _ _ _99 N/A.

B R 2-PT-0 5 Page 12 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Table 2 Braidwood Station Alloy 82/182 Pressurizer Welds Examination History Basis for Concluding Regulatory Requirements are SatisfiJd As stated above, the Braidwood Station pressurizer connections affected by this Bulletin are limited to the ten (five per unit), Alloy 82/182 full penetration nozzle to safe-end welds. The completion of volumetric, surface, and visual examinations without any evidence of recordable, relevant indications, through-wall leakage or any recordable wastage of the low alloy steel surface, is assurance of the prior integrity of the Alloy 82/182 connections.

Ongoing integrity of the Braidwood Station pressurizer steam space Alloy 82/182 connections is assured by performing 100% bare metal visual (BMV) examinations, at a minimum, each refueling outage (approximately 18 months). The examination will be performed until mitigation (i.e., weld overlays of the Alloy 82/182 locations) is performed on all pressurizer steam space connections.

The specific regulatory requirements are listed below with the associated response addressing how the requirement is met.

Compliance with Design Requirements: 10 CFR 50. Appendix A - General Design Criteria (GDC)

Criterion 14 - Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

The Braidwood Station pressurizer steam space connections are designed, fabricated, tested, and examined in accordance with the requirements of the ASME Code Section Ill, "Requirements for Design and Manufacture of Nuclear Power Plant Components,"

and the ASME Code Section Xl. In general, the controls established by these construction and inspection codes assure that the reactor coolant pressure boundary maintains an extremely low probability of rapidly propagating failure and gross rupture.

The BMV examination technique to be used in the Braidwood Station, Unit 1 Fall 2004 refueling outage, and in the Unit 2 Spring 2005 refueling outage, is a reliable means for identifying the very low leakage rates potentially associated with alloy 82/182 cracking.

Therefore, based on the design, materials, and examination methods, the Braidwood Station pressurizers continue to comply with the requirements of GDC 14.

In addition, in the case of the pressurizer steam space Alloy 82/182 locations, modifications in the form of weld overlays using primary water stress corrosion cracking (PWSCC) resistant material will be performed to provide added assurance that these connections will have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Page 13 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Criterion 31- Fracture Prevention of Reactor Coolant Pressure Boundary "Thereactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner, and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient thermal stresses, and (4) size of flaws."

The Braidwood Station pressurizer steam space connections are designed in accordance with the requirements of the ASME Code Section III with sufficient margin to the stresses encountered during operating, maintenance, testing, and postulated accident conditions. The pressurizer steam space connections, even the Alloy 82/182 welds, will continue to behave in a non-brittle manner. Ongoing BMV examinations of the pressurizer steam space connections at Braidwood Station will assure sufficient margin from rapidly propagating fracture until the susceptibility of Alloy 82/182 to PWSCC has been acceptably mitigated.

Criterion 32 - Inspection of Reactor Coolaht Pressure' Boundary "Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features, to assess their structural and leak tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel."

The Braidwood Station pressurizer steam space connections were designed to accommodate the visual, surface, and volumetric examinations of the ASME Code Section Xl. While the Alloy 82/182 to safe-end connections present limitations to current, fully qualified performance demonstration initiative (PDI) volumetric examination, ongoing BMV examinations will assure the structural and leak tight integrity of the pressurizer steam space connections at Braidwood Station.

Compliance with Operating Requirement: 10 CFR 50.36 - Plant Technical Specifications Braidwood Station Technical Specifications include requirements and associated action statements addressing reactor coolant pressure boundary (RCPB) leakage. The Braidwood Station Technical Specification limits for reactor coolant system operational leakage are one gallon per minute (gpm) for unidentified leakage, 10 gpm for identified leakage, and no pressure boundary leakage (reference Braidwood Station Technical Specifications Section 3.4.13, "RCS Operational Leakage"). Compliance with the zero non-isolable leakage criteria is met by conducting inspections and repairs in accordance with ASME B&PV Code, Section Xl, and 10 CFR 50.55a, "Codes and standards," as described below.

The unidentified leakage limit of one gpm is established as a quantity which can be accurately measured while sufficiently low to ensure early detection of leakage.

Page 14 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Leakage of this magnitude ban be reasonably detected within a short time, thus providing confidence that cracks associated with such leakage will not develop into a critical size before mitigating actions can be taken. If a through-wall boundary leak increases to the point where it is detected by the containment radiation monitor, mass balance calculations, or reactor containment sump level readings, the leak must be evaluated in accordance with the specified acceptance criteria and the plant must be shut down if the leak is determined to be a non-isolable reactor coolant system (RCS) pressure boundary fault.

In addition, Braidwood Station has implemented controls and expectations to address RCS leakage below Technical Specification limits. Exelon Generation Company (EGC) procedure ER-AP-331-1003, 'RCS Leakage Monitoring and Action Plan," has been implemented to assure adequate monitoring of RCS leakage and to provide minimum actions that could be taken at various RCS leakage levels.

Compliance with Inspection Requirements: 10 CFR 50.55a and the ASME Code Section Xl 10 CFR 50.55a, "Codes and standards," requires that inservice inspection and testing be performed in accordance with the requirements of the ASME Code, Section Xl, "Inservice Inspection of Nuclear Plant Components."Section XI contains applicable rules for examination, evaluation, and repair of code class components, including the RCPB.

However, as discussed above, Braidwood Station is using a risk-informed methodology forthe selection and examination of similar and dissimilar metal piping welds. While the Alloy 82/182 pressurizer steam space piping connections contain limitations in the performance of a fully qualified PDI volumetric examination, the current ISI program does not require that these welds be selected for volumetric examination. To compensate for the volumetric examination limitations, the Braidwood Station pressurizer steam space Alloy 82/182 connections will be visually examined each refueling outage until appropriate mitigation has been employed.

Compliance with Quality Assurance Requirements: 10 CFR 50, Appendix B Criterion V of Appendix B to 10 CFR 50 Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

The ASME Code Section Xl required visual examinations are performed using procedures that contain specific acceptance criteria or detailed recording criteria that are subsequently evaluated for acceptability. The visual examinations are performed using detailed instructions with a combination of qualitative and quantitative standards for the essential examination variables. Supplemental BMV examinations of the pressurizer Page 15of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 steam space Alloy 82/182 connections at Braidwood Station will be performed using standardized EGC procedures, which include appropriate acceptance criteria.

Criterion IX of Appendix B to 10 CFR 50 Criterion IX of Appendix B to 10 CFR 50 states that special processes, including nondestructive testing, shall be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

The pressurizer steam space connection BMV examinations at Braidwood Station will be performed by certified Level II or Level III examiners using EGC approved procedures with additional detailed instructions, as necessary.

Criterion XVI of Appendix B to 10 CFR 50 Criterion XVI of Appendix B to 10 CFR 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

For significant conditions adverse to quality, the measures taken shall include root cause determination and corrective action to preclude repetition of the adverse conditions.

The identification of an unacceptable visual indication requires repair, replacement or acceptance by analytical evaluation. In all cases, these indications would be tracked by tle EGC Corrective Action Program (CAP). In the case of a significant adverse condition, the CAP requires determination of the cause of the failure, evaluation of the extent of condition, and assignment of appropriate corrective actions to preclude recurrence. The EGC CAP meets the requirements of 10 CFR 50, Appendix B, Criterion XVI.

Requested Information I (c)

A description of the Alloy 82/182/600 pressurizer penetration and steam space piping connection inspection program that will be implemented at your plant during the next and subsequent refueling outages. The description should include the areas, penetrations and steam space piping connections to be inspected; the extent (percentage) of coverage to be achieved for each location; inspection methods to be used; qualification standards for the inspection methods and personnel; the process used to resolve any inspection indications; the inspection documentation to be generated; and the basis for concluding that your plant will satisfy applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizerpenetrations or steam space piping connections are found, indicate what followup NDE will be performed to characterize flaws in the leaking penetrations. Provide your plans for expansion of the scope of NDE to be performed if circumferential flaws are found in any portion of the leaking pressurizer penetrations or steam space piping connections.

Page 16of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2

Response

Braidwood Station will be performing a BMV examination of all affected pressurizer penetrations and steam space piping connections during the next refueling outages for each unit. The visual examinations will be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331-1001, "Boric Acid Corrosion Control (BACC) Inspection Locations, Implementation, and Inspection Guidelines,' and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation." In addition, Braidwood Station will continue to perform this BMV examination in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

Braidwood Station will use the guidance of ER-AP-331-1002 and LS-AA-125, 'Corrective Action Program (CAP) Procedure," to evaluate the source of any indications and to resolve inspection indications. Any evidence of pressure boundary leakage will require disposition under TS 3.4.13, "RCS Operational Leakage,! 10 CFR 50.72, 'Immediate notification requirements for operating nuclear power reactors," and the EGC Corrective Action Program.

The examinations will be documented in accordance with ER-AA-335-015 and ER-AP-331-1002 with written reports. All affected pressurizer penetrations and steam space piping connections have met and will continue to'meet all requirements related to the structural and leakage integrity of the RCPB. This is assured by compliance with the examination requirements of the ASME Code Section Xl and-the augmented examinations performed in accordance with this bulletin.

The basis for concluding that Braidwood Station satisfies applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

If a leaking penetration is found, a determination will be made, based on the location and nature of the indication, if additional NDE examinations will be performed or whether the location will be directly repaired by a weld overlay. In the evaluation required by the corrective action program, a determination will be made as to the extent of scope expansion and the type of NDE to be performed. All pressurizer upper head penetrations to steam space piping Alloy 82/182 weld connections are of such a configuration that a fully qualified PDI ultrasonic examination is not possible. Therefore, depending on the particulars of the indication, a best effort ultrasonic examination would be performed to characterize the flaw in the leaking penetration and to assess the condition of the other connections. Based on the results, and the quality of the examination technique, EGC may elect to preventively overlay some or all pressurizer upper head Alloy 82/182 locations.

Page 17of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Requested Information 1 (d)

In light of the information discussed in this bulletin and your understanding of the relevance of recent industry operating experience to your facility, explain why the inspection program identified in your response to item (1)(c) above is adequate for the purpose of maintaining the integrity of your facility's RCPB and for meeting all applicable regulatory requirements which pertain to your facility.

Response

As stated in response to question 1(c) above, Braidwood Station will be performing a BMV examination of all affected pressurizer penetrations and steam space piping connections during the next refueling outages for each unit. The visual examinations will be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331 -1001, "Boric Acid Corrosion Control (BACC)

Inrspection Locations, Implementation, and Inspection Guidelines," and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation."

In addition, Braidwood Station will continue to perform BMV examinations in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

The basis for concluding that the Braidwood Station BMV examination program meets all applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

Page 18 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Page 1 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Requested Information 1 (a)

A description of the pressurizer penetrations and steam space piping connections at your plant. At a minimum, this description should include materials of construction (e.g.,

stainless steel piping and/or weld metal, Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.), joint design (e.g., partial penetration welds, full penetration welds, bolted connections, etc.), and, in the case of weldedjoints, whether or not the weld was stress-relieved prior to being put into service. Additional information relevant with respect to determining the susceptibility of your plant's pressurizer penetrations and steam space piping connections to PWSCC should also be included.

Response

Byron Station Unit 1 and Unit 2 are 4-loop pressurized water reactors with the nuclear steam supply system designed by Westinghouse Electric Company, LLC (Westinghouse). The Byron Station units have one Westinghouse Model 84 pressurizer in each unit. Byron Station, Unit 1 began commercial operation on September 15, 1985.

Byron Station, Unit 2 began commercial operation on August 21, 1987.

The pressurizers were fabricated by Westinghouse at their Pensacola, Florida facility.

Each of the pressurizers has six piping connections: fiVe steam space connections and one surge line connection. The five steam space connections are: one spray line connection, one relief valve line connection, and three safety valve line connections.

The pressurizer manways (one per pressurizer) are not discussed in this response since, they do not contain any Alloy 600/82/182. Table'1 lists the pressurizer steam space connections and corresponding identifier.

Table 1 Byron Station Pressurizer Steam Space Connection Listing Unit Steam Space Connection Identifier 1 safety valve line connection (NOZZLE A-4C) (1RY01S) PN-04-F4 1 safety valve line connection (NOZZLE B-4B1) (RYO1S) PN-05-F5 1 safety valve line connection (NOZZLE C-4A) (1RY01S) PN-06-F6 1 l relief valve line connection (NOZZLE D) (1RY01 S) PN-03-F3 1 spray line connection (NOZZLE E) (1RY01 S) PN-02-F2 2 safety valve line connection (NOZZLE A-4C) (2RY01S) PN-03-F3 2 safety valve line connection (NOZZLE B-4B) (2RY01S) PN-04-F4 2 safety valve line connection (NOZZLE C-4A) (2RY01S) PN-05-F5 2 relief valve line connection (NOZZLE D) (2RY01S) PN-06-F6 2 spray line connection (NOZZLE E) (2RY01 S) PN-02-F2 All six of the pressurizer connections utilize an Alloy 82/182 weld connection from the low alloy steel pressurizer nozzles to the stainless steel safe-end attachments. These six locations are the only Alloy 600/82/182 based connections on the Byron Station pressurizers. The piping downstream of these connections do not have any Alloy 600/82/182 based components in any application. Also, the eight, 3/4 inch instrument line connections and the 3 inch sample line connection do not incorporate any Alloy Page 2 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 600/82/182 based components in their connection to the pressurizer or in any of their downstream connections. The Byron Station pressurizers have heater penetrations that use stainless steel sleeves connected to the pressurizer by stainless steel welds.

Pressurizer Steam Space Connections - Byron Station, Unit I Connection (1RY01S) PN-04-F4 (IRY01 S) PN-04-F4 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "C" pressurizer safety relief valve (I RY801 OC) to the pressurizer upper head nozzle (PN-04). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (1RYO1 S) PN-04-F4, identified in shop records as weld 4C on nozzle A, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were repairs to this weld during construction with a final acceptable radiography examination (RT) performed.

Fabrication records do not indicate size, depth or pertinent details of actual repairs performed.

Connection (1RY01 S) PN-05-F5 (1RY01 S) PN-05-F5 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "B" pressurizer safety relief valve (1RY801 OB) to the pressurizer upper head nozzle (PN-05). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-182, Grade F-316 flange. The flange provides the bolted Page 3 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (1RYO1 S) PN-05-F5, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were repairs to this weld during construction with final acceptable RT performed. Fabrication records do not indicate size, depth or pertinent details of actual repairs performed.

Connection (1RY01S) PN-06-F6 (1RY01 S) PN-06-F6 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "A"pressurizer safety relief valve (1RY801 OA) to the pressurizer upper head nozzle (PN-06). The welded. connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. D6wnstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (1RY01S) PN-06-F6, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were repairs to this weld during construction with final acceptable RT performed. Fabrication records do not indicate size, depth or pertinent details of actual repairs performed.

Connection (ORY01S) PN-02-F2 (lRYOIS) PN-02-F2 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (PN-02).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Page 4 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Upstream of the nozzle to safe-end weld is a weld that joins the four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the SA-403, Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455B and 1RY455C (SA-403, Grade WP304 valve bodies).

During fabrication, weld (1RY01 S) PN-02-F2, identified in shop records as weld PN F2 on nozzle E, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (1RY01S) PN-03-F3 (IRYO1S) PN-03-F3 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (1RY455A and 1RY456) to the pressurizer upper head nozzle. The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 821182 Weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304, 450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455A and 1RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld (IRY01S) PN-03-F3, identified in shop records as weld PN F3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Page 5 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Pressurizer Steam Space Connections - Byron Station. Unit 2 Connection (2RYOIS) PN-03-F3 (2RYO1S) PN-03-F3 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "C" pressurizer safety relief valve (2RY801 OC) to the pressurizer upper head nozzle (PN-03). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a 8A-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (2RY01S) PN-03-F3, identified in shop records as weld 4C on noizle A, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-04-F4 (2RY01S) PN-04-F4 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "B" pressurizer safety relief valve (2RY801 OB) to the pressurizer upper head nozzle (PN-04). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (2RY01S) PN-04-F4, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

Page 6 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units I and 2 A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-05-F5 (2RY01S) PN-05-F5 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the 'A"pressurizer safety relief valve (2RY801 OA) to the pressurizer upper head nozzle (PN-05). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade' F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (2RY01 S) PN-05-F5, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle-to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-02-F2 (2RY01S) PN-02-F2 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (PN-06).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Upstream of the nozzle to safe-end weld is a weld that joins the four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel Page 7 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 weld that joins the SA-1 82, Grade F-316L safe-end material to the SA-403,' Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455B and 2RY455C (SA-403, Grade WP304 valve bodies).

During fabrication, weld (2RY01S) PN-02-F2 identified in shop records as weld 2 on nozzle E, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-06-F6 (2RY01 S) PN-06-F6 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (2RY455A and 2RY456) to the pressurizer upper head nozzle. The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304, 450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455A and 2RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld (2RY01S) PN-06-F6, identified in shop records as weld 3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Page8of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Requested Information I (b)

A description of the inspection program for Alloy 82/182/600 pressurizer penetrations and steam space piping connections that has been implement~d at your plant. The description should include when the inspections were performed; the areas, penetrations and steam space piping connections inspected; the extent (percentage) of coverage achieved for each location which was inspected; the inspection methods used; the process used to resolve any inspection findings; the quality of the documentation of the inspections (e.g., written report, video record, photographs);, and, the basis for concluding that your plant satisfies applicable regulatory requirements related to the integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections were found, indicate what followup NDE was performed to characterize flaws in the leaking penetrations.

Response

The Alloy 82/182/600 pressurizer penetration and steam space connections at Byron Station are limited to the ten (five per unit) low alloy steel nozzle to stainless steel, Alloy 82/182, welded safe-end connections. All the pressurizer welds subject to the actions of this bulletin (identified in Table 2) have had volumetric'and surface examinations performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code Category B-F, 5 Code Item number B5.40. 6 The first inservice inspection (ISI) Interval examinations were performed in accordance with the 1983 Edition, Summer 1983 addenda, of the ASME Code Section Xl. Examinations performed in the second ISI interval were performed in accordance with the 1989 Edition, no addenda, of the ASME Code Section Xl. The examination areas for these welds were those identified in the ASME Code Section Xl, Figure IWB-2500-8, Similar and Dissimilar Metal Welds in Components and Piping." This examination area is the same for the ASME Code Section Xl Editions 1983 through 1989.

Byron Station Unit 1 and Byron Station Unit 2 are in the 2nd Interval with the ISI program, as stated above, based on the 1989 Edition of the ASME Code Section XI. On February 20, 2001 the NRC approved the'use of a risk informed methodology for the selection and examination of ASME Code Class 1 and Code Class 2 piping welds; however, none of the subject pressurizer safe-end welds have been re-examined under the risk informed ISI program. In addition, there have been no volumetric re-examinations performed on the pressurizer welds subject to the actions of this bulletin since the implementation of Supplement 10 7 to Appendix Vil 8of the ASME Code Section Xl since its implementation in November of 2002.

5 "Pressure Retaining Dissimilar Metal Welds" 6 Pressurizer, Nozzle-to-Safe-end Butt Welds, Nominal Pipe Size 4 Inch or Larger' "Qualification Requirements for Dissimilar Metal Piping Welds" 8 Performance Demonstration for Ultrasonic Examination Systems" Page 9 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 All examinations, surface and volumetric, were recorded on hard copy data sheets. No video or photographs were used to supplement these examinations. There have been no recordable indications on these ten welds that required disposition. All surface examinations covered 100% of the required examination area. The ultrasonic examinations, although partially obstructed in some cases, achieved greater than 90% of the required examination volume coverage.

In addition to the nondestructive examinations listed in Table 2, all steam space pressurizer Alloy 82/182 welds were visually examined each refueling outage, at a minimum, in accordance with the pressure test requirements of the ASME Code Section Xl, Category B-P. Also, Byron Station performed a 100% bare metal visual (BMV) examination of all affected Unit 2 pressurizer penetrations and steam space piping connections during the Spring 2004 refueling outage. This examination was performed in accordance with the industry's Materials Reliability Program (MRP) recommendations as described in a Leslie Hartz, Chair, MRP Senior Representatives, letter to PWR licensees dated January 20, 2004.

There has not been any indication of leaking pressurizer penetrations or steam space piping connections at Byron Station and therefore, there has not been a need for dispositions, additional nondestructive examinations (NDE), examination expansions, or flaw characterizations.

Page 10 of 18

PD Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units I and 2 Table 2 Byron Station Alloy 82/182 Pressurizer Welds Examination History Weld Exam Date Exam Method Exam Technique Results Disposition/

Identifier . . Data Sheet March 1987 Surface (S) Dye Penetrant (PT) NRI NPA (1RY01S) PN-02-F2 October 450 longitudinal wave transducer - N/A 198 Volumetric; (Vol) (40Lwv)NRI 88BYI-UT-084 ahd 1988(45 L-wve)88BYI-UT-086 Acceptable January Vol . 450 L-wave 3600 beam redirection Indication 90BY1-UT-Ill and 1990 N/A S PT NRI 90BY1-PT-020 (1RY0lS) PN-03-F3 450 shear wave transducer Acceptable November Vol (450 S-wave) and Laminations and ID geometry 97BY1-UTD-080 and 1997 450 L-Wave 97BYl-UTD-083 S *PT NRI NIA 97BY1-PT-005 Vol 450 and 60 0 L-wave NRI N/A (1RY01S) PN-04-F4 April 1996 96BY-UTD-143

. S PT-*NRI PT -NRI -N/A 96BY1-PT-016 Acceptable Vol 450 L-Wave 3600 beam redirection Indication 90BY1-UT-084 and January _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _9 B I U- 1 1990 NM s PT NRI 90BY1-PT-20 (1RY01S) PN-05-F5 Acceptable 1997 Vol 450 S-Wave and 450 L-Wave ID geometry 97BY1-UTD-081 and Novmbe 97BYI-UTD-084 1997NRIN/A S PT NRI 97BY1-PT-008 March 1987 5 PT NRI PZR-PT-73 OctoberN/A (ORY01S) PN-06-F6 1988 Vol 450 L-Wave NRI 88BY1-UT-083 and (IRYlS) N-0-F6 98888BYI-UT-085 Acceptable November Vol 450 S-Wave and 450 L-Wave ID geometry 97BY1-UTD-082 and 1997 975LYI-epTD-085 (R 1)PN2-2

__________PN_02 _F2 October 1990 _

Vol__ __ _ I___ _ _ _

550 L-wave

_ _ _ _ _ _ _ _ I__ _ _ _

ID_geometry

_ __ _ _ _ _ _ _ I__

Acceptable 90BY2-UT-084 Page 11 of 18

Attachment 2 . ,

Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Table 2 Byron Station Alloy 82/182 Pressurizer Welds Examination History (continued)

Weld Exam Date Exam Method Exam Technique Results Disposition/

Identifler .

  • Data Sheet (2RY01S) PN-02-F2 N/A September S PT NRI 90BY2-PT-044 1990 SP R March 1995 Vol 700 and 450 L-Wave NRI N/A

_____ ____95BY2-UTD-066, February S PT NRI N5Y2PTA 6 (2RY01S) PN-03-F3 1995 95BY2-PT-062 Vol 450 and 600 L-Wave NRI B2R09-UT-053 April 2001 N/A9UT05 S PNRI B2R09-PT-018 March 1995 Vol 700 and 450 L-Wave NRI 95BY2-UTD-066 February (ROS954- 5 PTTNIN/A . NRI 95BY2-PT-078 October Vol OctoberN/A700 and 450 L-Wave NRI 93BY2-UT-123 and 19 93BY2-UT-124 S~eptember PT NRI N/A (2RY01S) PN-05-F5 1993 S PT NRI 93BY2-PT-032 Vol 450 and 600 L-Wave NRI N/A AprIl 2001 132R09-UT-070-N/A S PT NRI 32R09-PT-018 October Vol 550 L-Wave ID geometry Acceptable 1990 S TNIN/A (2RY01S) PN-06-F6 S PT NRI 90BY2-PT-048 April 2001 Vol 450 and 600 L-Wave ID geometry B2R09-UTal1

. . S PT NRI B2R09-PT-019 Page 12 of 18

Y Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Basis for Concluding Regulatory Requirements are Satisfied As stated above, the Byron Station pressurizer connections affected by this Bulletin are limited to the ten (five per unit), Alloy 82/182 full penetration nozzle to safe-end welds.

The completion of volumetric, surface, and visual examinations without any evidence of recordable, relevant indications, through-wall leakage or any recordable wastage of the low alloy steel surface, is assurance of the previous integrity of the Alloy 82/182 connections.

Ongoing integrity of the Byron Station pressurizer steam space Alloy 82/182 connections is assured by performing BMV examinations, at a minimum, each refueling outage (approximately 18 months). The examination will be performed until mitigation (i.e., weld overlays of the Alloy 82/182 locations) is performed on all pressurizer steam space connections.

The specific regulatory requirements are listed below with the associated response addressing how the requirement is met.

Compliance with Design Requirements: 10 CFR 50, Appendix A - General Desiqn Criteria (GDC)

Criterion 14 - Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely 16w probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

The Byron Station pressurizer steam space connections are designed, fabricated, tested, and examined in accordance with the requirements of the ASME Code Section l1l, "Requirements for Design and Manufacture of Nuclear Power Plant Components,"

and the ASME Code Section Xl. In general, the controls established by these construction and inspection codes assures that the reactor coolant pressure boundary maintains an extremely low probability of rapidly propagating failure and gross rupture.

The BMV examination technique to be used in the Byron Station, Unit 1 Spring 2005 refueling outage, and in the Unit 2 Fall 2005 refueling outage, is a reliable means for identifying the very low leakage rates potentially associated with Alloy 82/182 cracking.

Therefore, based on the design, materials, and examination methods, the Byron Station pressurizers continue to comply with the requirements of GDC 14.

In addition, in the case of the pressurizer steam space Alloy 82/182 locations, modifications in the form of weld overlays using primary water stress corrosion cracking (PWSCC) resistant material will be performed to provide added assurance that these connections will have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Page 13 of 18

-Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient thermal stresses, and (4) size of flaws."

The Byron Station pressurizer steam space connections are designed in accordance with the requirements of the ASME Code Section III with sufficient margin to the stresses encountered during operating, maintenance, testing, and postulated accident conditions.

The pressurizer steam space connections, even the Alloy 82/182 welds, will continue to behave in a non-brittle manner. Ongoing BMV examinations of the pressurizer steam space connections at Byron Station will assure sufficient margin from rapidly propagating fracture until the susceptibility of Alloy 82/182 to PWSCC has been acceptably mitigated.

Criterion 32- Inspection of Reactor Coolant Pressure Boundary

."Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features

, to assess their structural and leak tight integrity, and (2) an appropriate material

. surveillance program for the reactor pressure vessel."

The Byron Station pressurizer steam space connections were designed to accommodate the visual, surface, and volumetric examinations of the ASME Code Section Xl. While the Alloy 82/182 to safe-end connections present limitations to current, fully qualified performance demonstration initiative (PDI) volumetric examination, ongoing BMV examinations will assure the structural and leak tight integrity of the pressurizer steam space connections at Byron Station.

Compliance with Operating Requirement: 10 CFR 50.36 - Plant Technical Specifications Byron Station Technical Specifications include requirements and associated action statements addressing reactor coolant pressure boundary (RCPB) leakage. The Byron Station Technical Specification limits for reactor coolant system operational leakage are one gallon per minute (gpm) for unidentified leakage, 10 gpm for identified leakage, and no pressure boundary leakage (reference Byron Station Technical Specifications Section 3.4.13, "RCS Operational Leakage"). Compliance with the zero non-isolable leakage criteria is met by conducting inspections and repairs in accordance with ASME B&PV Code, Section Xl, and 10 CFR 50.55a, "Codes and standards," as described below.

The unidentified leakage limit of one gpm is established as a quantity which can be accurately measured while sufficiently low to ensure early detection of leakage.

Leakage of this magnitude can be reasonably detected within a short time, thus providing confidence that cracks associated with such leakage will not develop into a Page 14 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 critical size before mitigating actions can be taken. If a through-wall boundary leak increases to the point where it is detected by the containment radiation monitor, mass balance calculations, or reactor containment sump level readings, the leak must be evaluated in accordance with the specified acceptance criteria and the plant must be shut down if the leak is determined to be a non-isolable reactor coolant system (RCS) pressure boundary fault.

In addition, Byron Station has implemented controls and expectations to address RCS leakage below Technical Specification limits. Exelon Generation Company (EGC) procedure ER-AP-331-1003, ."RCS Leakage Monitoring and Action Plan," has been implemented to assure adequate monitoring of RCS leakage and to provide minimum actions that could be taken at various RCS leakage levels.

Compliance with Inspection Requirements: 10 CFR 50.55a and the ASME Code Section Xl 10 CFR 50.55a, "Codes and standards," requires that inservice inspection and testing be performed in accordance with the requirements of the ASME B&PV Code, Section Xl, "Inservice Inspection of Nuclear Plant Components." Section Xl contains applicable rules for examination, evaluation, and repair of code class components, including the RCPB.

However, as discussed above, Byron Station is using a risk-informed methodology for the selection and examination of similar and dissimilar metal piping welds. While the Alloy 82/182 pressurizer steam space piping connections contain limitations in the performance of a'fully qualified PDI volumetric examination, the current ISI program does not require that these welds be selected for volumetric examination. To compensate for the volumetric examination limitations, the Byron Station pressurizer steam space Alloy 82/182 connections will be visually examined each refueling outage until appropriate mitigation has been employed.

Compliance with Quality Assurance Requirements: 10 CFR 50. Appendix B Criterion V of Appendix B to 10 CFR 50 Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

The ASME Code Section Xl required visual examinations are performed using procedures that contain specific acceptance criteria or detailed recording criteria that are subsequently evaluated for acceptability. The visual examinations are performed using detailed instructions with a combination of qualitative and quantitative standards for the essential examination variables. Supplemental BMV examinations of the pressurizer steam space Alloy 82/182 connections at Byron Station will be performed using standardized EGC procedures, which include appropriate acceptance criteria.

Page 15 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Criterion IX of Appendix B to 10 CFR 50 Criterion IX of Appendix B to 10 CFR 50 states that special processes, including nondestructive testing, shall be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements:

The pressurizer steam space connection BMV examinations at Byron Station will be performed by certified Level II or Level IlIl examiners using EGC approved procedures with additional detailed instructions, as necessary.

Criterion XVI of Appendix B to 10 CFR 50 Criterion XVI of Appendix B to 10 CFR 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

For significant conditions adverse to quality, the measures taken shall include root cause determination and corrective action to preclude repetition of the adv6rse conditions.

The identification of an unacceptable visual indication requires repair, replacement or acceptance by analytical evaluation. In all cases, these indications would be tracked by the EGC Corrective Action Program (CAP). In the case of a significant adverse condition, the CAP requires determination of the cause of the failure, evaluation of the extent of condition, and assignment of appropriate corrective actions to preclude recurrence. The EGC CAP meets the requirements of 10 CFR 50, Appendix B, Criterion Xv~l.

Requested Information I (c)

A description of the Alloy 82/182/600 pressurizer penetration and steam space piping connection inspection program that will be implemented at your plant during the next and subsequent refueling outages. The description should include the areas, penetrations and steam space piping connections to be inspected; the extent (percentage) of coverage to be achieved for each location; inspection methods to be used; qualification standards for the inspection methods and personnel; the process used to resolve any inspection indications; the inspection documentation to be generated; and the basis for concluding that your plant will satisfy applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections are found, indicate what followup NDE will be performed to characterize flaws in the leaking penetrations. Provide your plans for expansion of the scope of NDE to be performed if circumferential flaws are found in any portion of the leaking pressurizer penetrations or steam space piping connections.

Response

As stated in the response to question 1(b), Byron Station performed a 100% BMV examination of the Unit 2 Alloy 82/182 pressurizer steam space connections during the recent Spring 2004 refueling outage. Byron Station will continue to perform a 100%

Page 16 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 BMV examination of all affected pressurizer penetrations and piping connections during the next refueling outages for each unit. The visual examinations will continue to be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331-1001, "Boric Acid Corrqsion Control (BACC)

Inspection Locations, Implementation, and Inspection Guidelines," and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation." In addition, Byron Station will continue to perform this BMV examination in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

Byron Station will use the guidance of ER-AP-331-1002 and LS-AA-125, 'Corrective Action Program (CAP) Procedure," to evaluate the source of any indications and to resolve inspection indications. Any evidence of pressure boundary leakage will require disposition under TS 3.4.13, 'RCS Operati6nal Leakage," 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors," and the EGC Corrective Action Program.

The examinations will be documented in accordance with ER-AA-335-015 and ER-AP-331-1002 with written reports. All affected pressurizer penetrations and steam space piping connections have met and will continue to meet all requirements related to the structural and leakage integrity of the RCPB. This is assured by compliance with the examination requirements of the ASME Code Section Xl and the augmented examinations performed in accordancewith this bulletin.

The basis for concluding that Byron Station satisfies applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

If a leaking penetration is found, a determination will be made, based on the location and nature of the indication, if additional NDE examinations will be performed or whether the location will be directly repaired by a weld overlay. In the evaluation required by the corrective action program, a determination will be made as to the extent of scope expansion and the type of NDE to be performed. All pressurizer upper head penetrations to steam space piping Alloy 82/182 connections are of such a configuration that a fully qualified PDI ultrasonic examination is not possible. Therefore, depending on the particulars of the indication, a best effort ultrasonic examination would be performed to characterize the flaw in the leaking penetration and to assess the condition of the other connections. Based on the results, and the quality of the examination technique, EGC may elect to preventively overlay some or all pressurizer upper head Alloy 82/182 connections.

Page 17 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Requested Information 1 (d)

In light of the information discussed in this bulletin and your understanding of the relevance of recent industry operating experience to your facility, explain why the inspection program identified in your response to item (1)(c) above is adequate for the purpose of maintaining the integrity of your facility's RCPB and for meeting all applicable regulatory requirements which pertain to your facility.

Response

As stated in response to question 1(c) above, Byron Station Unit 2 performed a BMV examination of all affected pressurizer penetrations and steam space piping connections during the Spring 2004 refueling outage. There were no indications of leakage or boric acid deposits identified during this examination.

Byron Station will be performing a BMV examination of all affected pressurizer penetrations and steam space piping connections during the next refueling outages for each unit. The visual examinations will be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331-1001, "Boric Acid Corrosion Control (BACC) Inspection Locations, Implementation, and Inspection Guidelines," and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation."

In addition, Byron Station will continue to perform BMV examinations in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

The basis for concluding that the Byron Station BMV examination program meets all applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

Page 18 of 18

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Page 1 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Requested Information 1 (a)

A description of the pressurizer penetrations and steam space piping connections at your plant.

At a minimum, this description should include materials of construction (e.g., stainless steel piping and/or weld metal, Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.),

joint design (e.g., partial penetration welds, full penetration welds, bolted connections, etc.),

and, in the case of weldedjoints, whether or not the weld was stress-relieved prior to being put into service. Additional information relevant with respect to determining the susceptibility of your plant's pressurizer penetrations and steam space piping connections to PWSCC should also be included.

Response

Three Mile Island Nuclear Station (TMI), Unit No.1 is a single-loop pressurized water reactor with the nuclear steam supply system designed by Babcox and Willcox Company (B&W). TMI, Unit I has one B&W 177-FA pressurizer (Figure 1). TMI, Unit 1 began commercial operation on September 2, 1974.

The pressurizer was fabricated by B&W. The pressurizer has eight (8) water space connections, eight (8) steam space connections, one (1) surge line connection, and one (1)

Manway. The eight (8) steam space connections are: one (1) spray line connection; one (1) relief valve line connection; two (2) safety valve line connections; three (3) level sensing connections, and; one (1) vent connection. A description of pressurizer connections materials utilized is contained herein.

Pressurizer Connections - TMI. Unit 1 The pressurizer vessel contains the following 18 penetrations.

Penetration Type Number 1-inch vent nozzle 1 212-inch pressure relief nozzles 3 4-inch spray nozzle 1 I 1/-inch thermowell 1 1-inch level sensing nozzles 6 1-inch sampling nozzle 1 19-inch heater bundle openings 3 10-inch surge nozzle (outside the scope of this document, not steam space) 1 16-inch manway (not discussed in this document as it contains no Alloy 600/182/82)

The pressurizer steam space connections on the upper head include a 16-inch manway, a 1-inch vent and sampling nozzle, three 21/-inch relief nozzles, and a 4-inch spray nozzle. The spray line and attached spray head are connected to the inside end of the spray nozzle and suspended from the upper head, as shown in Figure 2A.

Page 2 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 The following is a discussion of'the penetrations identified above (except for the surge line and manway).

1. Vent Nozzle A 1-inch, Schedule 160 vent nozzle is located at the top center of the upper head to allow complete venting and to permit sampling from the steam space. The vent nozzle was fabricated from Alloy 600 bar (SB-166). The vent nozzle is joined to the interior of the upper head with a partial penetration weld (also called J-groove weld) as illustrated in Figure 2B. Based on information from the vendor, the J-groove weld is buttered at TMI, Unit 1. In addition to the J-groove weld, the vent nozzle installed at TMI, Unit 1 is a two-piece construction as illustrated in Figure 2B. The two Alloy 600 pieces (the top one is called a safe-end) is jointed, without weld butter, by a full penetration V-groove weld with Alloy 82/182. In addition to the J-groove weld, TMI, Unit 1 has an Alloy 82/182 weld boss welded to the pressurizer upper head outside diameter (O.D.) surface as illustrated in Figure 2B. The weld boss is not in contact with the reactor coolant (RC) water. The weld boss was applied during the pressurizer vessel fabrication as a contingency in the event a weld repair is needed.

Table 1 - Originally Installed 1-inch Vent Nozzle Assembly (The pressurizer has 1 vent nozzle, see Figure 2A and 2B)

Connection - a-Material 1-inch vent nozzle, Alloy 600 MK-78 Vent nozzle to upper Alloy 82 or head J-groove weld, 182 WP-95 Butter for J-groove Alloy 82 or weld, WP-94 182 Vent nozzle safe- Alloy 600 end, MK-77 Vent nozzle to safe- Alloy 82 or end weld, WP-92 182 Vent nozzle weld Alloy 82 or boss (not wetted by 182 RC)

Safe-end to stainless Alloy 82 or steel component 182 weld18 Page 3 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

2. 2%-inch Pressure Relief Nozzles Three 21/2-inch pressure relief nozzles are located near the top of the vessel for attachment of pressure relief devices. The nozzles are manufactured from carbon steel with a stainless steel cladding welded on the inner diameter (I.D.). The nozzles are joined to the upper head with a full penetration weld (carbon steel weld) as illustrated in Figure 3.

A type 316 stainless steel safe-end (also called long weld necks) are welded to the top of the nozzles with full penetration V-groove with Alloy 182 as illustrated in Figure 3. Based on information from the vendor, the nozzles were buttered with Alloy 82/182 before the V-groove welding. The flanges on the long weld necks contain eight equally spaced bolt holes for attaching the pressure relief valves or power operated relief valve block valve to the nozzle.

The materials used for the pressure relieve nozzle assembly are summarized in Table 2.

Table 2 - Originally Installed 21/2-inch Pressure Relief Nozzle Assembly.

(The'pressurizer has 3 pressure relief nozzles, see Figure 3)

-Connection Material 21/z-inch pressure. SS Clad relief nozzle carbon steel Safe-end (also called Type 316, long weld neck) SA-182 Pressure relief nozzle to safe-end weld, Alloy 182/82 WP-91 (a)

Pressure Relief Alloy 182 Nozzle Butter, WP-73 (a) For WP-91, the tack and root layer was Alloy 82 and the balance of the weld and weld repair was Alloy 182.

3. 4-inch Spray Nozzle The 4-inch, Schedule 120 spray nozzle is located on the upper head of the pressurizer vessel, connecting the external 2Y.-inch stainless steel spray line from the discharge of a reactor coolant pump with the internal stainless steel spray line and spray head. The nozzle is mounted normal to the pressurizer upper head, entering at a 450 angle from the horizontal. The nozzle body is carbon steel with stainless steel weld clad at the l.D. surface. The nozzle body is joined to the upper head with a full penetration weld (i.e., carbon steel weld).

Alloy 600 transition pieces are welded on both ends of the nozzle, the 8-inch long safe-end connecting with the stainless steel external spray line, and the 5-inch long extension pin (i.e.,

Page 4 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 pipe section) that connects to the internal stainless steel spray line as illustrated in Figure 4. At the top, the Alloy 600 safe-end is attached to the 4-inch spray nozzle by a full penetration Alloy 82 V-groove weld. The Alloy 600 extension pin is attached to the stainless steel weld build-up on the inside of the carbon steel shell by a full penetration Alloy 82/182 V-groove weld without butter. The extension pin is attached to the internal stainless'steelIspray pipe by a full penetration Alloy 82 V-groove weld without butter.

A stainless steel thermal sleeve is installed inside the nozzle assembly to provide protection from thermal stresses. Based on information from the vendor, as illustrated in Figure 5, there are four Alloy 82/182-weld buttons on the inside surface just above the thermal sleeve and four Alloy 82/182 weld pads just below the thermal sleeve. The buttons and pads minimize the chances of the thermal sleeve becoming a loose part. The materials used for the 4-inch spray, nozzle assembly at each unit are summarized in Table 3.

Table 3 - Originally Installed 4-inch Spray Nozzle Assembly (The pressurizer has I spray nozzle, see Figure 4 and Figure 5)

Connection. Material 4-inch spray nozzle, SS clad carbon steel Safe-end, MK-45 Alloy 600 Safe-end to spray Alloy 82 nozzle weld, WP-45 Extension pin, MK-46 Alloy 600 Extension pin to spray nozzle weld, Alloy 82 WP-46 Extension pin to Alloy 82 or internal spray pipe, 182 WP-104 4 upper weld buttons, Alloy 82 or WP-103 182 4 lower weld pads, Alloy 82 or WP-79 182 Safe-end to stainless Alloy 82 or steel component 182 weld Page 5 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

4. 1%-inch Thermowell The 1%/-inch thermowell is a closed penetration located in the side of the pressurizer shell about 9 feet from the bottom of the vessel. The nozzle is located just above the upper heater bundle, extending 4 inches into the water space. The thermowell is made of Alloy 600. The outer end of the thermowell is threaded to hold the RTE sensor. The thermowell is welded to the interior of the vessel wall with an Alloy 82/182 partial penetration (J-groove) weld as illustrated in Figure
6. Alloy 82/182 butter was applied. The weld was generally performed similar to the 1-inch vent J-groove weld (See Figure 2B and Figure 6). In addition to the J-groove weld, TMI, Unit 1 had Alloy 82/182 weld boss welded to the pressure vessel O.D. surface (not wetted, see Figure 6).

The weld boss was applied during the fabrication to facilitate future weld repair if needed. The materials used for the 11/2-inch thermowell assembly at each unit are summarized in Table 4.

Table 4 - Originally Installed 11%-inch Thermowell Nozzle Assembly (The pressurizer has 1 thermowell nozzle, see Figure 6)

Connection . Material 1%2-inch thermowell, Alloy 600 SB-166 Pressurizer butter for Alloy 82 or J-grpove weld, WP- 182 80 Thermowell to pressurizer (I.D. side) Alloy 182 J-groove weld, WP-81 Weld boss on pressurizer (O.D.) Alloy 82 or WP-107 (not wetted 182 by RC)

5. Level Sensing Nozzles The six 1-inch, Schedule 160 level sensing nozzles (also called level taps) are located in the pressurizer vessel consisting of two pairs of three nozzles each at high and low elevations. The nozzle bodies are of carbon steel with stainless steel weld clad on the l.D. surface. Each level sensing nozzle is joined to the carbon steel pressurizer shell with a full penetration carbon steel weld. At the top, as illustrated in Figure 7, an Alloy 600 safe-end is attached to each level-sensing nozzle with a full penetration Alloy 82 V-groove weld without butter. The materials used for the level sensing nozzle assembly are summarized in Table 5.

Page 6 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

6. 1-inch Sampling Nozzle The pressurizer has one 1-inch Schedule 160 sampling nozzle (also called sampling tap) connecting to a sampling line. The sampling nozzle is located at the same elevation as the 1A-inch thermowell described in Section 4. The sampling nozzle is mbunted similar to the level sensing nozzles (described in Section 5 and Figure 7). The materials used for the 1-inch level sampling nozzle assembly are summarized in Table 5.

Table 5 - Originally Installed Level Sensing Nozzle and Sampling Nozzle Assembly (The pressurizer has 6 level sensing nozzles and 1 sampling nozzle, see Figure 7)

Cnne'&tion' -'Mate'ral 1-inch level sensing nozzle and SS clad 1-inch sampling nozzle, carbon steel Level sensing nozzle and Alloy 600 sampling nozzle safe-end, Safe-end to level sensing nozzle and sampling nozzley weld, WP-63

7. Pressurizer Heater Bundle The heater belt forgings have three openings to accommodate the heater bundle assemblies.

Each heater bundle is an assembly consisting of three parallel disks drilled to hold 39 individual immersion heaters as shown in Figure 9. The outermost of the three disks, the heater bundle diaphragm plate, mates with and forms a seating surface with the heater belt forging penetration. The other two discs, called support plates, are fabricated from stainless steel.

Figure 10 gives a cross-sectional view of the pressurizer heater bundle assembly pressure retaining items. The heater bundle cover plates are bolted on the outer surface of the diaphragm plates, holding the diaphragm plates against the mating surface and providing support for the heater bundle assembly. As illustrated in Figure 11, a seal weld provides the pressure boundary between the heater bundle diaphragm plate and the heater belt forging. No structural credit is given for the seal weld.

The original heater bundles contain Alloy 600/182/82. The diaphragm plate is fabricated from Alloy 600 plate (SB-168). In addition, there are short heater sleeves machined from Alloy 600 bar and are attached to the inner side of the Alloy 600 diaphragm plate by an Alloy 82 partial penetration groove weld without buttering as illustrated in Figure 12. Each Type 316L stainless steel heater sheath passes through the diaphragm plate and sleeve and is attached to Alloy 600 sleeve by an Alloy 82 fillet weld shown in Figure 12. The heater bundles materials discussed in this section for each unit are summarized in Table 6.

Page 7 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 2003 Replacement Heat Bundle The original lower heater bundle was found leaking and was replaced in 2003. The replacement heater bundle has 12 larger diameter (1.25-inch) higher power heaters in lieu of the 39 heaters (0.66-inch) in the original heater bundle. The replacement diaphragm plate is fabricated from Type 304 stainless steel, instead of Alloy 600. Due to the original seal weld (original Alloy 600 diaphragm plate to the stainless steel cladding weld) being performed with Alloy 82 and subsequent weld repairs being performed with Alloy 152/52, the new seal weld for the replacement heater bundle diaphragm plate used Alloy 152/52. The new seal weld is illustrated in Figure 11.

Table 6 - Originally Installed Heater Bundle Nozzle Assembly (The pressurizer has 3 heater bundles)

Alloy 600 Diaphgm -pType! 2 Mat-::

Heater Bunde C ompbnents-' .

Diaphragm plate, Alloy 600, SB-168 (1 per heater bUndle)

Heater sheath,Tye36,S21 (39 per heater bundle) Type 316L SA-213 Heater sleeve, Alloy 600, SB-166 (39 pler heater bundle)

Diaphragm plate seal (Field) weld Alloy 82 or 182 (1 per heater bundle)

Heater sleeve to diaphragm Alloy 82 plate weld, WP-119 (39 per heater bundle)

Heater sheath to diaphragm Alloy 82 plate weld, WP-120 (39 per heater bundle)

Castellated seal nuts to diaphragm plate tack weld, WP- Alloy 82 126 Page 8 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Table 7 - Replacement Heater Bundle Nozzle Assembly

'Stainless'Steel' Diaphragm TypeHeater- Matral b'-

Bundle Diaphragm plate, (3 per pressurizer) Stainless Steel Heater sheath, (12 Per heater bundle) Stainless Steel Diaphragm plate seal (Field) Weld (1 per Alloy 52/152 heater bundle)

Heater sheath to diaphragm plate weld, Stainless Steel WP-121 (WP-123 for Weld DB)

(a) I he two original bundles Tor the original lower neater bundle.

(b) The replacement heater bundle for the original lower heater bundle

8. Post-Weld Stress Relief Heat Treatment The post-weld stress relieve heat treatment (PWHT)'of pressurizer connections were typically performed at 1100-11 50 0F. Based on information from the vendor, the Alloy 600 heater bundle diaphragm plates and heater sleeves did not receive any PWHT: In addition, none of the field welds performed either at the time of the driginal pressurizer installation or during field repair and modification would have received any PWHT. This includes the following known replacement or field repairs:

The heater bundle diaphragm plate to shell cladding seal weld (including the replacement heater bundle in 2003).

. All welds connecting the nozzles (or safe-ends) to external piping.

Page 9 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 VENT AND SAMPLING NOZZLE SPRAY NOZZLE

.4, --

HEAD -

" SPRAY UNE AND SUPPORTS

% INTERNAL SUPPORTS LEVEL SENSING NOZZLE

> SHELL a SUPPORT PLATE ASSEMBLY

. ,, I I THERMOWELL - HEATER BELT LEVEL SENSING NOZZLE SURGE DIFFUSER

- SURGE NOZZLE Figure 1 - The general arrangement of pressurizer penetrations at B&W 177-FA units. TMI, Unit I also has 36 Internal stainless steel ladder rungs welded to the I.D. surface of the pressurizer vessel.

Page 10 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 TOP VIEW UFTING LUGS PRESSURE RELIEF NOZZLES NOZZLE MANAY

> VENT AND SAMPLING NOZZLE PRSSR RELIE SIDE VIEW

.- 1E PRESSURE RELIEF I NOZZLE (1 OF 3)

MANWAY FORGING w dSPRAY NOZZLE SPRAY UNE WeldBi TMI-1 Figure 2A - Penetrations at the upper head of pressurizer vessel. Note, the manway boss was weld buildup at TMI-1, instead of a forging.

Page 11 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Final Bore Alloy 600 Safe End, MK-77 Alloy 600 Vent and Sampling Nozzle,'

I MK-78 Carbon Allt

'Steel Upper No; Head, MK-5 OD surface of

>X the pressurizer upper head Figure 2B - Top, 1-inch Schedule 160 Alloy 600 vent to Alloy 600 safe-end weld. Bottom, detail of the 1-inch vent and sampling nozzle to pressurizer upper head J-groove weld and weld boss (not wetted by RC).

Page 12 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Integral flange and bolt holes I Stainless steel safe end (also called long weld neck) MK-32 PT, WP-73 I

Carbon steel pressure relief nozzle, MK-31 Carbon Steel A'- I I Pressurizer Upper Head I-Stainless Steel Cladding ' _

Figure 3 - Detail of the three 2/2-Inch pressure relief nozzle to stainless steel safe-end weld.

Page 13 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 WP-102 Stainless Steel Thermal Sleeve Carbon steel spray nozzle, MK-9 I

. I. .*

Stainless Steel Cladding Alloy 600 Extension Pin (MK-46)

Alloy 600 Extension Pin Stainless 1/8" dia.

7- 'Steel Interal Spray Pipe electrode Figure 4 - Top, detail of the 4-inch spray nozzle assembly. Bottom, detail of the extension pin to the internal spray pipe weld. See Figure 5 for detail of the weld buttons for positioning the thermal sleeve.

Page 14 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

'WP-103 Upper Weld Buttons Lay the four %" dia.weld buttons @900 after the 4 lower weld pads and the stainless steel sleeve are in place.

(The buttons are on the Alloy 600 safe end).

Stainless Steel WP-79 ?-  %'

4 Weld Pads @900

/

Lay the lower weld pads after inserting the stainless steel thermal sleeve (The lower weld pads are on the Alloy 600 Extension Pin)

Figure 5 - Detail of the weld buttons for positioning the stainless steel thermal sleeve inside the 4-inch spray nozzle.

Page 15 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Outside the pressurizer Alloy 600 Thermowell (MK-29) vessel Inside the I pressurizer vessel Carbon steel pressurizer shell Stainless steel cladding Alloy 60 Thermowell

{/ D MK-29 O.D. surface of the pressurizer shell Figure 6 - Top, detail of the 1%A-inchthermowell nozzle to the pressurizer shell l.D. surface J-groove weld. Bottom, detail of the weld boss (not wetted) to the pressurizer shell O.D.

surface. The weld boss was intended for future welding replacement thermowell nozzle if needed.

Page 16 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 f Stainless Steel Clad Carbon Steel Sampling or Level Sensing Nozzles (MK-30) Welded to Carbon Steel Pressurizer Shell Figure 7 - Detail of the 1-inch level sensing nozzle or 1-inch sampling nozzle to Alloy 600 safe-end weld.

Diaphragm Plate (MK-13)

Alloy 600 (SB-168)

Support Plate (MK-17), Type 304,SA-240 Heater Sheath, MK-28 (5 of 39 for each bundle), Type 316L, SA-213 Figure 8 - Original Heater Bundle Assembly.

Page 17 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

. .. II I

FORGING Figure 9 - Cross sectional view of the original heater bundle assembly.

Page 18 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Alloy 82/182 Field weld Diaphragm plate (MK-13)

Alloy 600 Stainless Steel Cladding Carbon Steel Heater Belt Forging MK-4 Figure 10 - Detail of the Alloy 600 diaphragm plate to stainless steel cladding seal weld (Alloy 82/182)for the original heater bundles.

REPLACEMENT A/

DIAPHRAGM PLATE PIN 1167775-613 (SA-240) BASE METAL HEATER BUNDLE DIAPHRAGM PLATE SHALL BE HELD BUTTERING FLUSH TO BOTTOM OF C'BORE (STAINLESS STEEL)

DURING WELDING Figure 11 - Detail of the stainless steel diaphragm plate to stainless steel cladding seal weld for the replacement lower heater bundle In 2003. The weld metal is Alloy 521152.

Page 19 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Aloy 600 Diaphragm Plate MK-13 WP-119 Deal Nut Diaphragm Plate Type 316L Heater Sheath MK-28 Figure 12 - Top, detail of the Alloy 600 heater sleeve to the Alloy 600 diaphragm plate weld Bottom, detail of the Type 316L heater sheath to the Alloy 600 heater sleeve.

Page 20 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Requested Information 1 (b)

A description of the inspection program forAlloy 82/182/600 pressurizer penetrations and steam space piping connections that has been implemented at your plant. The description should include when the inspections were performed; the areas, penetrations and steam space piping connections inspected; the extent (percentage) of coverage achieved for each location which was inspected; the inspection methods used; the process used to resolve any inspection findings; the quality of the documentation of the inspections (e.g., written report, video record, photographs); and, the basis for concluding that your plant satisfies applicable regulatory requirements related to the integrity of pressurizer penetrations 'and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections were found, indicate what followup NDE was performed to characterize flaws in the leaking penetrations.

Response

The Alloy 82/182/600 pressurizer penetration and steam space connections at TMI are limited to the thirteen carbon steel nozzles to alloy 600/82/182 welded safe-end connections and the two alloy to stainless steel field weld connections. Of the fifteen welded connections, only five are subject to inservice inspection examinations as a result of their size or connection type. Two of the welds (4-inch spray noZzle and spray nozzle safe-end to stainless field weld connection) subject to the actions of this bulletin have had volumetri6 and surface examinations performed in accordance with the requirements of the American Society of Mechahical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, uRules for Inservice Inspection of Nuclear Power.

Plant Components," Code Category B-F, Code Item number B5.40, in the 2nd and 3d ISI intervals for the 4-inch spray nozzle weld and Code Item B5.1 30 in the 2nd interval for the spray nozzle safe-end to stainless field weld. The remaining three welds subject to this bulletin were examined in accordance with the requirements of the ASME Code Section XI, Code Category B-F, Code Item number B5.50 during the 2nd ISI Interval. The ISI examinations performed during the 2nd IS Interval were done in accordance with the 1986 Edition of the ASME Code Section Xl. The ISI examination performed during the 3r Interval was done in accordance with the 1995 Edition of the ASME Code through the 1996 Addenda.

TMI, Unit 1 is now in the 3d Interval with the ISI program, and complies with the 1995 Edition through the 1996 Addenda of the ASME Code Section XI. In a safety evaluation report dated November 7, 2003, the NRC approved the use of a risk-informed methodology for the selection and examination of ASME Code Class 1 and Code Class 2 piping welds. However, none of the subject pressurizer safe-end welds have been re-examined under the risk-informed ISI program. In addition, two welds (4-inch spray nozzle and spray nozzle safe-end to stainless field weld connection) have had a volumetric examination performed in accordance with Supplement 10 to Appendix VIII of the ASME Code Section XI since its implementation in November of 2002.

All examinations, surface and volumetric, were recorded on hard copy data sheets. No video or photographs were used to supplement these examinations. There have been no recordable indications on these five welds that required disposition. All surface Page 21 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 examinations covered 100% of the required examination area. The ultrasonic examinations, although partially obstructed in some cases, achieved greater than 90% of the required examination volume coverage.

In addition to the nondestructive examinations listed below, all steam space pressurizer Alloy 82/182 welds were visually examined each refueling outage, at a minimum, in accordance with the pressure test requirements of the ASME Code Section Xl, Category B-P.

The following is a listing of the previous inspections:

Weld RCT0002PROO09BMWELD (4-inch spray nozzle) This weld was first examined in October 2001. The weld was examined by dye penetrant and by ultrasonic methods.

The ultrasonic examination was performed using 30 degree and 45 degree shear wave, and 45 degree longitudinal wave transducers. There were no indications recorded from either the dye penetrant or ultrasonic examinations; therefore, no disposition was required.

This weld was examined again in November 2003 as part of an expanded scope inspection. The weld was examined by dye penetrant and by ultrasonic (UT) methods.

The UT examination was performed using performance demonstration initiative (PDI) qualified 60 degree longitudinal/dual, 45 degree shear/single,'and 45 degree longitudinal/dual techniques. There were no indications recorded from either the dye penetrant or for the ultrasonic examinations.

Weld RCT0002PROO08BMWELD (2%.-inch pressure relief nozzle) This weld was examined in September 1997. The weld was examined by the dye penetrant method.

There were no indications recorded from this examination.

Weld RCT0002PROO07BMWELD (2%-inch pressure relief nozzle) This weld was examined in September 1997. The weld was examined by the dye penetrant method.

There were no indications recorded from this examination.

Weld RCT0002PROO06BMWELD (2%.-inch pressure relief nozzle) This weld was examined in September 1997. The weld was examined by the dye penetrant method.

There were no indications recorded from this examination.

Weld SP0021 BMWELD (4-inch safe-end to stainless steel component field weld) This weld was examined in September 1999. The weld was examined by the dye penetrant and ultrasonic methods. There was one geometric indication outside the weld that was recordable.

Basis for Concluding Regulatory Requirements are Satisfied As stated above, the TMI, Unit 1 pressurizer connections affected by this bulletin are limited to the 13 Alloy 82/182 full penetration nozzle to safe-end welds and two safe-end to stainless steel field welds. The completion of volumetric, surface, and visual examinations required by the ASME Code Section Xl without any evidence of Page 22 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 recordable, relevant indications, or through-wall leakage of the carbon steel surface, is assurance of the previous integrity of the Alloy 82/182 connhebtions.

The specific regulatory requirements are listed below with the associated response addressing how the requirement is met.

Compliance with Design Requirements: 10 CFR 50, Appendix A- General Design Criteria (GOC)

Criterion 14 - Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

The TMI, Unit 1 pressurizer connections are designegd, fabricated, tested, and examined in accordance with the requirements of the'ASME Code Section III, 'Requirements for Design and Manufacture of Nuclear Power Plant Components," and the ASME Code Section Xl. In general, the controls established by these construction and inspection codes assures that the reactor vessel maintains an extremely low probability of rapidly propagating failure and gross rupture.

The BMV examination technique to be used in the TMI, Unit 1, Fall 2005 refueling outage is a reliable means for identifying the very low leakage rates potentially associated with Alloy 82/182 cracking. Therefore, based on the design, materials, and examination methods, the TMI, Unit 1 pressurizer continues to comply with the requirements of GDC 14.

In addition, in the case of the pressurizer steam space Alloy 82/182 locations, appropriate mitigation will be performed to provide added assurance that these connections will have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner, and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient thermal stresses, and (4) size of flaws."

The reactor coolant pressure boundary is designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident Page 23 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 conditions, the boundary behaves in a non-brittle manner, and the probability of rapidly propagating fracture is minimized.

Criterion 32 - Inspection of Reactor Coolant Pressure Boundary "Components which are part of the reactor coolant pressure boundary shall be designed to permit (1)periodic inspection and testing of important areas and features to assess their structural and leak tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

The TMI, Unit 1 pressurizer connections described in this bulletin, which are part of the reactor coolant pressure boundary (RCPB), were designed to accommodate the visual, surface, and volumetric examination requirements of the ASME Code Section Xl. I Ongoing ASME Code Section Xl examinations will ensure the continued structural and leak tight integrity of these connections.

Compliance with Operating Requirement: 10 CFR 50.36 - Plant Technical Specifications TMI, Unit 1 Technical Specifications include requirements and associated action statements addressing reactor coolant leakage. The TMI, Unit 1 Technical Specification limits for reactor coolant operational leakage are one gallon per minute (gpm) for unidentified leakage, 10 gpm for identified leakage, and no reactor coolant system strength boundary leakage (reference TMI, Unit 1 Technical Specifications 3.1.6, "Leakage"). Compliance with the zero non-isolable leakage criteria is met by conducting inspections and repairs in accordance with ASME Code, Section Xl, and 10 CFR R 50.55a,"Codes and standards," as described below.

Compliance with Inspection Requirements: 10 CFR 50.55a and the ASME Code Section Xl 10 CFR 50.55a, "Codes and standards," requires that inservice inspection and testing be performed in accordance with the requirements of the ASME Code, Section Xl, "Inservice Inspection of Nuclear Plant Components." Section Xl contains applicable rules for examination, evaluation, and repair of code class components, including the RCPB.

Compliance with Quality Assurance Requirements: 10 CFR 50, Appendix B Criterion V of Appendix B to 10 CFR 50 Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Page 24 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 The ASME Code Section Xl. required visual examinations are performed using procedures that contain specific acceptance criteria or detailed recording criteria that are subsequently evaluated for acceptability. The visual examinations are performed using detailed instructions with a combination of qualitative and quantitative standards for the essential examination variables. Examinations of the pressurizer steam space Alloy 82/182 connections will be performed using standardized AmeFGen procedures, which include appropriate acceptance criteria.

Criterion IX of Appendix B to 10 CFR 50 Criterion IX of Appendix B to 10 CFR 50 states that special processes, including nondestructive testing, shall be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requiremients.

The pressurizer examinations at will be performed by certified Level II or Level IlIl VT-2 examiners using AmerGen approved procedures with additional detailed instructions as necessary.

Criterion XVI of Appendix B to 10 CFR Part 50 Criterion XVI of Appendix B to 10 CFR Part 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. For significant conditions adverse' to quality, the measures taken shall*

include root cause determination and corrective action to preclude repetition of the adverse conditions.

The identification of an unacceptable visual indication requires repair, replacement or acceptance by analytical evaluation. In all cases, these indications would be tracked by the TMI, Unit 1 Corrective Action Program (CAP).

Requested Information I (c)

A description of the Alloy 82/182/600 pressurizer penetration and steam space piping connection inspection program that will be implemented at your plant during the next and subsequent refueling outages. The description should include the areas, penetrations and steam space piping connections to be inspected; the extent (percentage) of coverage to be achieved for each location; inspection methods to be used; qualification standards for the inspection methods and personnel; the process used to resolve any inspection indications; the inspection documentation to be generated; and the basis for concluding that your plant will satisfy applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections are found, indicate what followup NDE will be performed to characterize flaws in the leaking penetrations. Provide yourplans for expansion of the scope of NDE to be performed if Page 25 of 27

N Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 circumferential flaws are found in any portion of the leaking pressurizer penetrations or steam space piping connections.

Response

TMI, Unit 1 will be performing a 100% bare metal visual (BMV) examination of the affected pressurizer penetrations described in this bulletin during the next refueling outage. The visual examinations will be performed by certified Level Il/Ill VT-2 visual examiners. In addition, TMI, Unit 1 will continue to perform this BMV examination in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility. Currently, plans are being developed to mitigate susceptible pressurizer upper head locations. TMI, Unit 1 will evaluate the source of any indications and resolve inspection indications.

The affected pressurizer connections described in this bulletin have met, and will continue to meet, all requirements related to the structural and leakage integrity of the FRCPB. This is assured by compliance with the examination requirements of the ASME Code Section Xi and the augmented examinations performed in accordance with this bulletin.

If a leaking penetration is found, a determination would be made, based on the location and nature of the indication, if additional NDE examinations will be performed and the type of repair methodology that will be utilized. In the evaluation required by the corrective action program, a determination would be made ,as to the extent of scope expansion and the type of NDE to be performed. Depending on the particulars of the indication, a best effort ultrasonic examination would be performed to characterize the flaw in the leaking penetration and to assess the condition of the other connections.

The basis for concluding that TMI, Unit 1 satisfies applicable regulatory requirements related to the structural and leakage integrity of the affected pressurizer connections described by this bulletin is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

Requested Information 1 (d)

In light of the information discussed in this bulletin and your understanding of the relevance of recent industry operating experience to your facility, explain why the inspection program identified in your response to item (1)(c) above is adequate for the purpose of maintaining the integrity of your facility's RCPB and for meeting all applicable regulatory requirements which pertain to your facility.

Response

As stated in response to question 1(c) above, TMI, Unit 1 will be performing a BMV examination of the affected pressurizer penetrations described in this bulletin during the next refueling outage. The visual examinations will be performed by certified Level Il/Ill VT-2 visual examiners. In addition, TMI, Unit 1 will continue to perform BMV examinations in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility.

Page 26 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 The basis for concluding that TMI, Unit 1 BMV examinations meet applicable regulatory requirements related to the structural and leakage integrity is provided above in the

'Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response. I Page 27 of 27

Text

7 - -.- - _t I - Z-r AmerGen b Exekmn.

www.exelonCOTP.COM An Exelon Company AmerGen Energy Company, LLC Nuclear 4300 Winfield Road Exelon Generation Warrenville, IL 60555 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.54 (f)

RS-04-109 5928-04-20180 July 27, 2004 United States Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, Maryland 20852 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 Three Mile Island Nuclear Station, Unit No. 1 Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Initial response to NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized-Water Reactors"

Reference:

NRC Bulletin 2004-01, "Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurized-Water Reactors," dated May 28, 2004 The purpose of this letter is to provide the Exelon Generation Company, LLC and the AmerGen Energy Company, LLC initial, sixty-day, response to the referenced NRC bulletin. The responses to NRC Bulletin 2004-01 questions 1(a), 1(b), 1(c), and 1(d) detailing the description and fabrication of the pressurizer connections; the current and proposed inspection program for the components; and the basis for assuring reactor coolant pressure boundary integrity for pressurizer penetrations and steam space connections are provided in the attachments to this letter.

,At

U. S. Nuclear Regulatory Commission July 27, 2004 Page 2 Should you have any questions concerning this letter, please contact David J.

Chrzanowski at (630) 657-2816.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 2720L4 A /.X Keith R. Jury Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC AmerGen Energy Company, LLC Attachments: Attachment 1, Initial Response to NRC Bulletin 2004-01, Braidwood Station, Units 1 and 2 Attachment 2, Initial Response to NRC Bulletin 2004-01, Byron Station, Units 1 and 2 Attachment 3, Initial Response to NRC Bulletin 2004-01, Three Mile Island Nuclear Station, Unit 1

aI- -J Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Page 1 of 18

Attachment I Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units I and 2 Requested Information 1 (a)

A description of the pressurizer penetrations and steam space piping connections at your plant. At a minimum, this description should include materials of construction (e.g.,

stainless steel piping and/or weld metal, Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.), joint design (e.g., partial penetration welds, full penetration welds, bolted connections, etc.), and, in the case of weldedjoints, whether or not the weld was stress-relieved prior to being put into service. Additional information relevant with respect to determining the susceptibility of your plant's pressurizer penetrations and steam space piping connections to PWSCC should also be included.

Response

Braidwood Station Unit 1 and Unit 2 are 4-loop pressurized water reactors with the nuclear steam supply system designed by Westinghouse Electric Company, LLC (Westinghouse). The Braidwood Station units have one Westinghouse Model 84 pressurizer in each unit. Braidwood Station, Unit 1 began commercial operation on July 29, 1988. Braidwood Station, Unit 2 began commercial operation on October 17, 1988.

The pressurizers were fabricated by Westinghouse at their Pensacola, Florida facility.

Each of the pressurizers has six piping connections: five steam space connections and one surge line connection. The five steam space connections are: one spray line connection, one relief valve line connection, and three safety valve line connections.

The pressurizer manways (one per pressurizer) are not discussed in this response since, they do not contain any Alloy 600/82/182. Table'1 lists-the pressurizer steam space connections and corresponding identifier.

Table I Braidwood Station Pressurizer Steam Space Connection Listing Unit Steam Space Connection Identifier 1 safety valve line connection 1PZR-01-SE-02 1 safety valve line connection 1PZR-01-SE-03 1 safety valve line connection 1PZR-01-SE-04 1 relief valve line connection 1PZR-01-SE-05 I spray line connection 1PZR-01 -SE-06 2 safety valve line connection 2PZR-01-SE-02 2 safety valve line connection 2PZR-01-SE-03 2 safety valve line connection 2PZR-01-SE-04 2 relief valve line connection 2PZR-01-SE-05 2 spray line connection 2PZR-01-SE-06 All six of the pressurizer connections have an Alloy 82/182 weld connection from the low alloy steel pressurizer nozzles to the stainless steel safe-end attachments. These six locations are the only Alloy 600/82/182 based connections on the Braidwood Station pressurizers. Piping connections downstream of these safe-end connections do not have any Alloy 600/82/182 based components in any application. Also, the eight, 3/4 inch instrument line connections and the 3/4 inch sample line connection do not incorporate Page 2 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 any Alloy 600/82/182 based components in their connections to the pressurizer or in any of their downstream connections. The Braidwood Station pressurizers have heater penetrations that use stainless steel sleeves connected to the pressurizer by stainless steel welds.

Pressurizer Steam Space Connections - Braidwood Station. Unit 1 Connection 1PZR-01-SE-02 1PZR-01-SE-02 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "A" pressurizer safety relief valve (1RY8001 A) to the pressurizer upper head nozzle (1PZR-01-N4A). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld IPZR-01-SE-02, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment. A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 1PZR-01-SE-03 1PZR-01-SE-03 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the 'Bi"pressurizer safety relief valve (1RY8001 B) to the pressurizer upper head nozzle (1PZR-01-N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-182, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

Page 3 of 18

7 Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 During fabrication, weld 1PZR-01-SE-03, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were five repairs performed on this weld: 2.25" long and 3/8" deep from the inside diameter (ID) between radiography location marks 0 to 1; 1.5" long and 11/16" deep from the outside diameter (OD) between radiography location marks 5 to 6; 0.75" long and 11/16" deep from the OD between radiography location marks 6 to 7; 1" long and 11/16" deep from the OD also between radiography location marks 6 to 7; and 1.25" long and 1/4" to 3/8" deep from the OD between radiography location marks 11 to 0.

Subsequent to the repairs, a final, acceptable, radiography examination (RT) was performed.

Connection IPZR-01-SE-04 1PZR-01 -SE-04 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "C" pressurizer safety relief valve (I RY8001 C) to the pressurizer upper head nozzle (1PZR-01 -N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade'F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 1PZR-01-SE-04, identified in shop records as weld 4C on nozzle A, did not receive any post-weld heat treatment. A review of the fabrication records indicated that there were five spot repairs performed on this weld. The records did not specify the length for any of the repair areas which were located between radiography location marks: 2 to 3, 4 to 5, 5 to 6, 6 to 7, and 7 to 8. All repairs were listed as being 0.2" deep from the OD.

Subsequent to the repairs, a final, acceptable, RT was performed.

Connection 1PZR-01-SE-05 1PZR-01 -SE-05 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (1PZR-01-N2).

Page 4 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182,.Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Up stream of the nozzle to safe-end weld is a weld that joins the four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the SA-403, Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule 160 piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455B and 1RY455C (SA-403, Grade WP304 valve bodies).

During fabrication, weld 1PZR-01-SE-05, identified in shop records as weld 2 on nozzle E, did not receive any post-weld heat treatmnent. A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 1PZR-01-SE-06 '

1PZR-01 -SE-06 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (.1 RY455A and 1RY456) to the pressurizer upper head nozzle (1PZR-01 -N3). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304, 450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455A and 1RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld 1PZR-01-SE-06, identified in shop records as weld 3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Page 5 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Pressurizer Steam Space Connections - Braidwood Station, Unit 2 Connection 2PZR-01-SE-02 2PZR-01-SE-02 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "A" pressurizer safety relief valve (2RY8001 A) to the pressurizer upper head nozzle (2PZR-01-N4A).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160,' SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 2PZR-01-SE-02, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment.

Additional Information for 2PZR-01-SE-02 A review of the fabrication records indicated that there were four repairs performed on this weld. The records did not specify the length for any of the repair areas which were located between radiography location marks: 6 to 7, 10 to 11, 11 to 12, and 0 to 1. All repairs were listed as being approximately 0.5" deep from the OD.

Subsequent to the repairs, a final, acceptable, RT was performed.

Connection 2PZR-01-SE-03 2PZR-01-SE-03 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "B" pressurizer safety relief valve (2RY8001 B) to the pressurizer upper head nozzle (2PZR-01-N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Page 6 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 2PZR-01-SE-03, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there was one repair performed on this weld. The repair area is listed as being 0.5" long and approximately 0.625" deep from the ID between radiography location mark 5 to 6.

Subsequent to the repair, a final, acceptable, RT was performed.

Connection 2PZR-01-SE-04 2PZR-01-SE-04 is a six inch nominal pipe diameter butt welded connection joiniing the pipe that connects the UC" pressurizer safety relief valve (2RY8001 C) to the pressurizer upper head nozzle (2PZR-01-N4B). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-182, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld 2PZR-01-SE-04, identified in shop records as weld 4C on nozzle A, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 2PZR-01-SE-05 2PZR-01-SE-05 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (2PZR-01-N2).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal Page 7of 18

V Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Up stream of the nozzle to safe-end weld is a weld that joins tlee four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the SA-403, Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455B and 2RY455C (SA-403, Grade WP304 valve bodies).

During fabrication,'weld 2PZR-01-SE-05, identified in shop records as weld 2 on nozzle E, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Connection 2PZR-01-SE-06 2PZR-01-SE-06 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (2RY455A and 2RY456) to the pressurizer upper head nozzle (2PZR-01-N3). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304,450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455A and 2RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld 2PZR-01-SE-06, identified in shop records as weld 3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were no repairs to this weld during construction.

Page 8 of 18

Attachment 1 InitiaI Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Requested Information 1 (b)

A description of the inspection program for Alloy 82/182/600 pressurizer penetrations and steam space piping connections that has been implemented at your plant. The description should include when the inspections were performed; the areas, penetrations and steam space piping connections inspected; the extent (percentage) of coverage achieved for each location which was inspected; the inspection methods used; the process used to resolve any inspection findings; the quality of the documentation of the inspections (e.g., written report, video record, photographs); and, the basis for concluding that your plant satisfies applicable regulatory requirements related to the integrity of pressurizerpenetrations and steam space piping connections. If leaking pressurizerpenetrations or steam space piping connections were found, indicate what followup NDE was performed to characterize flaws in the leaking penetrations.

Response

The Alloy 82/182/600 pressurizer penetration and steam space connections at Braidwood Station are limited to the ten (five per unit) low alloy steel nozzle to stainless steel, Alloy 82, welded safe-end connections. All the pressurizer welds subject to the actions of this bulletin (identified in Table 2) have had volumetric and surface examinations performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components,' Code Category B-F,1 Code Item number B5.40.2 The first inservice inspection (ISI) Interval examinations were performed in accordance with the 1983 Edition, Summer 1983 addenda, of the ASME Code Section Xl. Examinations performed in the second ISI interval were performed in accordance with the 1989 Edition, no addenda, of the ASME Code Section Xl. The examination areas for these welds were those identified in the ASME Code Section Xl, Figure IWB-2500-8, 'Similar and Dissimilar Metal Welds in Components and Piping.'

This examination area is the same for the ASME Code Section Xl Editions 1983 through 1989.

Both Braidwood Station units are now in the 2nd Interval with the IS program, as stated above, based on the 1989 Edition of the ASME Code Section Xl. On February 20, 2001, the NRC approved the use of a risk informed methodology for the selection and examination of ASME Code Class 1 and Code Class 2 piping welds; however, none of the subject pressurizer safe-end welds have been re-examined under the risk informed ISI program. In addition, there have been no volumetric re-examinations performed on the pressurizer welds subject to the actions of this bulletin since the implementation of Supplement 1 to Appendix VII 4 of the ASME Code Section XI since its implementation in November of 2002.

1 "Pressure Retaining Dissimilar Metal Welds" 2 Pressurizer, Nozzle-to-Safe-end Butt Welds, Nominal Pipe Size 4 Inch or Larger" 3 Qualification Requirements for Dissimilar Metal Piping Welds" 4'Performance Demonstration for Ultrasonic Examination Systems" Page 9 of 18

Attachment 1 Inital Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 All examinations, surface and volumetric, were recorded on hard copy data sheets. No video or photographs were used to supplement these examinations. There have been no recordable indications on these ten welds that required disposition. All surface examinations covered 100% of the required examination area. The ultrasonic examinations, although partially obstructed in some cases, achieved greater than 90% of the required examination volume coverage. I In addition to the nondestructive examinations listed below, all steam space pressurizer Alloy 82 welds were visually examined, at a minimum, each refueling outage in accordance with the pressure test requirements of the ASME Code Section Xl, Category B-P. These examinations have, in the past, been performed with the insulation in place with a four-hour hold time at normal operating pressure and temperature.

There has not been any indication of leaking pressurizer penetrations or steam space piping connections at Braidwood Station and therefore, there has not been a need for dispositions, additional nondestructive examinations (NDE), examination expansions, or flaw characterizations.

Page 10 of 18

Attachment 1 ,.

Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Table 2 Braidwood Station Alloy 82/182 Pressurizer Welds Examination History Weld Exam Date Exam Method Exam Technique Results Disposition/

Identifier . Data Sheet Volumetric (Vol) 450 longitudinal wave-transducer No Recordable Indication(s) (NRI) N/A 1PZR-01-SE-02 October 1995 (450 L-wave) ._95BR1_UT-052 Surface (S) Dye Penetrant (PT) NRI N/A 95BR1PT-026 N/A Vol 450 L-wave and 700 shear wave NRI 94BR1-UT-061 and IPZR-01-SE-03 March 1994 transducer (S-wave) Ni94BRI-UT-062' S PT Three arc strike Indications Acceptable

._ 94BR1-PT-021 N/A Vol 450 L-wave and 700 S-wave NRI 94BR1-UT-061 and 1PZR-01-SE-04 March 1994 . 94BR1-UT-062 S PT NRI N/A Acceptable Vol 450 L-wave 360° root geometry Indication 89BR1-UT-110 and October 1989 89BR1-UT-111 1PZR-01-SE-05 S PT NRI 89BR1-PT-056 Vol 450 L-wave NRI N/A 988R1-UTD-040 September 1998 N/A S S PTPT NRIN/

N~i98BR1-PT-039 Acceptable Vol 450 L-wave 3600 root geometry Indication 89BR1-UT-108 and October 1989 . 89BR1-UT-109 1PZR-01-SE-06 0 S PT .NRI TNi898R1-PT-057 N/A Vol 450 L-wave NRI N/A September 1998 98BR1-UTD-045 S PT -NRI N/A S PT N 98BR1-PT-040 Vol 450 L-wave NRI 95BR2-PT-064 2PZR-01-SE-02 April 1996 N/A S PT NRI 958R1-UT-156 Page 11 of 18

'I Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Table 2 Braidwood Station Alloy 821182 Pressurizer Welds Examination History (continued)

Weld Exam Date Exam Method Exam Technique -. Results Disposition/

Identifier . Data Sheet Vol 450 L-wave NRI 94BRN-UT-060 2PZR-01-SE-03 October 1994 N1A 2PZR-A1-SE-04 October 1994 9 NBR2-UT-/0A PNRI 94BR2-PT-038 Indication was dispositioned as Vol S 55° L-wave Pt NRI from the alloy Ultrasonic Indications non-relevantN/A and acceptable April 1990 821182 cladding Interface 90BR2-UT-100 90BR2-UT-101and s PT NRIN/A 2PZR-01-SE-05 NRI 90R2-PT-043 Pipe Inner diameter geometry Acceptable Vol 450 L-wave Indications 99BR2-UTD-034 April 1999

. PT NRI N/A 99BR2-PT-049 Indication was dispositioned as Vol55, -wveUltrasonic indications from the alloy non-relevant and acceptable Vol 55 11wave82/182 cladding Interface -90BR2-UT-102 and April199090BR2-UT-103 s PT NRIN/A 2PZR-01-SE-06 S TNI90BR2-PTO 048 Pipe inner diameter geometry Acceptable Vol 4501 L-wave' Indications 99BR2-UTD-035 April 1999 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

S PT_ _ _ _ _ _ _ _ _ _ _ _ __ _ _ __ _ _ __

NRI

_ _ __ _ _ _ _ _ _ _ _ _99 N/A.

B R 2-PT-0 5 Page 12 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Table 2 Braidwood Station Alloy 82/182 Pressurizer Welds Examination History Basis for Concluding Regulatory Requirements are SatisfiJd As stated above, the Braidwood Station pressurizer connections affected by this Bulletin are limited to the ten (five per unit), Alloy 82/182 full penetration nozzle to safe-end welds. The completion of volumetric, surface, and visual examinations without any evidence of recordable, relevant indications, through-wall leakage or any recordable wastage of the low alloy steel surface, is assurance of the prior integrity of the Alloy 82/182 connections.

Ongoing integrity of the Braidwood Station pressurizer steam space Alloy 82/182 connections is assured by performing 100% bare metal visual (BMV) examinations, at a minimum, each refueling outage (approximately 18 months). The examination will be performed until mitigation (i.e., weld overlays of the Alloy 82/182 locations) is performed on all pressurizer steam space connections.

The specific regulatory requirements are listed below with the associated response addressing how the requirement is met.

Compliance with Design Requirements: 10 CFR 50. Appendix A - General Design Criteria (GDC)

Criterion 14 - Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

The Braidwood Station pressurizer steam space connections are designed, fabricated, tested, and examined in accordance with the requirements of the ASME Code Section Ill, "Requirements for Design and Manufacture of Nuclear Power Plant Components,"

and the ASME Code Section Xl. In general, the controls established by these construction and inspection codes assure that the reactor coolant pressure boundary maintains an extremely low probability of rapidly propagating failure and gross rupture.

The BMV examination technique to be used in the Braidwood Station, Unit 1 Fall 2004 refueling outage, and in the Unit 2 Spring 2005 refueling outage, is a reliable means for identifying the very low leakage rates potentially associated with alloy 82/182 cracking.

Therefore, based on the design, materials, and examination methods, the Braidwood Station pressurizers continue to comply with the requirements of GDC 14.

In addition, in the case of the pressurizer steam space Alloy 82/182 locations, modifications in the form of weld overlays using primary water stress corrosion cracking (PWSCC) resistant material will be performed to provide added assurance that these connections will have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Page 13 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Criterion 31- Fracture Prevention of Reactor Coolant Pressure Boundary "Thereactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner, and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient thermal stresses, and (4) size of flaws."

The Braidwood Station pressurizer steam space connections are designed in accordance with the requirements of the ASME Code Section III with sufficient margin to the stresses encountered during operating, maintenance, testing, and postulated accident conditions. The pressurizer steam space connections, even the Alloy 82/182 welds, will continue to behave in a non-brittle manner. Ongoing BMV examinations of the pressurizer steam space connections at Braidwood Station will assure sufficient margin from rapidly propagating fracture until the susceptibility of Alloy 82/182 to PWSCC has been acceptably mitigated.

Criterion 32 - Inspection of Reactor Coolaht Pressure' Boundary "Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features, to assess their structural and leak tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel."

The Braidwood Station pressurizer steam space connections were designed to accommodate the visual, surface, and volumetric examinations of the ASME Code Section Xl. While the Alloy 82/182 to safe-end connections present limitations to current, fully qualified performance demonstration initiative (PDI) volumetric examination, ongoing BMV examinations will assure the structural and leak tight integrity of the pressurizer steam space connections at Braidwood Station.

Compliance with Operating Requirement: 10 CFR 50.36 - Plant Technical Specifications Braidwood Station Technical Specifications include requirements and associated action statements addressing reactor coolant pressure boundary (RCPB) leakage. The Braidwood Station Technical Specification limits for reactor coolant system operational leakage are one gallon per minute (gpm) for unidentified leakage, 10 gpm for identified leakage, and no pressure boundary leakage (reference Braidwood Station Technical Specifications Section 3.4.13, "RCS Operational Leakage"). Compliance with the zero non-isolable leakage criteria is met by conducting inspections and repairs in accordance with ASME B&PV Code, Section Xl, and 10 CFR 50.55a, "Codes and standards," as described below.

The unidentified leakage limit of one gpm is established as a quantity which can be accurately measured while sufficiently low to ensure early detection of leakage.

Page 14 of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Leakage of this magnitude ban be reasonably detected within a short time, thus providing confidence that cracks associated with such leakage will not develop into a critical size before mitigating actions can be taken. If a through-wall boundary leak increases to the point where it is detected by the containment radiation monitor, mass balance calculations, or reactor containment sump level readings, the leak must be evaluated in accordance with the specified acceptance criteria and the plant must be shut down if the leak is determined to be a non-isolable reactor coolant system (RCS) pressure boundary fault.

In addition, Braidwood Station has implemented controls and expectations to address RCS leakage below Technical Specification limits. Exelon Generation Company (EGC) procedure ER-AP-331-1003, 'RCS Leakage Monitoring and Action Plan," has been implemented to assure adequate monitoring of RCS leakage and to provide minimum actions that could be taken at various RCS leakage levels.

Compliance with Inspection Requirements: 10 CFR 50.55a and the ASME Code Section Xl 10 CFR 50.55a, "Codes and standards," requires that inservice inspection and testing be performed in accordance with the requirements of the ASME Code, Section Xl, "Inservice Inspection of Nuclear Plant Components."Section XI contains applicable rules for examination, evaluation, and repair of code class components, including the RCPB.

However, as discussed above, Braidwood Station is using a risk-informed methodology forthe selection and examination of similar and dissimilar metal piping welds. While the Alloy 82/182 pressurizer steam space piping connections contain limitations in the performance of a fully qualified PDI volumetric examination, the current ISI program does not require that these welds be selected for volumetric examination. To compensate for the volumetric examination limitations, the Braidwood Station pressurizer steam space Alloy 82/182 connections will be visually examined each refueling outage until appropriate mitigation has been employed.

Compliance with Quality Assurance Requirements: 10 CFR 50, Appendix B Criterion V of Appendix B to 10 CFR 50 Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

The ASME Code Section Xl required visual examinations are performed using procedures that contain specific acceptance criteria or detailed recording criteria that are subsequently evaluated for acceptability. The visual examinations are performed using detailed instructions with a combination of qualitative and quantitative standards for the essential examination variables. Supplemental BMV examinations of the pressurizer Page 15of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 steam space Alloy 82/182 connections at Braidwood Station will be performed using standardized EGC procedures, which include appropriate acceptance criteria.

Criterion IX of Appendix B to 10 CFR 50 Criterion IX of Appendix B to 10 CFR 50 states that special processes, including nondestructive testing, shall be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

The pressurizer steam space connection BMV examinations at Braidwood Station will be performed by certified Level II or Level III examiners using EGC approved procedures with additional detailed instructions, as necessary.

Criterion XVI of Appendix B to 10 CFR 50 Criterion XVI of Appendix B to 10 CFR 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

For significant conditions adverse to quality, the measures taken shall include root cause determination and corrective action to preclude repetition of the adverse conditions.

The identification of an unacceptable visual indication requires repair, replacement or acceptance by analytical evaluation. In all cases, these indications would be tracked by tle EGC Corrective Action Program (CAP). In the case of a significant adverse condition, the CAP requires determination of the cause of the failure, evaluation of the extent of condition, and assignment of appropriate corrective actions to preclude recurrence. The EGC CAP meets the requirements of 10 CFR 50, Appendix B, Criterion XVI.

Requested Information I (c)

A description of the Alloy 82/182/600 pressurizer penetration and steam space piping connection inspection program that will be implemented at your plant during the next and subsequent refueling outages. The description should include the areas, penetrations and steam space piping connections to be inspected; the extent (percentage) of coverage to be achieved for each location; inspection methods to be used; qualification standards for the inspection methods and personnel; the process used to resolve any inspection indications; the inspection documentation to be generated; and the basis for concluding that your plant will satisfy applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizerpenetrations or steam space piping connections are found, indicate what followup NDE will be performed to characterize flaws in the leaking penetrations. Provide your plans for expansion of the scope of NDE to be performed if circumferential flaws are found in any portion of the leaking pressurizer penetrations or steam space piping connections.

Page 16of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2

Response

Braidwood Station will be performing a BMV examination of all affected pressurizer penetrations and steam space piping connections during the next refueling outages for each unit. The visual examinations will be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331-1001, "Boric Acid Corrosion Control (BACC) Inspection Locations, Implementation, and Inspection Guidelines,' and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation." In addition, Braidwood Station will continue to perform this BMV examination in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

Braidwood Station will use the guidance of ER-AP-331-1002 and LS-AA-125, 'Corrective Action Program (CAP) Procedure," to evaluate the source of any indications and to resolve inspection indications. Any evidence of pressure boundary leakage will require disposition under TS 3.4.13, "RCS Operational Leakage,! 10 CFR 50.72, 'Immediate notification requirements for operating nuclear power reactors," and the EGC Corrective Action Program.

The examinations will be documented in accordance with ER-AA-335-015 and ER-AP-331-1002 with written reports. All affected pressurizer penetrations and steam space piping connections have met and will continue to'meet all requirements related to the structural and leakage integrity of the RCPB. This is assured by compliance with the examination requirements of the ASME Code Section Xl and-the augmented examinations performed in accordance with this bulletin.

The basis for concluding that Braidwood Station satisfies applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

If a leaking penetration is found, a determination will be made, based on the location and nature of the indication, if additional NDE examinations will be performed or whether the location will be directly repaired by a weld overlay. In the evaluation required by the corrective action program, a determination will be made as to the extent of scope expansion and the type of NDE to be performed. All pressurizer upper head penetrations to steam space piping Alloy 82/182 weld connections are of such a configuration that a fully qualified PDI ultrasonic examination is not possible. Therefore, depending on the particulars of the indication, a best effort ultrasonic examination would be performed to characterize the flaw in the leaking penetration and to assess the condition of the other connections. Based on the results, and the quality of the examination technique, EGC may elect to preventively overlay some or all pressurizer upper head Alloy 82/182 locations.

Page 17of 18

Attachment 1 Initial Response to NRC Bulletin 2004-01 Braidwood Station, Units 1 and 2 Requested Information 1 (d)

In light of the information discussed in this bulletin and your understanding of the relevance of recent industry operating experience to your facility, explain why the inspection program identified in your response to item (1)(c) above is adequate for the purpose of maintaining the integrity of your facility's RCPB and for meeting all applicable regulatory requirements which pertain to your facility.

Response

As stated in response to question 1(c) above, Braidwood Station will be performing a BMV examination of all affected pressurizer penetrations and steam space piping connections during the next refueling outages for each unit. The visual examinations will be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331 -1001, "Boric Acid Corrosion Control (BACC)

Inrspection Locations, Implementation, and Inspection Guidelines," and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation."

In addition, Braidwood Station will continue to perform BMV examinations in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

The basis for concluding that the Braidwood Station BMV examination program meets all applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

Page 18 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Page 1 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Requested Information 1 (a)

A description of the pressurizer penetrations and steam space piping connections at your plant. At a minimum, this description should include materials of construction (e.g.,

stainless steel piping and/or weld metal, Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.), joint design (e.g., partial penetration welds, full penetration welds, bolted connections, etc.), and, in the case of weldedjoints, whether or not the weld was stress-relieved prior to being put into service. Additional information relevant with respect to determining the susceptibility of your plant's pressurizer penetrations and steam space piping connections to PWSCC should also be included.

Response

Byron Station Unit 1 and Unit 2 are 4-loop pressurized water reactors with the nuclear steam supply system designed by Westinghouse Electric Company, LLC (Westinghouse). The Byron Station units have one Westinghouse Model 84 pressurizer in each unit. Byron Station, Unit 1 began commercial operation on September 15, 1985.

Byron Station, Unit 2 began commercial operation on August 21, 1987.

The pressurizers were fabricated by Westinghouse at their Pensacola, Florida facility.

Each of the pressurizers has six piping connections: fiVe steam space connections and one surge line connection. The five steam space connections are: one spray line connection, one relief valve line connection, and three safety valve line connections.

The pressurizer manways (one per pressurizer) are not discussed in this response since, they do not contain any Alloy 600/82/182. Table'1 lists the pressurizer steam space connections and corresponding identifier.

Table 1 Byron Station Pressurizer Steam Space Connection Listing Unit Steam Space Connection Identifier 1 safety valve line connection (NOZZLE A-4C) (1RY01S) PN-04-F4 1 safety valve line connection (NOZZLE B-4B1) (RYO1S) PN-05-F5 1 safety valve line connection (NOZZLE C-4A) (1RY01S) PN-06-F6 1 l relief valve line connection (NOZZLE D) (1RY01 S) PN-03-F3 1 spray line connection (NOZZLE E) (1RY01 S) PN-02-F2 2 safety valve line connection (NOZZLE A-4C) (2RY01S) PN-03-F3 2 safety valve line connection (NOZZLE B-4B) (2RY01S) PN-04-F4 2 safety valve line connection (NOZZLE C-4A) (2RY01S) PN-05-F5 2 relief valve line connection (NOZZLE D) (2RY01S) PN-06-F6 2 spray line connection (NOZZLE E) (2RY01 S) PN-02-F2 All six of the pressurizer connections utilize an Alloy 82/182 weld connection from the low alloy steel pressurizer nozzles to the stainless steel safe-end attachments. These six locations are the only Alloy 600/82/182 based connections on the Byron Station pressurizers. The piping downstream of these connections do not have any Alloy 600/82/182 based components in any application. Also, the eight, 3/4 inch instrument line connections and the 3 inch sample line connection do not incorporate any Alloy Page 2 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 600/82/182 based components in their connection to the pressurizer or in any of their downstream connections. The Byron Station pressurizers have heater penetrations that use stainless steel sleeves connected to the pressurizer by stainless steel welds.

Pressurizer Steam Space Connections - Byron Station, Unit I Connection (1RY01S) PN-04-F4 (IRY01 S) PN-04-F4 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "C" pressurizer safety relief valve (I RY801 OC) to the pressurizer upper head nozzle (PN-04). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (1RYO1 S) PN-04-F4, identified in shop records as weld 4C on nozzle A, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were repairs to this weld during construction with a final acceptable radiography examination (RT) performed.

Fabrication records do not indicate size, depth or pertinent details of actual repairs performed.

Connection (1RY01 S) PN-05-F5 (1RY01 S) PN-05-F5 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "B" pressurizer safety relief valve (1RY801 OB) to the pressurizer upper head nozzle (PN-05). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-182, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-182, Grade F-316 flange. The flange provides the bolted Page 3 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (1RYO1 S) PN-05-F5, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were repairs to this weld during construction with final acceptable RT performed. Fabrication records do not indicate size, depth or pertinent details of actual repairs performed.

Connection (1RY01S) PN-06-F6 (1RY01 S) PN-06-F6 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "A"pressurizer safety relief valve (1RY801 OA) to the pressurizer upper head nozzle (PN-06). The welded. connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. D6wnstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (1RY01S) PN-06-F6, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were repairs to this weld during construction with final acceptable RT performed. Fabrication records do not indicate size, depth or pertinent details of actual repairs performed.

Connection (ORY01S) PN-02-F2 (lRYOIS) PN-02-F2 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (PN-02).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Page 4 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Upstream of the nozzle to safe-end weld is a weld that joins the four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the SA-403, Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455B and 1RY455C (SA-403, Grade WP304 valve bodies).

During fabrication, weld (1RY01 S) PN-02-F2, identified in shop records as weld PN F2 on nozzle E, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (1RY01S) PN-03-F3 (IRYO1S) PN-03-F3 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (1RY455A and 1RY456) to the pressurizer upper head nozzle. The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 821182 Weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304, 450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 1RY455A and 1RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld (IRY01S) PN-03-F3, identified in shop records as weld PN F3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Page 5 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Pressurizer Steam Space Connections - Byron Station. Unit 2 Connection (2RYOIS) PN-03-F3 (2RYO1S) PN-03-F3 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "C" pressurizer safety relief valve (2RY801 OC) to the pressurizer upper head nozzle (PN-03). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a 8A-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (2RY01S) PN-03-F3, identified in shop records as weld 4C on noizle A, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-04-F4 (2RY01S) PN-04-F4 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the "B" pressurizer safety relief valve (2RY801 OB) to the pressurizer upper head nozzle (PN-04). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA-403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (2RY01S) PN-04-F4, identified in shop records as weld 4B on nozzle B, did not receive any post-weld heat treatment.

Page 6 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units I and 2 A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-05-F5 (2RY01S) PN-05-F5 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the 'A"pressurizer safety relief valve (2RY801 OA) to the pressurizer upper head nozzle (PN-05). The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 and additional SA403, Grade WP304 elbows with the final steam space pipe connection terminating in a stainless steel welded connection between a SA-403, Grade WP304 elbow and a SA-1 82, Grade' F-316 flange. The flange provides the bolted connection between the pressurizer steam space piping and the pressurizer safety valve.

During fabrication, weld (2RY01 S) PN-05-F5, identified in shop records as weld 4A on nozzle C, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle-to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-02-F2 (2RY01S) PN-02-F2 is a four inch nominal pipe diameter butt welded connection joining the pressurizer spray line piping to the four inch pressurizer upper head nozzle (PN-06).

The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld. The spray nozzle design incorporates a 0.120 inch thick thermal sleeve. The thermal sleeve is internally mounted in the nozzle and is attached by four 0.50 inch spot welds, to the stainless steel cladding, at the nozzle inner radius. The upper portion of the thermal sleeve is welded to the inner diameter of the stainless steel safe-end for approximately 450 of the inner circumference. All these attachment welds are stainless steel material, not Alloy 82/182.

Upstream of the nozzle to safe-end weld is a weld that joins the four inch safe-end to the four inch side of a 4" by 6" schedule 160, concentric reducer. This is a stainless steel Page 7 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 weld that joins the SA-1 82, Grade F-316L safe-end material to the SA-403,' Grade WP304 reducer. Upstream of the reducer are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 piping and additional SA-403, Grade WP304 elbows. The six inch schedule piping run transitions to a 4 inch schedule 160 piping run which also contains weld connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455B and 2RY455C (SA-403, Grade WP304 valve bodies).

During fabrication, weld (2RY01S) PN-02-F2 identified in shop records as weld 2 on nozzle E, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Connection (2RY01S) PN-06-F6 (2RY01 S) PN-06-F6 is a six inch nominal pipe diameter butt welded connection joining the pipe that connects the power operated relief valves (2RY455A and 2RY456) to the pressurizer upper head nozzle. The welded connection joins the SA-508, Class 2 low alloy steel pressurizer nozzle material to SA-1 82, Grade F-316L safe-end material with an Alloy 82/182 weld.

Downstream of the nozzle to safe-end weld is a safe-end to elbow weld. This is a stainless steel weld that joins the SA-1 82, Grade F-316L safe-end material to the six inch, schedule 160, SA-403, Grade WP304, 450 elbow. Downstream of the elbow are all stainless steel welded connections between sections of six inch, schedule 160, SA-376, Grade TP 304 pipe segments and additional SA-403, Grade WP304 elbows connecting to a branching 6 by 6 by 3 reducing tee (SA-403, Grade WP304). Two, three inch piping runs continue downstream with welded connections between SA-376, Grade TP 304 pipe segments and SA-403, Grade WP304 elbows. The piping run terminates with welded connections to valves 2RY455A and 2RY456 (SA-1 82, Grade F316 valve bodies).

During fabrication, weld (2RY01S) PN-06-F6, identified in shop records as weld 3 on nozzle D, did not receive any post-weld heat treatment.

A review of the fabrication records indicated that there were cladding repairs and a reject of one area of the Safe-End to Nozzle Weld. No specific information to size, length or depth of repairs was noted. Final RT of the nozzle to safe-end weld was performed and determined to be acceptable.

Page8of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Requested Information I (b)

A description of the inspection program for Alloy 82/182/600 pressurizer penetrations and steam space piping connections that has been implement~d at your plant. The description should include when the inspections were performed; the areas, penetrations and steam space piping connections inspected; the extent (percentage) of coverage achieved for each location which was inspected; the inspection methods used; the process used to resolve any inspection findings; the quality of the documentation of the inspections (e.g., written report, video record, photographs);, and, the basis for concluding that your plant satisfies applicable regulatory requirements related to the integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections were found, indicate what followup NDE was performed to characterize flaws in the leaking penetrations.

Response

The Alloy 82/182/600 pressurizer penetration and steam space connections at Byron Station are limited to the ten (five per unit) low alloy steel nozzle to stainless steel, Alloy 82/182, welded safe-end connections. All the pressurizer welds subject to the actions of this bulletin (identified in Table 2) have had volumetric'and surface examinations performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code Category B-F, 5 Code Item number B5.40. 6 The first inservice inspection (ISI) Interval examinations were performed in accordance with the 1983 Edition, Summer 1983 addenda, of the ASME Code Section Xl. Examinations performed in the second ISI interval were performed in accordance with the 1989 Edition, no addenda, of the ASME Code Section Xl. The examination areas for these welds were those identified in the ASME Code Section Xl, Figure IWB-2500-8, Similar and Dissimilar Metal Welds in Components and Piping." This examination area is the same for the ASME Code Section Xl Editions 1983 through 1989.

Byron Station Unit 1 and Byron Station Unit 2 are in the 2nd Interval with the ISI program, as stated above, based on the 1989 Edition of the ASME Code Section XI. On February 20, 2001 the NRC approved the'use of a risk informed methodology for the selection and examination of ASME Code Class 1 and Code Class 2 piping welds; however, none of the subject pressurizer safe-end welds have been re-examined under the risk informed ISI program. In addition, there have been no volumetric re-examinations performed on the pressurizer welds subject to the actions of this bulletin since the implementation of Supplement 10 7 to Appendix Vil 8of the ASME Code Section Xl since its implementation in November of 2002.

5 "Pressure Retaining Dissimilar Metal Welds" 6 Pressurizer, Nozzle-to-Safe-end Butt Welds, Nominal Pipe Size 4 Inch or Larger' "Qualification Requirements for Dissimilar Metal Piping Welds" 8 Performance Demonstration for Ultrasonic Examination Systems" Page 9 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 All examinations, surface and volumetric, were recorded on hard copy data sheets. No video or photographs were used to supplement these examinations. There have been no recordable indications on these ten welds that required disposition. All surface examinations covered 100% of the required examination area. The ultrasonic examinations, although partially obstructed in some cases, achieved greater than 90% of the required examination volume coverage.

In addition to the nondestructive examinations listed in Table 2, all steam space pressurizer Alloy 82/182 welds were visually examined each refueling outage, at a minimum, in accordance with the pressure test requirements of the ASME Code Section Xl, Category B-P. Also, Byron Station performed a 100% bare metal visual (BMV) examination of all affected Unit 2 pressurizer penetrations and steam space piping connections during the Spring 2004 refueling outage. This examination was performed in accordance with the industry's Materials Reliability Program (MRP) recommendations as described in a Leslie Hartz, Chair, MRP Senior Representatives, letter to PWR licensees dated January 20, 2004.

There has not been any indication of leaking pressurizer penetrations or steam space piping connections at Byron Station and therefore, there has not been a need for dispositions, additional nondestructive examinations (NDE), examination expansions, or flaw characterizations.

Page 10 of 18

PD Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units I and 2 Table 2 Byron Station Alloy 82/182 Pressurizer Welds Examination History Weld Exam Date Exam Method Exam Technique Results Disposition/

Identifier . . Data Sheet March 1987 Surface (S) Dye Penetrant (PT) NRI NPA (1RY01S) PN-02-F2 October 450 longitudinal wave transducer - N/A 198 Volumetric; (Vol) (40Lwv)NRI 88BYI-UT-084 ahd 1988(45 L-wve)88BYI-UT-086 Acceptable January Vol . 450 L-wave 3600 beam redirection Indication 90BY1-UT-Ill and 1990 N/A S PT NRI 90BY1-PT-020 (1RY0lS) PN-03-F3 450 shear wave transducer Acceptable November Vol (450 S-wave) and Laminations and ID geometry 97BY1-UTD-080 and 1997 450 L-Wave 97BYl-UTD-083 S *PT NRI NIA 97BY1-PT-005 Vol 450 and 60 0 L-wave NRI N/A (1RY01S) PN-04-F4 April 1996 96BY-UTD-143

. S PT-*NRI PT -NRI -N/A 96BY1-PT-016 Acceptable Vol 450 L-Wave 3600 beam redirection Indication 90BY1-UT-084 and January _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _9 B I U- 1 1990 NM s PT NRI 90BY1-PT-20 (1RY01S) PN-05-F5 Acceptable 1997 Vol 450 S-Wave and 450 L-Wave ID geometry 97BY1-UTD-081 and Novmbe 97BYI-UTD-084 1997NRIN/A S PT NRI 97BY1-PT-008 March 1987 5 PT NRI PZR-PT-73 OctoberN/A (ORY01S) PN-06-F6 1988 Vol 450 L-Wave NRI 88BY1-UT-083 and (IRYlS) N-0-F6 98888BYI-UT-085 Acceptable November Vol 450 S-Wave and 450 L-Wave ID geometry 97BY1-UTD-082 and 1997 975LYI-epTD-085 (R 1)PN2-2

__________PN_02 _F2 October 1990 _

Vol__ __ _ I___ _ _ _

550 L-wave

_ _ _ _ _ _ _ _ I__ _ _ _

ID_geometry

_ __ _ _ _ _ _ _ I__

Acceptable 90BY2-UT-084 Page 11 of 18

Attachment 2 . ,

Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Table 2 Byron Station Alloy 82/182 Pressurizer Welds Examination History (continued)

Weld Exam Date Exam Method Exam Technique Results Disposition/

Identifler .

  • Data Sheet (2RY01S) PN-02-F2 N/A September S PT NRI 90BY2-PT-044 1990 SP R March 1995 Vol 700 and 450 L-Wave NRI N/A

_____ ____95BY2-UTD-066, February S PT NRI N5Y2PTA 6 (2RY01S) PN-03-F3 1995 95BY2-PT-062 Vol 450 and 600 L-Wave NRI B2R09-UT-053 April 2001 N/A9UT05 S PNRI B2R09-PT-018 March 1995 Vol 700 and 450 L-Wave NRI 95BY2-UTD-066 February (ROS954- 5 PTTNIN/A . NRI 95BY2-PT-078 October Vol OctoberN/A700 and 450 L-Wave NRI 93BY2-UT-123 and 19 93BY2-UT-124 S~eptember PT NRI N/A (2RY01S) PN-05-F5 1993 S PT NRI 93BY2-PT-032 Vol 450 and 600 L-Wave NRI N/A AprIl 2001 132R09-UT-070-N/A S PT NRI 32R09-PT-018 October Vol 550 L-Wave ID geometry Acceptable 1990 S TNIN/A (2RY01S) PN-06-F6 S PT NRI 90BY2-PT-048 April 2001 Vol 450 and 600 L-Wave ID geometry B2R09-UTal1

. . S PT NRI B2R09-PT-019 Page 12 of 18

Y Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Basis for Concluding Regulatory Requirements are Satisfied As stated above, the Byron Station pressurizer connections affected by this Bulletin are limited to the ten (five per unit), Alloy 82/182 full penetration nozzle to safe-end welds.

The completion of volumetric, surface, and visual examinations without any evidence of recordable, relevant indications, through-wall leakage or any recordable wastage of the low alloy steel surface, is assurance of the previous integrity of the Alloy 82/182 connections.

Ongoing integrity of the Byron Station pressurizer steam space Alloy 82/182 connections is assured by performing BMV examinations, at a minimum, each refueling outage (approximately 18 months). The examination will be performed until mitigation (i.e., weld overlays of the Alloy 82/182 locations) is performed on all pressurizer steam space connections.

The specific regulatory requirements are listed below with the associated response addressing how the requirement is met.

Compliance with Design Requirements: 10 CFR 50, Appendix A - General Desiqn Criteria (GDC)

Criterion 14 - Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely 16w probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

The Byron Station pressurizer steam space connections are designed, fabricated, tested, and examined in accordance with the requirements of the ASME Code Section l1l, "Requirements for Design and Manufacture of Nuclear Power Plant Components,"

and the ASME Code Section Xl. In general, the controls established by these construction and inspection codes assures that the reactor coolant pressure boundary maintains an extremely low probability of rapidly propagating failure and gross rupture.

The BMV examination technique to be used in the Byron Station, Unit 1 Spring 2005 refueling outage, and in the Unit 2 Fall 2005 refueling outage, is a reliable means for identifying the very low leakage rates potentially associated with Alloy 82/182 cracking.

Therefore, based on the design, materials, and examination methods, the Byron Station pressurizers continue to comply with the requirements of GDC 14.

In addition, in the case of the pressurizer steam space Alloy 82/182 locations, modifications in the form of weld overlays using primary water stress corrosion cracking (PWSCC) resistant material will be performed to provide added assurance that these connections will have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Page 13 of 18

-Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient thermal stresses, and (4) size of flaws."

The Byron Station pressurizer steam space connections are designed in accordance with the requirements of the ASME Code Section III with sufficient margin to the stresses encountered during operating, maintenance, testing, and postulated accident conditions.

The pressurizer steam space connections, even the Alloy 82/182 welds, will continue to behave in a non-brittle manner. Ongoing BMV examinations of the pressurizer steam space connections at Byron Station will assure sufficient margin from rapidly propagating fracture until the susceptibility of Alloy 82/182 to PWSCC has been acceptably mitigated.

Criterion 32- Inspection of Reactor Coolant Pressure Boundary

."Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features

, to assess their structural and leak tight integrity, and (2) an appropriate material

. surveillance program for the reactor pressure vessel."

The Byron Station pressurizer steam space connections were designed to accommodate the visual, surface, and volumetric examinations of the ASME Code Section Xl. While the Alloy 82/182 to safe-end connections present limitations to current, fully qualified performance demonstration initiative (PDI) volumetric examination, ongoing BMV examinations will assure the structural and leak tight integrity of the pressurizer steam space connections at Byron Station.

Compliance with Operating Requirement: 10 CFR 50.36 - Plant Technical Specifications Byron Station Technical Specifications include requirements and associated action statements addressing reactor coolant pressure boundary (RCPB) leakage. The Byron Station Technical Specification limits for reactor coolant system operational leakage are one gallon per minute (gpm) for unidentified leakage, 10 gpm for identified leakage, and no pressure boundary leakage (reference Byron Station Technical Specifications Section 3.4.13, "RCS Operational Leakage"). Compliance with the zero non-isolable leakage criteria is met by conducting inspections and repairs in accordance with ASME B&PV Code, Section Xl, and 10 CFR 50.55a, "Codes and standards," as described below.

The unidentified leakage limit of one gpm is established as a quantity which can be accurately measured while sufficiently low to ensure early detection of leakage.

Leakage of this magnitude can be reasonably detected within a short time, thus providing confidence that cracks associated with such leakage will not develop into a Page 14 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 critical size before mitigating actions can be taken. If a through-wall boundary leak increases to the point where it is detected by the containment radiation monitor, mass balance calculations, or reactor containment sump level readings, the leak must be evaluated in accordance with the specified acceptance criteria and the plant must be shut down if the leak is determined to be a non-isolable reactor coolant system (RCS) pressure boundary fault.

In addition, Byron Station has implemented controls and expectations to address RCS leakage below Technical Specification limits. Exelon Generation Company (EGC) procedure ER-AP-331-1003, ."RCS Leakage Monitoring and Action Plan," has been implemented to assure adequate monitoring of RCS leakage and to provide minimum actions that could be taken at various RCS leakage levels.

Compliance with Inspection Requirements: 10 CFR 50.55a and the ASME Code Section Xl 10 CFR 50.55a, "Codes and standards," requires that inservice inspection and testing be performed in accordance with the requirements of the ASME B&PV Code, Section Xl, "Inservice Inspection of Nuclear Plant Components." Section Xl contains applicable rules for examination, evaluation, and repair of code class components, including the RCPB.

However, as discussed above, Byron Station is using a risk-informed methodology for the selection and examination of similar and dissimilar metal piping welds. While the Alloy 82/182 pressurizer steam space piping connections contain limitations in the performance of a'fully qualified PDI volumetric examination, the current ISI program does not require that these welds be selected for volumetric examination. To compensate for the volumetric examination limitations, the Byron Station pressurizer steam space Alloy 82/182 connections will be visually examined each refueling outage until appropriate mitigation has been employed.

Compliance with Quality Assurance Requirements: 10 CFR 50. Appendix B Criterion V of Appendix B to 10 CFR 50 Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

The ASME Code Section Xl required visual examinations are performed using procedures that contain specific acceptance criteria or detailed recording criteria that are subsequently evaluated for acceptability. The visual examinations are performed using detailed instructions with a combination of qualitative and quantitative standards for the essential examination variables. Supplemental BMV examinations of the pressurizer steam space Alloy 82/182 connections at Byron Station will be performed using standardized EGC procedures, which include appropriate acceptance criteria.

Page 15 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Criterion IX of Appendix B to 10 CFR 50 Criterion IX of Appendix B to 10 CFR 50 states that special processes, including nondestructive testing, shall be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements:

The pressurizer steam space connection BMV examinations at Byron Station will be performed by certified Level II or Level IlIl examiners using EGC approved procedures with additional detailed instructions, as necessary.

Criterion XVI of Appendix B to 10 CFR 50 Criterion XVI of Appendix B to 10 CFR 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

For significant conditions adverse to quality, the measures taken shall include root cause determination and corrective action to preclude repetition of the adv6rse conditions.

The identification of an unacceptable visual indication requires repair, replacement or acceptance by analytical evaluation. In all cases, these indications would be tracked by the EGC Corrective Action Program (CAP). In the case of a significant adverse condition, the CAP requires determination of the cause of the failure, evaluation of the extent of condition, and assignment of appropriate corrective actions to preclude recurrence. The EGC CAP meets the requirements of 10 CFR 50, Appendix B, Criterion Xv~l.

Requested Information I (c)

A description of the Alloy 82/182/600 pressurizer penetration and steam space piping connection inspection program that will be implemented at your plant during the next and subsequent refueling outages. The description should include the areas, penetrations and steam space piping connections to be inspected; the extent (percentage) of coverage to be achieved for each location; inspection methods to be used; qualification standards for the inspection methods and personnel; the process used to resolve any inspection indications; the inspection documentation to be generated; and the basis for concluding that your plant will satisfy applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections are found, indicate what followup NDE will be performed to characterize flaws in the leaking penetrations. Provide your plans for expansion of the scope of NDE to be performed if circumferential flaws are found in any portion of the leaking pressurizer penetrations or steam space piping connections.

Response

As stated in the response to question 1(b), Byron Station performed a 100% BMV examination of the Unit 2 Alloy 82/182 pressurizer steam space connections during the recent Spring 2004 refueling outage. Byron Station will continue to perform a 100%

Page 16 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 BMV examination of all affected pressurizer penetrations and piping connections during the next refueling outages for each unit. The visual examinations will continue to be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331-1001, "Boric Acid Corrqsion Control (BACC)

Inspection Locations, Implementation, and Inspection Guidelines," and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation." In addition, Byron Station will continue to perform this BMV examination in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

Byron Station will use the guidance of ER-AP-331-1002 and LS-AA-125, 'Corrective Action Program (CAP) Procedure," to evaluate the source of any indications and to resolve inspection indications. Any evidence of pressure boundary leakage will require disposition under TS 3.4.13, 'RCS Operati6nal Leakage," 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors," and the EGC Corrective Action Program.

The examinations will be documented in accordance with ER-AA-335-015 and ER-AP-331-1002 with written reports. All affected pressurizer penetrations and steam space piping connections have met and will continue to meet all requirements related to the structural and leakage integrity of the RCPB. This is assured by compliance with the examination requirements of the ASME Code Section Xl and the augmented examinations performed in accordancewith this bulletin.

The basis for concluding that Byron Station satisfies applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

If a leaking penetration is found, a determination will be made, based on the location and nature of the indication, if additional NDE examinations will be performed or whether the location will be directly repaired by a weld overlay. In the evaluation required by the corrective action program, a determination will be made as to the extent of scope expansion and the type of NDE to be performed. All pressurizer upper head penetrations to steam space piping Alloy 82/182 connections are of such a configuration that a fully qualified PDI ultrasonic examination is not possible. Therefore, depending on the particulars of the indication, a best effort ultrasonic examination would be performed to characterize the flaw in the leaking penetration and to assess the condition of the other connections. Based on the results, and the quality of the examination technique, EGC may elect to preventively overlay some or all pressurizer upper head Alloy 82/182 connections.

Page 17 of 18

Attachment 2 Initial Response to NRC Bulletin 2004-01 Byron Station, Units 1 and 2 Requested Information 1 (d)

In light of the information discussed in this bulletin and your understanding of the relevance of recent industry operating experience to your facility, explain why the inspection program identified in your response to item (1)(c) above is adequate for the purpose of maintaining the integrity of your facility's RCPB and for meeting all applicable regulatory requirements which pertain to your facility.

Response

As stated in response to question 1(c) above, Byron Station Unit 2 performed a BMV examination of all affected pressurizer penetrations and steam space piping connections during the Spring 2004 refueling outage. There were no indications of leakage or boric acid deposits identified during this examination.

Byron Station will be performing a BMV examination of all affected pressurizer penetrations and steam space piping connections during the next refueling outages for each unit. The visual examinations will be performed by certified Level Il/Ill visual examiners qualified in the detection and assessment of boric acid leakage in accordance with EGC procedures ER-AA-335-015, "VT-2 Visual Examination," ER-AP-331-1001, "Boric Acid Corrosion Control (BACC) Inspection Locations, Implementation, and Inspection Guidelines," and ER-AP-331-1002, "Boric Acid Corrosion Control Program Identification, Assessment, and Evaluation."

In addition, Byron Station will continue to perform BMV examinations in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility of the pressurizer penetrations and steam space piping connections. The current plan is to perform complete weld overlays, with PWSCC resistant material, on all susceptible pressurizer upper head locations.

The basis for concluding that the Byron Station BMV examination program meets all applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

Page 18 of 18

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Page 1 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Requested Information 1 (a)

A description of the pressurizer penetrations and steam space piping connections at your plant.

At a minimum, this description should include materials of construction (e.g., stainless steel piping and/or weld metal, Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.),

joint design (e.g., partial penetration welds, full penetration welds, bolted connections, etc.),

and, in the case of weldedjoints, whether or not the weld was stress-relieved prior to being put into service. Additional information relevant with respect to determining the susceptibility of your plant's pressurizer penetrations and steam space piping connections to PWSCC should also be included.

Response

Three Mile Island Nuclear Station (TMI), Unit No.1 is a single-loop pressurized water reactor with the nuclear steam supply system designed by Babcox and Willcox Company (B&W). TMI, Unit I has one B&W 177-FA pressurizer (Figure 1). TMI, Unit 1 began commercial operation on September 2, 1974.

The pressurizer was fabricated by B&W. The pressurizer has eight (8) water space connections, eight (8) steam space connections, one (1) surge line connection, and one (1)

Manway. The eight (8) steam space connections are: one (1) spray line connection; one (1) relief valve line connection; two (2) safety valve line connections; three (3) level sensing connections, and; one (1) vent connection. A description of pressurizer connections materials utilized is contained herein.

Pressurizer Connections - TMI. Unit 1 The pressurizer vessel contains the following 18 penetrations.

Penetration Type Number 1-inch vent nozzle 1 212-inch pressure relief nozzles 3 4-inch spray nozzle 1 I 1/-inch thermowell 1 1-inch level sensing nozzles 6 1-inch sampling nozzle 1 19-inch heater bundle openings 3 10-inch surge nozzle (outside the scope of this document, not steam space) 1 16-inch manway (not discussed in this document as it contains no Alloy 600/182/82)

The pressurizer steam space connections on the upper head include a 16-inch manway, a 1-inch vent and sampling nozzle, three 21/-inch relief nozzles, and a 4-inch spray nozzle. The spray line and attached spray head are connected to the inside end of the spray nozzle and suspended from the upper head, as shown in Figure 2A.

Page 2 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 The following is a discussion of'the penetrations identified above (except for the surge line and manway).

1. Vent Nozzle A 1-inch, Schedule 160 vent nozzle is located at the top center of the upper head to allow complete venting and to permit sampling from the steam space. The vent nozzle was fabricated from Alloy 600 bar (SB-166). The vent nozzle is joined to the interior of the upper head with a partial penetration weld (also called J-groove weld) as illustrated in Figure 2B. Based on information from the vendor, the J-groove weld is buttered at TMI, Unit 1. In addition to the J-groove weld, the vent nozzle installed at TMI, Unit 1 is a two-piece construction as illustrated in Figure 2B. The two Alloy 600 pieces (the top one is called a safe-end) is jointed, without weld butter, by a full penetration V-groove weld with Alloy 82/182. In addition to the J-groove weld, TMI, Unit 1 has an Alloy 82/182 weld boss welded to the pressurizer upper head outside diameter (O.D.) surface as illustrated in Figure 2B. The weld boss is not in contact with the reactor coolant (RC) water. The weld boss was applied during the pressurizer vessel fabrication as a contingency in the event a weld repair is needed.

Table 1 - Originally Installed 1-inch Vent Nozzle Assembly (The pressurizer has 1 vent nozzle, see Figure 2A and 2B)

Connection - a-Material 1-inch vent nozzle, Alloy 600 MK-78 Vent nozzle to upper Alloy 82 or head J-groove weld, 182 WP-95 Butter for J-groove Alloy 82 or weld, WP-94 182 Vent nozzle safe- Alloy 600 end, MK-77 Vent nozzle to safe- Alloy 82 or end weld, WP-92 182 Vent nozzle weld Alloy 82 or boss (not wetted by 182 RC)

Safe-end to stainless Alloy 82 or steel component 182 weld18 Page 3 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

2. 2%-inch Pressure Relief Nozzles Three 21/2-inch pressure relief nozzles are located near the top of the vessel for attachment of pressure relief devices. The nozzles are manufactured from carbon steel with a stainless steel cladding welded on the inner diameter (I.D.). The nozzles are joined to the upper head with a full penetration weld (carbon steel weld) as illustrated in Figure 3.

A type 316 stainless steel safe-end (also called long weld necks) are welded to the top of the nozzles with full penetration V-groove with Alloy 182 as illustrated in Figure 3. Based on information from the vendor, the nozzles were buttered with Alloy 82/182 before the V-groove welding. The flanges on the long weld necks contain eight equally spaced bolt holes for attaching the pressure relief valves or power operated relief valve block valve to the nozzle.

The materials used for the pressure relieve nozzle assembly are summarized in Table 2.

Table 2 - Originally Installed 21/2-inch Pressure Relief Nozzle Assembly.

(The'pressurizer has 3 pressure relief nozzles, see Figure 3)

-Connection Material 21/z-inch pressure. SS Clad relief nozzle carbon steel Safe-end (also called Type 316, long weld neck) SA-182 Pressure relief nozzle to safe-end weld, Alloy 182/82 WP-91 (a)

Pressure Relief Alloy 182 Nozzle Butter, WP-73 (a) For WP-91, the tack and root layer was Alloy 82 and the balance of the weld and weld repair was Alloy 182.

3. 4-inch Spray Nozzle The 4-inch, Schedule 120 spray nozzle is located on the upper head of the pressurizer vessel, connecting the external 2Y.-inch stainless steel spray line from the discharge of a reactor coolant pump with the internal stainless steel spray line and spray head. The nozzle is mounted normal to the pressurizer upper head, entering at a 450 angle from the horizontal. The nozzle body is carbon steel with stainless steel weld clad at the l.D. surface. The nozzle body is joined to the upper head with a full penetration weld (i.e., carbon steel weld).

Alloy 600 transition pieces are welded on both ends of the nozzle, the 8-inch long safe-end connecting with the stainless steel external spray line, and the 5-inch long extension pin (i.e.,

Page 4 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 pipe section) that connects to the internal stainless steel spray line as illustrated in Figure 4. At the top, the Alloy 600 safe-end is attached to the 4-inch spray nozzle by a full penetration Alloy 82 V-groove weld. The Alloy 600 extension pin is attached to the stainless steel weld build-up on the inside of the carbon steel shell by a full penetration Alloy 82/182 V-groove weld without butter. The extension pin is attached to the internal stainless'steelIspray pipe by a full penetration Alloy 82 V-groove weld without butter.

A stainless steel thermal sleeve is installed inside the nozzle assembly to provide protection from thermal stresses. Based on information from the vendor, as illustrated in Figure 5, there are four Alloy 82/182-weld buttons on the inside surface just above the thermal sleeve and four Alloy 82/182 weld pads just below the thermal sleeve. The buttons and pads minimize the chances of the thermal sleeve becoming a loose part. The materials used for the 4-inch spray, nozzle assembly at each unit are summarized in Table 3.

Table 3 - Originally Installed 4-inch Spray Nozzle Assembly (The pressurizer has I spray nozzle, see Figure 4 and Figure 5)

Connection. Material 4-inch spray nozzle, SS clad carbon steel Safe-end, MK-45 Alloy 600 Safe-end to spray Alloy 82 nozzle weld, WP-45 Extension pin, MK-46 Alloy 600 Extension pin to spray nozzle weld, Alloy 82 WP-46 Extension pin to Alloy 82 or internal spray pipe, 182 WP-104 4 upper weld buttons, Alloy 82 or WP-103 182 4 lower weld pads, Alloy 82 or WP-79 182 Safe-end to stainless Alloy 82 or steel component 182 weld Page 5 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

4. 1%-inch Thermowell The 1%/-inch thermowell is a closed penetration located in the side of the pressurizer shell about 9 feet from the bottom of the vessel. The nozzle is located just above the upper heater bundle, extending 4 inches into the water space. The thermowell is made of Alloy 600. The outer end of the thermowell is threaded to hold the RTE sensor. The thermowell is welded to the interior of the vessel wall with an Alloy 82/182 partial penetration (J-groove) weld as illustrated in Figure
6. Alloy 82/182 butter was applied. The weld was generally performed similar to the 1-inch vent J-groove weld (See Figure 2B and Figure 6). In addition to the J-groove weld, TMI, Unit 1 had Alloy 82/182 weld boss welded to the pressure vessel O.D. surface (not wetted, see Figure 6).

The weld boss was applied during the fabrication to facilitate future weld repair if needed. The materials used for the 11/2-inch thermowell assembly at each unit are summarized in Table 4.

Table 4 - Originally Installed 11%-inch Thermowell Nozzle Assembly (The pressurizer has 1 thermowell nozzle, see Figure 6)

Connection . Material 1%2-inch thermowell, Alloy 600 SB-166 Pressurizer butter for Alloy 82 or J-grpove weld, WP- 182 80 Thermowell to pressurizer (I.D. side) Alloy 182 J-groove weld, WP-81 Weld boss on pressurizer (O.D.) Alloy 82 or WP-107 (not wetted 182 by RC)

5. Level Sensing Nozzles The six 1-inch, Schedule 160 level sensing nozzles (also called level taps) are located in the pressurizer vessel consisting of two pairs of three nozzles each at high and low elevations. The nozzle bodies are of carbon steel with stainless steel weld clad on the l.D. surface. Each level sensing nozzle is joined to the carbon steel pressurizer shell with a full penetration carbon steel weld. At the top, as illustrated in Figure 7, an Alloy 600 safe-end is attached to each level-sensing nozzle with a full penetration Alloy 82 V-groove weld without butter. The materials used for the level sensing nozzle assembly are summarized in Table 5.

Page 6 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

6. 1-inch Sampling Nozzle The pressurizer has one 1-inch Schedule 160 sampling nozzle (also called sampling tap) connecting to a sampling line. The sampling nozzle is located at the same elevation as the 1A-inch thermowell described in Section 4. The sampling nozzle is mbunted similar to the level sensing nozzles (described in Section 5 and Figure 7). The materials used for the 1-inch level sampling nozzle assembly are summarized in Table 5.

Table 5 - Originally Installed Level Sensing Nozzle and Sampling Nozzle Assembly (The pressurizer has 6 level sensing nozzles and 1 sampling nozzle, see Figure 7)

Cnne'&tion' -'Mate'ral 1-inch level sensing nozzle and SS clad 1-inch sampling nozzle, carbon steel Level sensing nozzle and Alloy 600 sampling nozzle safe-end, Safe-end to level sensing nozzle and sampling nozzley weld, WP-63

7. Pressurizer Heater Bundle The heater belt forgings have three openings to accommodate the heater bundle assemblies.

Each heater bundle is an assembly consisting of three parallel disks drilled to hold 39 individual immersion heaters as shown in Figure 9. The outermost of the three disks, the heater bundle diaphragm plate, mates with and forms a seating surface with the heater belt forging penetration. The other two discs, called support plates, are fabricated from stainless steel.

Figure 10 gives a cross-sectional view of the pressurizer heater bundle assembly pressure retaining items. The heater bundle cover plates are bolted on the outer surface of the diaphragm plates, holding the diaphragm plates against the mating surface and providing support for the heater bundle assembly. As illustrated in Figure 11, a seal weld provides the pressure boundary between the heater bundle diaphragm plate and the heater belt forging. No structural credit is given for the seal weld.

The original heater bundles contain Alloy 600/182/82. The diaphragm plate is fabricated from Alloy 600 plate (SB-168). In addition, there are short heater sleeves machined from Alloy 600 bar and are attached to the inner side of the Alloy 600 diaphragm plate by an Alloy 82 partial penetration groove weld without buttering as illustrated in Figure 12. Each Type 316L stainless steel heater sheath passes through the diaphragm plate and sleeve and is attached to Alloy 600 sleeve by an Alloy 82 fillet weld shown in Figure 12. The heater bundles materials discussed in this section for each unit are summarized in Table 6.

Page 7 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 2003 Replacement Heat Bundle The original lower heater bundle was found leaking and was replaced in 2003. The replacement heater bundle has 12 larger diameter (1.25-inch) higher power heaters in lieu of the 39 heaters (0.66-inch) in the original heater bundle. The replacement diaphragm plate is fabricated from Type 304 stainless steel, instead of Alloy 600. Due to the original seal weld (original Alloy 600 diaphragm plate to the stainless steel cladding weld) being performed with Alloy 82 and subsequent weld repairs being performed with Alloy 152/52, the new seal weld for the replacement heater bundle diaphragm plate used Alloy 152/52. The new seal weld is illustrated in Figure 11.

Table 6 - Originally Installed Heater Bundle Nozzle Assembly (The pressurizer has 3 heater bundles)

Alloy 600 Diaphgm -pType! 2 Mat-::

Heater Bunde C ompbnents-' .

Diaphragm plate, Alloy 600, SB-168 (1 per heater bUndle)

Heater sheath,Tye36,S21 (39 per heater bundle) Type 316L SA-213 Heater sleeve, Alloy 600, SB-166 (39 pler heater bundle)

Diaphragm plate seal (Field) weld Alloy 82 or 182 (1 per heater bundle)

Heater sleeve to diaphragm Alloy 82 plate weld, WP-119 (39 per heater bundle)

Heater sheath to diaphragm Alloy 82 plate weld, WP-120 (39 per heater bundle)

Castellated seal nuts to diaphragm plate tack weld, WP- Alloy 82 126 Page 8 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Table 7 - Replacement Heater Bundle Nozzle Assembly

'Stainless'Steel' Diaphragm TypeHeater- Matral b'-

Bundle Diaphragm plate, (3 per pressurizer) Stainless Steel Heater sheath, (12 Per heater bundle) Stainless Steel Diaphragm plate seal (Field) Weld (1 per Alloy 52/152 heater bundle)

Heater sheath to diaphragm plate weld, Stainless Steel WP-121 (WP-123 for Weld DB)

(a) I he two original bundles Tor the original lower neater bundle.

(b) The replacement heater bundle for the original lower heater bundle

8. Post-Weld Stress Relief Heat Treatment The post-weld stress relieve heat treatment (PWHT)'of pressurizer connections were typically performed at 1100-11 50 0F. Based on information from the vendor, the Alloy 600 heater bundle diaphragm plates and heater sleeves did not receive any PWHT: In addition, none of the field welds performed either at the time of the driginal pressurizer installation or during field repair and modification would have received any PWHT. This includes the following known replacement or field repairs:

The heater bundle diaphragm plate to shell cladding seal weld (including the replacement heater bundle in 2003).

. All welds connecting the nozzles (or safe-ends) to external piping.

Page 9 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 VENT AND SAMPLING NOZZLE SPRAY NOZZLE

.4, --

HEAD -

" SPRAY UNE AND SUPPORTS

% INTERNAL SUPPORTS LEVEL SENSING NOZZLE

> SHELL a SUPPORT PLATE ASSEMBLY

. ,, I I THERMOWELL - HEATER BELT LEVEL SENSING NOZZLE SURGE DIFFUSER

- SURGE NOZZLE Figure 1 - The general arrangement of pressurizer penetrations at B&W 177-FA units. TMI, Unit I also has 36 Internal stainless steel ladder rungs welded to the I.D. surface of the pressurizer vessel.

Page 10 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 TOP VIEW UFTING LUGS PRESSURE RELIEF NOZZLES NOZZLE MANAY

> VENT AND SAMPLING NOZZLE PRSSR RELIE SIDE VIEW

.- 1E PRESSURE RELIEF I NOZZLE (1 OF 3)

MANWAY FORGING w dSPRAY NOZZLE SPRAY UNE WeldBi TMI-1 Figure 2A - Penetrations at the upper head of pressurizer vessel. Note, the manway boss was weld buildup at TMI-1, instead of a forging.

Page 11 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Final Bore Alloy 600 Safe End, MK-77 Alloy 600 Vent and Sampling Nozzle,'

I MK-78 Carbon Allt

'Steel Upper No; Head, MK-5 OD surface of

>X the pressurizer upper head Figure 2B - Top, 1-inch Schedule 160 Alloy 600 vent to Alloy 600 safe-end weld. Bottom, detail of the 1-inch vent and sampling nozzle to pressurizer upper head J-groove weld and weld boss (not wetted by RC).

Page 12 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Integral flange and bolt holes I Stainless steel safe end (also called long weld neck) MK-32 PT, WP-73 I

Carbon steel pressure relief nozzle, MK-31 Carbon Steel A'- I I Pressurizer Upper Head I-Stainless Steel Cladding ' _

Figure 3 - Detail of the three 2/2-Inch pressure relief nozzle to stainless steel safe-end weld.

Page 13 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 WP-102 Stainless Steel Thermal Sleeve Carbon steel spray nozzle, MK-9 I

. I. .*

Stainless Steel Cladding Alloy 600 Extension Pin (MK-46)

Alloy 600 Extension Pin Stainless 1/8" dia.

7- 'Steel Interal Spray Pipe electrode Figure 4 - Top, detail of the 4-inch spray nozzle assembly. Bottom, detail of the extension pin to the internal spray pipe weld. See Figure 5 for detail of the weld buttons for positioning the thermal sleeve.

Page 14 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

'WP-103 Upper Weld Buttons Lay the four %" dia.weld buttons @900 after the 4 lower weld pads and the stainless steel sleeve are in place.

(The buttons are on the Alloy 600 safe end).

Stainless Steel WP-79 ?-  %'

4 Weld Pads @900

/

Lay the lower weld pads after inserting the stainless steel thermal sleeve (The lower weld pads are on the Alloy 600 Extension Pin)

Figure 5 - Detail of the weld buttons for positioning the stainless steel thermal sleeve inside the 4-inch spray nozzle.

Page 15 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Outside the pressurizer Alloy 600 Thermowell (MK-29) vessel Inside the I pressurizer vessel Carbon steel pressurizer shell Stainless steel cladding Alloy 60 Thermowell

{/ D MK-29 O.D. surface of the pressurizer shell Figure 6 - Top, detail of the 1%A-inchthermowell nozzle to the pressurizer shell l.D. surface J-groove weld. Bottom, detail of the weld boss (not wetted) to the pressurizer shell O.D.

surface. The weld boss was intended for future welding replacement thermowell nozzle if needed.

Page 16 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 f Stainless Steel Clad Carbon Steel Sampling or Level Sensing Nozzles (MK-30) Welded to Carbon Steel Pressurizer Shell Figure 7 - Detail of the 1-inch level sensing nozzle or 1-inch sampling nozzle to Alloy 600 safe-end weld.

Diaphragm Plate (MK-13)

Alloy 600 (SB-168)

Support Plate (MK-17), Type 304,SA-240 Heater Sheath, MK-28 (5 of 39 for each bundle), Type 316L, SA-213 Figure 8 - Original Heater Bundle Assembly.

Page 17 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1

. .. II I

FORGING Figure 9 - Cross sectional view of the original heater bundle assembly.

Page 18 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Alloy 82/182 Field weld Diaphragm plate (MK-13)

Alloy 600 Stainless Steel Cladding Carbon Steel Heater Belt Forging MK-4 Figure 10 - Detail of the Alloy 600 diaphragm plate to stainless steel cladding seal weld (Alloy 82/182)for the original heater bundles.

REPLACEMENT A/

DIAPHRAGM PLATE PIN 1167775-613 (SA-240) BASE METAL HEATER BUNDLE DIAPHRAGM PLATE SHALL BE HELD BUTTERING FLUSH TO BOTTOM OF C'BORE (STAINLESS STEEL)

DURING WELDING Figure 11 - Detail of the stainless steel diaphragm plate to stainless steel cladding seal weld for the replacement lower heater bundle In 2003. The weld metal is Alloy 521152.

Page 19 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Aloy 600 Diaphragm Plate MK-13 WP-119 Deal Nut Diaphragm Plate Type 316L Heater Sheath MK-28 Figure 12 - Top, detail of the Alloy 600 heater sleeve to the Alloy 600 diaphragm plate weld Bottom, detail of the Type 316L heater sheath to the Alloy 600 heater sleeve.

Page 20 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 Requested Information 1 (b)

A description of the inspection program forAlloy 82/182/600 pressurizer penetrations and steam space piping connections that has been implemented at your plant. The description should include when the inspections were performed; the areas, penetrations and steam space piping connections inspected; the extent (percentage) of coverage achieved for each location which was inspected; the inspection methods used; the process used to resolve any inspection findings; the quality of the documentation of the inspections (e.g., written report, video record, photographs); and, the basis for concluding that your plant satisfies applicable regulatory requirements related to the integrity of pressurizer penetrations 'and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections were found, indicate what followup NDE was performed to characterize flaws in the leaking penetrations.

Response

The Alloy 82/182/600 pressurizer penetration and steam space connections at TMI are limited to the thirteen carbon steel nozzles to alloy 600/82/182 welded safe-end connections and the two alloy to stainless steel field weld connections. Of the fifteen welded connections, only five are subject to inservice inspection examinations as a result of their size or connection type. Two of the welds (4-inch spray noZzle and spray nozzle safe-end to stainless field weld connection) subject to the actions of this bulletin have had volumetri6 and surface examinations performed in accordance with the requirements of the American Society of Mechahical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, uRules for Inservice Inspection of Nuclear Power.

Plant Components," Code Category B-F, Code Item number B5.40, in the 2nd and 3d ISI intervals for the 4-inch spray nozzle weld and Code Item B5.1 30 in the 2nd interval for the spray nozzle safe-end to stainless field weld. The remaining three welds subject to this bulletin were examined in accordance with the requirements of the ASME Code Section XI, Code Category B-F, Code Item number B5.50 during the 2nd ISI Interval. The ISI examinations performed during the 2nd IS Interval were done in accordance with the 1986 Edition of the ASME Code Section Xl. The ISI examination performed during the 3r Interval was done in accordance with the 1995 Edition of the ASME Code through the 1996 Addenda.

TMI, Unit 1 is now in the 3d Interval with the ISI program, and complies with the 1995 Edition through the 1996 Addenda of the ASME Code Section XI. In a safety evaluation report dated November 7, 2003, the NRC approved the use of a risk-informed methodology for the selection and examination of ASME Code Class 1 and Code Class 2 piping welds. However, none of the subject pressurizer safe-end welds have been re-examined under the risk-informed ISI program. In addition, two welds (4-inch spray nozzle and spray nozzle safe-end to stainless field weld connection) have had a volumetric examination performed in accordance with Supplement 10 to Appendix VIII of the ASME Code Section XI since its implementation in November of 2002.

All examinations, surface and volumetric, were recorded on hard copy data sheets. No video or photographs were used to supplement these examinations. There have been no recordable indications on these five welds that required disposition. All surface Page 21 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 examinations covered 100% of the required examination area. The ultrasonic examinations, although partially obstructed in some cases, achieved greater than 90% of the required examination volume coverage.

In addition to the nondestructive examinations listed below, all steam space pressurizer Alloy 82/182 welds were visually examined each refueling outage, at a minimum, in accordance with the pressure test requirements of the ASME Code Section Xl, Category B-P.

The following is a listing of the previous inspections:

Weld RCT0002PROO09BMWELD (4-inch spray nozzle) This weld was first examined in October 2001. The weld was examined by dye penetrant and by ultrasonic methods.

The ultrasonic examination was performed using 30 degree and 45 degree shear wave, and 45 degree longitudinal wave transducers. There were no indications recorded from either the dye penetrant or ultrasonic examinations; therefore, no disposition was required.

This weld was examined again in November 2003 as part of an expanded scope inspection. The weld was examined by dye penetrant and by ultrasonic (UT) methods.

The UT examination was performed using performance demonstration initiative (PDI) qualified 60 degree longitudinal/dual, 45 degree shear/single,'and 45 degree longitudinal/dual techniques. There were no indications recorded from either the dye penetrant or for the ultrasonic examinations.

Weld RCT0002PROO08BMWELD (2%.-inch pressure relief nozzle) This weld was examined in September 1997. The weld was examined by the dye penetrant method.

There were no indications recorded from this examination.

Weld RCT0002PROO07BMWELD (2%-inch pressure relief nozzle) This weld was examined in September 1997. The weld was examined by the dye penetrant method.

There were no indications recorded from this examination.

Weld RCT0002PROO06BMWELD (2%.-inch pressure relief nozzle) This weld was examined in September 1997. The weld was examined by the dye penetrant method.

There were no indications recorded from this examination.

Weld SP0021 BMWELD (4-inch safe-end to stainless steel component field weld) This weld was examined in September 1999. The weld was examined by the dye penetrant and ultrasonic methods. There was one geometric indication outside the weld that was recordable.

Basis for Concluding Regulatory Requirements are Satisfied As stated above, the TMI, Unit 1 pressurizer connections affected by this bulletin are limited to the 13 Alloy 82/182 full penetration nozzle to safe-end welds and two safe-end to stainless steel field welds. The completion of volumetric, surface, and visual examinations required by the ASME Code Section Xl without any evidence of Page 22 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 recordable, relevant indications, or through-wall leakage of the carbon steel surface, is assurance of the previous integrity of the Alloy 82/182 connhebtions.

The specific regulatory requirements are listed below with the associated response addressing how the requirement is met.

Compliance with Design Requirements: 10 CFR 50, Appendix A- General Design Criteria (GOC)

Criterion 14 - Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

The TMI, Unit 1 pressurizer connections are designegd, fabricated, tested, and examined in accordance with the requirements of the'ASME Code Section III, 'Requirements for Design and Manufacture of Nuclear Power Plant Components," and the ASME Code Section Xl. In general, the controls established by these construction and inspection codes assures that the reactor vessel maintains an extremely low probability of rapidly propagating failure and gross rupture.

The BMV examination technique to be used in the TMI, Unit 1, Fall 2005 refueling outage is a reliable means for identifying the very low leakage rates potentially associated with Alloy 82/182 cracking. Therefore, based on the design, materials, and examination methods, the TMI, Unit 1 pressurizer continues to comply with the requirements of GDC 14.

In addition, in the case of the pressurizer steam space Alloy 82/182 locations, appropriate mitigation will be performed to provide added assurance that these connections will have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary "The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner, and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient thermal stresses, and (4) size of flaws."

The reactor coolant pressure boundary is designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident Page 23 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 conditions, the boundary behaves in a non-brittle manner, and the probability of rapidly propagating fracture is minimized.

Criterion 32 - Inspection of Reactor Coolant Pressure Boundary "Components which are part of the reactor coolant pressure boundary shall be designed to permit (1)periodic inspection and testing of important areas and features to assess their structural and leak tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

The TMI, Unit 1 pressurizer connections described in this bulletin, which are part of the reactor coolant pressure boundary (RCPB), were designed to accommodate the visual, surface, and volumetric examination requirements of the ASME Code Section Xl. I Ongoing ASME Code Section Xl examinations will ensure the continued structural and leak tight integrity of these connections.

Compliance with Operating Requirement: 10 CFR 50.36 - Plant Technical Specifications TMI, Unit 1 Technical Specifications include requirements and associated action statements addressing reactor coolant leakage. The TMI, Unit 1 Technical Specification limits for reactor coolant operational leakage are one gallon per minute (gpm) for unidentified leakage, 10 gpm for identified leakage, and no reactor coolant system strength boundary leakage (reference TMI, Unit 1 Technical Specifications 3.1.6, "Leakage"). Compliance with the zero non-isolable leakage criteria is met by conducting inspections and repairs in accordance with ASME Code, Section Xl, and 10 CFR R 50.55a,"Codes and standards," as described below.

Compliance with Inspection Requirements: 10 CFR 50.55a and the ASME Code Section Xl 10 CFR 50.55a, "Codes and standards," requires that inservice inspection and testing be performed in accordance with the requirements of the ASME Code, Section Xl, "Inservice Inspection of Nuclear Plant Components." Section Xl contains applicable rules for examination, evaluation, and repair of code class components, including the RCPB.

Compliance with Quality Assurance Requirements: 10 CFR 50, Appendix B Criterion V of Appendix B to 10 CFR 50 Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Page 24 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 The ASME Code Section Xl. required visual examinations are performed using procedures that contain specific acceptance criteria or detailed recording criteria that are subsequently evaluated for acceptability. The visual examinations are performed using detailed instructions with a combination of qualitative and quantitative standards for the essential examination variables. Examinations of the pressurizer steam space Alloy 82/182 connections will be performed using standardized AmeFGen procedures, which include appropriate acceptance criteria.

Criterion IX of Appendix B to 10 CFR 50 Criterion IX of Appendix B to 10 CFR 50 states that special processes, including nondestructive testing, shall be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requiremients.

The pressurizer examinations at will be performed by certified Level II or Level IlIl VT-2 examiners using AmerGen approved procedures with additional detailed instructions as necessary.

Criterion XVI of Appendix B to 10 CFR Part 50 Criterion XVI of Appendix B to 10 CFR Part 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. For significant conditions adverse' to quality, the measures taken shall*

include root cause determination and corrective action to preclude repetition of the adverse conditions.

The identification of an unacceptable visual indication requires repair, replacement or acceptance by analytical evaluation. In all cases, these indications would be tracked by the TMI, Unit 1 Corrective Action Program (CAP).

Requested Information I (c)

A description of the Alloy 82/182/600 pressurizer penetration and steam space piping connection inspection program that will be implemented at your plant during the next and subsequent refueling outages. The description should include the areas, penetrations and steam space piping connections to be inspected; the extent (percentage) of coverage to be achieved for each location; inspection methods to be used; qualification standards for the inspection methods and personnel; the process used to resolve any inspection indications; the inspection documentation to be generated; and the basis for concluding that your plant will satisfy applicable regulatory requirements related to the structural and leakage integrity of pressurizer penetrations and steam space piping connections. If leaking pressurizer penetrations or steam space piping connections are found, indicate what followup NDE will be performed to characterize flaws in the leaking penetrations. Provide yourplans for expansion of the scope of NDE to be performed if Page 25 of 27

N Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 circumferential flaws are found in any portion of the leaking pressurizer penetrations or steam space piping connections.

Response

TMI, Unit 1 will be performing a 100% bare metal visual (BMV) examination of the affected pressurizer penetrations described in this bulletin during the next refueling outage. The visual examinations will be performed by certified Level Il/Ill VT-2 visual examiners. In addition, TMI, Unit 1 will continue to perform this BMV examination in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility. Currently, plans are being developed to mitigate susceptible pressurizer upper head locations. TMI, Unit 1 will evaluate the source of any indications and resolve inspection indications.

The affected pressurizer connections described in this bulletin have met, and will continue to meet, all requirements related to the structural and leakage integrity of the FRCPB. This is assured by compliance with the examination requirements of the ASME Code Section Xi and the augmented examinations performed in accordance with this bulletin.

If a leaking penetration is found, a determination would be made, based on the location and nature of the indication, if additional NDE examinations will be performed and the type of repair methodology that will be utilized. In the evaluation required by the corrective action program, a determination would be made ,as to the extent of scope expansion and the type of NDE to be performed. Depending on the particulars of the indication, a best effort ultrasonic examination would be performed to characterize the flaw in the leaking penetration and to assess the condition of the other connections.

The basis for concluding that TMI, Unit 1 satisfies applicable regulatory requirements related to the structural and leakage integrity of the affected pressurizer connections described by this bulletin is provided above in the "Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response.

Requested Information 1 (d)

In light of the information discussed in this bulletin and your understanding of the relevance of recent industry operating experience to your facility, explain why the inspection program identified in your response to item (1)(c) above is adequate for the purpose of maintaining the integrity of your facility's RCPB and for meeting all applicable regulatory requirements which pertain to your facility.

Response

As stated in response to question 1(c) above, TMI, Unit 1 will be performing a BMV examination of the affected pressurizer penetrations described in this bulletin during the next refueling outage. The visual examinations will be performed by certified Level Il/Ill VT-2 visual examiners. In addition, TMI, Unit 1 will continue to perform BMV examinations in each subsequent refueling outage until appropriate mitigation has been implemented to eliminate the PWSCC susceptibility.

Page 26 of 27

Attachment 3 Initial Response to NRC Bulletin 2004-01 Three Mile Island Nuclear Station, Unit No.1 The basis for concluding that TMI, Unit 1 BMV examinations meet applicable regulatory requirements related to the structural and leakage integrity is provided above in the

'Basis for Concluding Regulatory Requirements are Satisfied" section of the question 1(b) response. I Page 27 of 27

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