RS-07-015, Inspection and Mitigation of Alloy 600/82/182 Pressurizer Butt Welds

Inspection and Mitigation of Alloy 600/82/182 Pressurizer Butt Welds

ML070310279
Person / Time
Site:	Byron, Braidwood, Crane
Issue date:	01/30/2007
From:	O'Neill T AmerGen Energy Co, Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-07-20021, RS-07-015
Download: ML070310279 (36)

Text

An Exelon Company AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 RS-07-015 5928-07-20021 January 30, 2007 erGen.

wwwexelonCorp.COM U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 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 :

Inspection and Mitigation of Alloy 600/82/182 Pressurizer Butt Welds Exel6n

Reference:

Letter from M. S. Fertel (Nuclear Energy Institute) to L. A. Reyes (U. S. NRC)

"Industry Actions Associated with Potential Generic Implications of Wolf Creek Inspection Findings," dated January 26, 2007 In October of 2006, while performing inspections of its pressurizer Alloy 82/182 butt welds in accordance with MRP-139, "Primary System Piping Butt Weld Inspection and Evaluation Guideline," a pressurized water reactor (PWR) licensee discovered several circumferential indications in its pressurizer surge, safety and relief nozzles. Because of the potential importance of this issue, Exelon Generation Company, LLC (EGC) and AmerGen Energy Company, LLC (AmerGen), are submitting this letter to notify the NRC of the actions taken or planned for inspecting or mitigating Alloy 600/82/182 butt welds on pressurizer connections for Braidwood Station Units 1 and 2, Byron Station Units 1 and 2, and Three Mile Island Nuclear Station Unit Number 1 (TMI Unit 1).

The details of the Alloy 600/82/182 pressurizer connections and timeframe for full compliance with the Materials Reliability Program (MRP) 139 guidelines for Braidwood Station, Byron Station, and TMI Unit 1 are provided in the attachments to this letter. A listing of EGC and AmerGen commitments related to the near-term enhancements to the pressurizer Alloy 600/82/182 programs is provided in Attachment 1.

Nuclear Exelon Generation 4300 Winfield Road Warrenville, IL 60555

U. S. Nuclear Regulatory Commission Page 2 January 30, 2007 In summary, by December 31, 2007, Braidwood Station Unit 1, Byron Station Units 1 and 2, and Three Mile Nuclear Station Unit No. 1 will be in full compliance with the MRP-139 inspection/mitigation requirements for pressurizer Alloy 600/82/182 components. Braidwood Station Unit 2 will be in full compliance with these requirements by the Spring of 2008. The justification for this extension is included in Attachment 2.

Consistent with the actions discussed in the referenced letter for those plants planning mitigation actions in 2008, EGC is evaluating improvements in the leakage monitoring procedures for Braidwood Station Unit 2 and will provide the details of these actions to the NRC by March 31, 2007.

In addition, EGC is assessing the feasibility of on-line monitoring equipment for Braidwood Station Unit 2 and will provide the details of these actions to the NRC by May 31, 2007. A summary of current EGC and AmerGen PWR reactor coolant system leakage thresholds and actions is provided in Attachment 5.

If Braidwood Station Unit 1, Braidwood Station Unit 2, or TMI Unit 1 should shut down due to excessive primary system unidentified leakage, and if the leakage cannot be confirmed to originate from a source other than the pressurizer, a bare metal visual examination of Alloy 600/82/182 butt weld locations on the pressurizer will be performed to determine whether the leakage originated at those locations identified in this response.

At this time, EGC does not intend to modify the leakage monitoring program or to impose any additional bare metal visual examination requirements at Byron Station since after restart from the Byron Unit 2 refueling outage, beginning April 1, 2007, the Alloy 82/182 pressurizer connections for both Byron Station Unit 1 and Unit 2 will be fully mitigated by the installation of full structural overlays (Byron Station Unit 1 Alloy 82/182 pressurizer connections were fully mitigated in the Fall of 2006).

EGC is also participating in a newly developed EPRI project to perform additional refined crack growth calculations of the limiting pressurizer nozzle Alloy 82/182 weld configuration using 3-dimensional finite element analysis. It is expected that modeling a changing crack shape, rather than a semi-elliptical flaw shape, will result in a significant increase in the estimated time period between through-wall penetration and rupture for the limiting case nozzle.

If you have any questions concerning this submittal, please contact David Chrzanowski at (630) 657-2816.

Thomas S. O'Neill Vice President - Regulatory Affairs Exelon Generation Company, LLC AmerGen Energy Company, LLC

U. S. Nuclear Regulatory Commission Page 3 January 30, 2007 : EGC and AmerGen Commitments - Alloy 600/82/182 Program Enhancements : Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds - Braidwood Station, Units 1 and 2 : Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds - Byron Station, Units 1 and 2 : Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds - Three Mile Island Nuclear Station, Unit 1 : Summary of EGC and AmerGen Reactor Coolant System Leakage Monitoring Thresholds and Actions

Ecc:

Site Vice President - Braidwood Station Site Vice President - Byron Station Site Vice President - Three Mile Island Vice President - Regulatory Affairs Regulatory Assurance Manager - Braidwood Station Regulatory Assurance Manager - Byron Station Regulatory Assurance Manager - Three Mile Island Director Licensing and Regulatory Affairs - Midwest Director Licensing and Regulatory Affairs - Mid-Atlantic Director Engineering - Braidwood Station Director Engineering - Byron Station Director Engineering - Three Mile Island Station Illinois Emergency Management Agency - Division of Nuclear Safety R. R. Janati, Commonwealth of Pennsylvania Exelon Document Control Desk Licensing (Electronic Copy)

KS Correspondence Document Desk Commitment Tracking Coordinator - Cantera Commitment Tracking Coordinator - Kennett Square Jim Riley - NEI Warren Fujimoto - Exelon NSRB Exelon Generation Company, LLC (EGC)

AmerGen Energy Company, LLC (AmerGen)

List of Commitments Alloy 600/82/182 Program Enhancements COMMITTED DATE COMMITMENT TYPE s,

COMMITMENT ONE-TIME PROGRAMMATIC ACTION ACTION YesM0 Yes/NO EGC will provide a submittal to the NRC describing an enhanced reactor coolant system leakage monitoring program for Braidwood Station March 31, 2007 Yes No unit 2. The implementation date of the enhanced program will be provided in the March 2007 submittal.

EGC will provide a submittal to the NRC describing plans for any additional capability which shows value in diagnosing leakage at Braidwood Station Unit 2. This may include, but is not limited to, video monitoring of pressurizer piping, acoustic monitoring in the area of the pressurizer, sensitive humidity monitoring, and other methods May 31, 2007 Yes No currently under evaluation. This program would be implemented at Braidwood Station Unit 2 until the installation of full structural weld overlays on all six Braidwood Station Unit 2 pressurizer Alloy 82/182 connections in May 2008. The implementation date of any leakage detection program will be provided in the May 2007 submittal.

If Braidwood Station Unit 1 should This contingency shut down due to excessive primary would be system unidentified leakage, and if implemented at the leakage cannot be confirmed to Braidwood Station originate from a source other than Unit 1 until the No Yes the pressurizer, a bare metal visual installation of full examination of Alloy 82/182 butt weld structural weld locations on the pressurizer will be overlays on all six performed to determine whether the pressurizer Alloy leakage originated at those locations.

82/182 connections in October 2007.

Exelon Generation Company, LLC (EGC)

AmerGen Energy Company, LLC (AmerGen)

List of Commitments Alloy 600/82/182 Program Enhancements COMMITTED DATE COMMITMENT TYPE COMMITMENT OR "OUTAGE" ONE-TIME PROG~nc Ac-nON AcnON Yes/No Yes/No If Braidwood Station Unit 2 should This contingency shut down due to excessive primary would be system unidentified leakage, and if implemented at No Yes the leakage cannot be confirmed to Braidwood Station originate from a source other than Unit 2 until the the pressurizer, a bare metal visual installation of full examination of Alloy 82/182 butt weld structural weld locations on the pressurizer will be overlays on all six performed to determine whether the pressurizer Alloy leakage originated at those locations.

82/182 connections in May 2008.

If TMI Unit 1 should shut down due to excessive primary system This contingency unidentified leakage, and if the would be leakage cannot be confirmed to implemented at TMI originate from a source other than Unit 1 until the No Yes the pressurizer, a bare metal visual inspection or examination of Alloy 600/82/182 butt mitigation plans are weld locations on the pressurizer will completed in be performed to determine whether October 2007.

the leakage originated at those locations.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Inspection of pressurizer Alloy 82/182 butt welds (there are no Alloy 600 product forms in the Braidwood Station pressurizers) at Braidwood Station Units 1 and 2 in accordance with Materials Reliability Program (MRP) 139 schedule have not yet been completed, but Exelon Generation Company, LLC (EGC) intends to complete mitigation activities on these locations in October 2007 for Braidwood Station Unit 1 and in May 2008 for Braidwood Station Unit 2. The listing of all Braidwood Station pressurizer Alloy 82/182 connections is provided in Table 2-1 below.

Table 2-1 Braidwood Station Alloy 82/182 Connections Details concerning Braidwood Station Unit 1 inspection and mitigation activities are provided in Table 2-2. After mitigation, future inspections of pressurizer butt welds at Braidwood Station Unit 1 will be performed in accordance with industry guidance (i.e.,

MRP-139 "Primary System Piping Butt Weld Inspection and Evaluation Guideline").

Unit Pressurizer Connection Listing Identifier 1

sure line connection 1PZR-01-SE-01 1

safe valve line connection 1 PZR-01-SE-02 1

safe valve line connection 1PZR-01-SE-03 1

safety valve line connection 1PZR-01-SE-04 1

relief valve line connection 1 PZR-01-SE-05 1

spray line connection 1 PZR-01-SE-06 2

sure line connection 2PZR-01-SE-01 2

safe 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 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Table 2-2 Mitigation Information Braidwood Station Unit 1 For Braidwood Station Unit 2, EGC intends to complete all inspection and mitigation activities on the pressurizer Alloy 82/182 locations by the end of May 2008, approximately five months after the December 31, 2007 MRP-139 target date.

Justification for this mitigation completion date for Braidwood Station Unit 2 is provided in this Attachment.

Details concerning Braidwood Station Unit 2 inspection and mitigation activities are provided in Table 2-3.

After mitigation, future inspections of pressurizer butt welds at Braidwood Station Unit 2 will be performed in accordance with industry guidance (i.e.,

MRP-139).

Nozzle MRP-139 Volumetric Inspection Mitigation Completed or to be Requirement Met or to be Met Completed Function /

Susceptible Material Outage Designation Start Date Outage Designation Designation Description (MM/YYYY)

Alloy 82/182 weld A1 R13 10/2007 Mitigation by Weld Surge line connection material Overlay 1 PZR-01-SE-01 10/2007 AlR13 Safety valve line Alloy 82/182 weld A1 R13 10/2007 Mitigation by Weld connection material Overlay 1 PZR-01-SE-02 10/2007 AlR13 Safety valve line Alloy 82/182 weld A1 R13 10/2007 Mitigation by Weld connection material Overlay 1 PZR-01-SE-03 10/2007 AIR13 Safety valve line Alloy 82/182 weld A1 R13 10/2007 Mitigation by Weld connection material Overlay 1 PZR-01-SE-04 10/2007 AlR13 Relief valve line Alloy 82/182 weld A1 R13 10/2007 Mitigation by Weld connection material Overlay 1 PZR-01-SE-05 10/2007 A1 R13 Spray valve line Alloy 82/182 weld A1 R13 10/2007 Mitigation by Weld connection material Overlay 1 PZR-01-SE-06 10/2007 A1 R13

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds 1. Previous Inspection Information Braidwood Station, Units 1 and 2 Table 2-3 Mitigation Information Braidwood Station Unit 2 All twelve of the Braidwood Station pressurizer safe-end welds subject to the actions of MRP-139 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 XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code Category B-F,'Code Item number B5.40 (see Table 2-4) ;

however, none of these pressurizer welds have been subsequently re-examined since the implementation of Supplement 10 to Appendix VIII (i.e., Performance Demonstration Initiative or PDI) of the ASME Code Section XI since its implementation in November of 2002.

Details of the examination history of the ten (five per Unit) steam space Alloy 82/182 pressurizer connections were provided in a letter from K. R. Jury (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "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"," dated July 27, 2004.

Nozzle MRP-139 Volumetric Inspection Mitigation Completed or to be Requirement Met or to be Met Completed Function /

Designation Susceptible Material Description Outage Designation Start Date (MM/YYYIr)

Outage Designation Alloy 82/182 weld A2R13 5/2008 Mitigation by Weld Surge line connection material Overlay 2PZR-01-SE-01 5/2008 A2R13 Safety valve line Alloy 82/182 weld A2R13 5/2008 Mitigation by Weld connection material Overlay 2PZR-01-SE-02 5/2008 A2R13 Safety valve line Alloy 82/182 weld A2R13 5/2008 Mitigation by Weld connection material Overlay 2PZR-01-SE-03 5/2008 A2R13 Safety valve line Alloy 82/182 weld A2R13 5/2008 Mitigation by Weld connection material Overlay 2PZR-01-SE-04 5/2008 A1R13 Relief valve line Alloy 82/182 weld A2R13 5/2008 Mitigation by Weld connection material Overlay 2PZR-01-SE-05 5/2008 A2R13 Spray valve line Alloy 82/182 weld A2R13 5/2008 Mitigation by Weld connection material Overlay 2PZR-01-SE-06 5/2008 A2R13

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds

2. Mitigation Information

3. RCS Leakage Monitoring Braidwood Station, Units 1 and 2 In accordance with commitments made to the NRC Bulletin 2004-01, Braidwood Station has been performing bare metal visuals (BMV) of all steam space pressurizer connections since 2004. BMV examinations were performed on all Alloy 82/182 pressurizer locations, including the surge line connections, during the previous refueling outages. For Braidwood Station Unit 1, the visual examinations were performed in the Spring of 2006 (i.e., A1 R12) ; for Braidwood Station Unit 2 the visual examinations were performed in the Fall of 2006 (i.e., A2R12). In both cases the examinations were acceptable with no indications of leakage or wastage.

Braidwood Station Unit 1 pressurizer connections within the scope of MRP-139 will be mitigated by installing full structural weld overlays (SWOL) in the Fall of 2007 (A1 R13).

A listing is provided in Table 2-2.

Braidwood Station Unit 2 pressurizer connections within the scope of MRP-139 will be mitigated by installing SWOL in May of 2008 (A2R13). A listing is provided in Table 2-3.

Braidwood Station Surveillance Procedures 1/2BwOSR 3.4.13.1, "Unit 1/2 Reactor Coolant System Water Inventory Balance" provide the steps necessary to determine reactor coolant system (RCS) identified and unidentified leakage. The Braidwood Station surveillance along with EGC procedure ER-AP-331-1003, "RCS Leakage Monitoring and Action Plan," establish the controls and expectations for monitoring RCS leakage well below the Technical Specification (TS) required action levels. A summary of the ER-AP-331-1003 leakage levels and corresponding actions is provided in.

Since June 2005, the baseline value for unidentified RCS leakage for Braidwood Station Unit 2 has been 0.08 gallons per minute (gpm) with an associated Action Level of 0.18 gpm.

4. Rationale for Performing Mitigation after December 31, 2007 Braidwood Station Unit 1 will have completed mitigation on all Alloy 82/182 pressurizer connections within the 2007 MRP-139 timeframe and no additional justification is required.

The following provides the justification for the extension for Braidwood Station Unit 2 to achieve full compliance with MRP-139 Alloy 82/182 pressurizer components. Following the application of SWOL in A2R13 (currently scheduled to start on April 21, 2008), all six Unit 2 pressurizer Alloy 82/182 butt welds will be ultrasonically inspected as well as fully mitigated from primary water stress corrosion cracking (PWSCC).

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 This schedule deferral represents an approximate five month extension from the MRP-139 required implementation schedule due date of December 31, 2007. Similar mitigation and inspection activities are planned for Braidwood Station Unit 1 during the A1 R13 outage in October 2007, which are in compliance with the MRP-139 schedule.

Prior to the development and approval of MRP-139, EGC was in the process of developing a proactive inspection and mitigation plan for Alloy 600 PWSCC susceptible locations. Braidwood Station and EGC Engineering developed a long-term asset management Alloy 600 PWSCC mitigation plan. A key input into the selected method of mitigation was the ability to perform volumetric examinations of these weld configurations. For the pressurizer nozzle Alloy 600 weld configurations, EGC NonDestructive Examination (NDE) Services and the EPRI NDE Center concluded that a qualified ultrasonic examination capable of interrogating at least 90% of the required weld examination volume could not be achieved on the existing Braidwood Station and pressurizer weld configurations using the current industry approved ultrasonic methods.

Other methods of volumetric NDE (radiography) were evaluated and attempted on PWSCC samples obtained from the EPRI NDE Center without success. Based on the lack of a qualified NDE technique along with the decision to proactively mitigate these locations, application of a preemptive SWOL was determined to be the best approach for mitigation and examination.

Once a SWOL is applied, the connection is not only mitigated against PWSCC, but now can be examined using approved volumetric examination techniques qualified specifically for weld overlays. The optimum outage timeframe to support pre-installation walk downs, design, schedule, and application of the SWOL was determined to be during a ten-year Inservice Inspection (ISI) outage if such an outage was scheduled in the near-term. For the Braidwood Station units, both ten-year ISI outages were scheduled in the relative near-term (Fall 2007 and Spring 2008) and the SWOL were scheduled for these outages as per the Alloy 600 mitigation plan.

The industry butt weld safety assessment, MRP-109, "Materials Reliability Program :

Alloy 82/182 Pipe Butt Weld Safety Assessment for the US PWR Plant Designs:

Westinghouse and CE Design Plants" has concluded that there is no immediate safety concern associated with PWSCC of Alloy 82/182 butt welds based on the following:

1) The case of the longitudinal flaw is a safe situation, because the length of the flaw is limited to the width of the weld material (-2.5"), which is less than the critical flaw size for burst. (MRP-109, Sect. 7)

2) The circumferential crack growth analysis case for the Westinghouse designed safety, relief and 14" surge line butt welds is considered safe because the stress intensity factors are < 9 MPa gym, which are too low to support circumferential crack propagation (MRP-109, Section 5.5, top of page 5-33)

3) The circumferential crack growth analysis results for the W designed spray nozzle showed it would take at least 2.5 years to propagate a flaw, producing a 1 gpm leak, to a critical crack length (MRP-109 Table 5-3 and Section 7).

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2

4) A very small number of leaks/cracks given the large number of locations worldwide (MRP-139, Section 4.3).

5) Axial cracking is much more likely than circumferential cracking (MRP-109 and MRP-139, Section 1).

6) For all Alloy 82/182 pressurizer joint locations except the surge line, any leak at operating conditions would be essentially pure water/steam with little boric acid carry-over. Therefore, the potential for significant boric acid corrosion (BAC) is considered low.

7) All six pressurizer Alloy 82/182 welds underwent a bare metal visual examination during the Fall 2006 (A2R12) refuel outage with no evidence of boric acid leakage or wastage.

The circumferential crack growth analysis presented in the MRP-109 is a bounding analysis because the most restrictive loads were chosen from all 48 Westinghouse PW Rs with Alloy 600 welds. These loads included Safe Shutdown Earthquake (SSE) loads that are directly applicable to Braidwood Station. Westinghouse has confirmed that MRP-109 is a bounding analysis for the Braidwood Station pressurizer nozzle configurations.

The analyses and results reported in MRP-109 have been re-evaluated considering the pressurizer nozzle inspection results from the Fall 2006 inspections at the Wolf Creek Station. MRP 2007-003 summarizes the assessments and conclusions of the potential for PWSCC flaws with aspect ratios larger than originally evaluated in MRP-109. The conclusions reported in MRP 2007-003 re-confirm the prior conclusions from MRP-109 with respect to pressurizer nozzles that may develop full 360° circumferential part-throughwall cracks. MRP 2007-003 concludes that long shallow cracks will grow throughwall and result in leakage before rupture when applied piping bending moments are considered (MRP 2007-003 Section 12).

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Previous Inspection Results :

Braidwood Station, Units 1 and 2 As stated before, the Braidwood Station pressurizer safe-end welds listed below have had examinations performed in accordance with the earlier requirements of ASME Section XI, and have not been re-examined with PDI qualified techniques.

Notes : Vol - ultrasonic (UT) volumetric technique ; S - surface exam ; L-Wave - refracted longitudinal wave UT technique; PT - dye penetrant surface exam ; NRI - no recordable indications Table 2-4 Previous Inspection Results Braidwood Station Unit 2 As mentioned above, Braidwood Station Unit 2 has been performing BMV of all steam space pressurizer connections since the Fall of 2004. In addition, BMV examinations were performed on all Braidwood Station Unit 2 Alloy 82/182 pressurizer locations, including the surge line connections, during the previous refueling outage in the Fall of 2006 (i.e., A2R12). In all cases the examinations were acceptable with no indications of leakage or wastage.

Weld Exam Date Exam Method Exam Technique Results Identifier Vol 45° L-wave Pipe inner diameter geometry 2PZR-01-SE-01 October 1997 indications S

PT NRI 2PZR-01-SE-02 April 1996 Vol 45° L-wave NRI S

PT NRI Vol 45° L-wave NRI 2PZR-01-SE-03 October 1994 S

PT NRI Vol 45° L-wave NRI 2PZR-01-SE-04 October 1994 S

PT NRI Vol 55° L-wave Ultrasonic indications from the alloy April 1990 82/182 cladding interface 2PZR-01-SE-05 S

PT NRI Pipe inner diameter geometry Vol 45° L-wave indications April 1999 S

PT NRI Vol 55° L-wave Ultrasonic indications from the alloy April 1990 82/182 cladding interface 2PZR-01-SE-06 S

PT NRI Pipe inner diameter geometry Vol 45° L-wave indications April 1999 S

PT NRI Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Fabrication History:

Braidwood Station, Units 1 and 2 2PZR-01-SE-01 A review of the fabrication records indicated that there were four repairs performed on this weld. The repair areas were located between radiography (RT) location marks:

Location 0-1 : Length 1.5" with depth of 0.3" from ID ; Location 1-2: Length 2.5" with depth of 0.5" from OD and length 3.75" with depth of 0.5" from ID ; Location 2-3: Length 2" with depth of 0.5" from OD ; Location 3-4 : Length 2" with depth 0.3" from ID ; and Location 6-7: Length 1" with depth 0.25" from OD and length 2.5" with depth 0.625" from ID ; and Location 7-0: Length 2" with depth 0.625" from ID. Subsequent to the repairs, a final, acceptable, RT was performed.

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.

2PZR-01-SE-03 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.

2PZR-01-SE-04 A review of the fabrication records indicated that there were documented no repairs to this weld during construction.

2PZR-01-SE-05 A review of the fabrication records indicated that there were no documented repairs to this weld during construction.

2PZR-01-SE-06 A review of the fabrication records indicated that there were no documented repairs to this weld during construction.

Leak Detection Capability :

The MRP-109 report assumes that action will be taken in accordance with Braidwood Station TS 3.4.13, "RCS Operational Leakage," if the 1.0 gpm unidentified or 10 gpm identified leakage is detected.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Braidwood Station Surveillance Procedure 2BwOSR 3.4.13.1, "Unit 2 Reactor Coolant System Water Inventory Balance 72 Hour Surveillance," provides the steps necessary to determine RCS identified and unidentified leakage. The Braidwood Station surveillance along with EGC procedure ER-AP-331-1003, "RCS Leakage Monitoring and Action Plan" establish the controls and expectations for monitoring RCS leakage well below the TS required action levels. A discussion of these expectations is provided in Attachment 5.

Braidwood Station Unit 2 is evaluating enhancements to the current leakage monitoring program as well as the feasibility of installing video cameras or other on-line monitoring capability at specific Braidwood Station Unit 2 pressurizer locations. EGC will provide the details of these leakage monitoring actions to the NRC by March 31, 2007 and any on-line equipment applications for Braidwood Station Unit 2 to the NRC by May 31, 2007.

If Braidwood Station Unit 1 or Unit 2 should shut down due to excessive primary system unidentified leakage, and if the leakage cannot be confirmed to originate from a source other than the pressurizer, a BMV examination of Alloy 82/182 butt weld locations on the pressurizer will be performed to determine whether the leakage originated at those locations. EGC will maintain these contingency actions in-place until the installation of full structural weld overlays on all six pressurizer Alloy 82/182 connections : October 2007 for Unit 1 and May 2008 for Unit 2.

Water Chemistry:

Key parameters requiring control to minimize PWSCC are defined in the EPRI guidelines. The following discussion includes a description of the key parameters required for PWSCC mitigation and, a description of the Braidwood Unit 2 Chemistry program for these key parameters in relation to stress corrosion cracking (SCC) mitigation on pressurizer materials.

Dissolved Oxygen Minimizing dissolved oxygen concentration is important for SCC mitigation, although it does not singularly prevent PWSCC. Oxygen and oxidizing species are produced in a PWR core by radiolysis and from makeup water and boric acid sources. Extensive modeling has shown that dissolved hydrogen concentrations as low as 1 - 5 cubic centimeters per kilogram (cc/kg) are sufficient to effectively scavenge oxygen and oxidizing species produced in the core Pressurized Water Reactor Primary Water Chemistry Guidelines (Reference 1). Analysis documented in the EPRI Guidelines, shows that oxygen from aerated water added at full power is effectively scavenged within the first few centimeters (<10 cm) of the core. Modeling performed using a hydrogen concentration of 25 cc/kg showed complete oxygen suppression downstream of the core for cases evaluated with and without oxygen introduction from makeup sources.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Reactor coolant hydrogen concentration is a control parameter in the EPRI guidelines, with action levels established for operation outside the control band of 25 - 50 cc/kg when the reactor is critical. EGC maintains strict controls for reactor coolant hydrogen within the EPRI guidelines. The average reactor coolant hydrogen for Braidwood Station Unit 2 cycle 12 was 34.9 cc/kg, and for cycle 13 (to date) is 37.3 cc/kg. The only time in cycle 12 or cycle 13 (to date) that the reactor coolant hydrogen concentration was below the action level 1 value of 25 cc/kg was for 3.17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> in mode 2 during startup after A2R12.

The Braidwood Station Unit 2 hydrogen concentration far exceeds the values required to ensure the concentration of oxygen exiting the reactor core is zero at all times. Since the pressurizer spray line is located on the loop-4 cold leg upstream of the location where partially oxygenated reactor makeup could be introduced from the charging line, the oxygen content in reactor coolant introduced to the pressurizer is zero at all times during operation.

Dissolved Hydrogen The material under consideration for the pressurizer welds is alloy 82/182, which displays different susceptibility characteristics than steam generator tubing materials.

Materials Reliability Program (MRP-115) report 1006696 (Reference 2) documents the effects of dissolved hydrogen concentration on alloy 82/182 weld materials.

The average Braidwood Station Unit 2 hydrogen concentration (- 36 cc/kg) is close to the middle of the range for U.S. PWRs on the plot of hydrogen versus temperature, relative to the Ni/NiO transition line, shown in Reference 2, figure C-15. The conclusion from MRP-115 (Reference 2) indicates no large bias due to the effect of the dissolved hydrogen concentration on electrochemical potential, indicating that Braidwood Station operating experience is expected to fall within the range of conditions for data used to develop the MRP-115 deterministic model. Furthermore, data from Reference 2, figure C-17 suggest that actual crack growth rates for these weld materials could be less than predicted by a factor of 2 for operation at the Braidwood Station Unit 2 average hydrogen concentration.

Reactor Coolant pH In 2003, Westinghouse conducted a program that examined the crack growth rate of many primary system materials, including Alloy 182 welds, as a function of primary water pH. The results of this work were compared to MRP-115 predicted crack growth rates in Reference 2, chapter 5. Data for three distinct pH regimes: 6.9, 7.2, and 7.4, are compared to the MRP-115 curve for Alloy 182 in figure 5-2 from the referenced report.

The data show a general grouping, independent of pH (within the range studied) that on average is well below the predicted crack growth rates for a particular stress intensity factor. The average reactor coolant pH during Braidwood Station Unit 2 cycle 12 was 7.16, and for cycle 13 (to date) is 7.16. These values are bounded by the Westinghouse Crack Growth Rate data shown in Reference 2 figure 5-2. The Braidwood Station reactor coolant pH program is not expected to adversely influence crack growth rate for the pressurizer weld materials.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Contaminant Minimization Braidwood Station, Units 1 and 2 Contaminant species such as chloride, fluoride, and sulfate are known to promote stress corrosion cracking of austenitic stainless steels. At least two PWRs have experienced Alloy 600 primary-side cracking due to reduced sulfur species, which in one case is reported to have occurred during power operation. The EPRI Guidelines (Reference 1) establish action level 2 values when these parameters are > 150 parts per billion (ppb).

EGC conservatively established action level 1 values of > 50 ppb for these parameters, and goals of < 10 ppb for chloride and fluoride, and < 50 ppb for sulfate.

Braidwood Station Unit 2 cycle 12 and cycle 13 performance for these parameters is shown in the table below. There were no instances of operation outside of EGC goals for these parameters during cycle 12 or cycle 13 (to date).

Table 2-5 Contaminant Levels Braidwood Station Unit 2 Operating Cycles 12 and 13 Braidwood Station maintains reactor coolant system contaminant concentrations as low as reasonably achievable. Contaminant concentrations in makeup water are tightly controlled for recycle water by sampling every batch tank of water prior to introduction to the primary water storage tank to ensure strict chemistry limits are met. Boric acid storage tank impurities are monitored and controlled to minimize the change of impurity ingress via this pathway.

The Westinghouse Chemical and Volume Control System demineralizer design includes robust weld-screen underdrain to prevent resin leakage. Each demineralizer is also equipped with a post-strainer that normally contains a 0.1 micron absolute rated filter cartridge to prevent resin media from entering the reactor coolant, preventing primary system contaminants resulting from resin introduction.

Zinc Injection Zinc is added to the Braidwood Station Unit 2 reactor coolant system for the primary purpose of radiation field reduction. The program, as described in plant procedures, is limited to a reactor coolant zinc concentration of 5 ppb due to crud induced power shift (CIPS) risk. As indicated in the Pressurized Water Reactor Primary Water Zinc Application Guidelines (Reference 3), the effect of PWR zinc on crack initiation and growth rate (CGR) has been studied for Alloy 600, but not Alloy 82/182. Testing of the effect of zinc on Alloy 182 Crack Growth rate is currently in progress. This testing will be completed by March 2008.

Parameter Cycle 12 Average Reactor Coolant Concentration Cycle 13 (to date) Average Reactor Coolant Concentration

" -"b Chloride 3.1 2.7 Fluoride 0.9 1.3 Sulfate 0.6 0.7 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 The EPRI Zinc Guidelines (Reference 3) propose a model (Figure 2-7) for predicting the effect of zinc on PWSCC initiation, using available Alloy 600 data. Crack initiation data were used to construct an approximate model for the expected benefit as a function of zinc concentration, e.g., the linear interpolation model shown in Figure 2-7. Using this model and the Braidwood Station Unit 2 reactor coolant zinc concentration of 5 ppb, an initiation delay factor for PWSCC of 1.25 is predicted.

The conclusion is made in Reference 3 that field data, in addition to the extensive laboratory data, demonstrate that zinc is beneficial in mitigating PWSCC initiation and some crack propagation. The effectiveness of zinc on PWSCC propagation is less certain, although steam generator NDE data show that zinc has reduced crack propagation by 17% to 60%. A maximum benefit was observed at 35 ppb, although some benefit was measured at levels as low as 5 ppb. Therefore, some benefit for PWSCC mitigation is expected for Braidwood Station Unit 2. This benefit will be quantifiable when the Alloy 82/182 data becomes available.

Water Chemistry References 1. Pressurized Water Reactor Primary Water Chemistry Guidelines, EPRI 1002884, Final Report, October 2003, Volume 1.

2.

Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115),

EPRI 1006696, Final Report, November 2004.

3.

Pressurized Water Reactor Primary Water Zinc Application Guidelines, EPRI 1013420, Final Report, December 2006.

Design and Weld Stress Considerations :

Table 2-6 below provides a comparison of the Braidwood Station Unit 2 pressurizer nozzle, faulted condition bending stress to the bounding bending stress used in MRP-109 to calculate the critical crack size.

Table 2-6 Braidwood Station Unit 2 Stress Margins The stress margins greater than 1.0 provide increased critical cracks sizes for the Braidwood Station Unit 2 pressurizer nozzles.

Braidwood Station Unit 2 BR-2 Stress Margin for Critical Crack Size Nozzle BR-2 Faulted Stress MRP-109 Bounding Stress Margin (ksi)

Stress Critical Crack (ksi)

Ratio MRP-109 to BR-2 Sure 14.24 18.88 1.33 Safe /Relief 4.28 5.13 1.20 Spray 11.04 11.04 1.00 Note - Braidwood Station Unit 2 Spray nozzle is the source of the bounding stress used to determine the critical crack size.

-I Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Another source of margin was the addition of the full thermal load to the loads used to determine the critical crack sizes. Based on the significant fracture toughness and ductility of the Alloy 82/182 weld metal, the critical crack size was determined using a net section collapse failure mechanism. The thermal bending loads are self-relieving with crack opening and local plasticity of the cracked cross-section.

Including the full thermal bending moments is a significant conservatism in the determination of the critical crack sizes. The following table shows the reduction in bending moments and increased stress margin factor when the full thermal loads are removed.

Table 2-7 Braidwood Station Unit 2 Stress Margins Without Thermal Loads In addition to the stress margins for determining the critical crack size, the Alloy 182 material tensile properties for the Braidwood Station Unit 2 nozzles will also affect the critical crack margin. Review of the Braidwood Station Unit 2 certified material test reports (CMTRs) for the pressurizer nozzle Alloy 82 and 182 weld filler metal has produced the following table. The lower of the 82/182 actual material tensile properties, corrected to an operating temperature of 650°F are compared to those used to calculate the critical crack size. The increased flow stress from the CMTRs will provide increased critical crack sizes for the Braidwood Station Unit 2 nozzles.

Braidwood Station Unit 2 (BR-2)Stress Margin for Critical Crack Size Without Thermal Loads Nozzle BR-2 Faulted Stress MRP-109 Bounding Stress Margin (ksi)

Stress Critical Crack (ksi)

Ratio MRP-109 to BR-2)

Sure 12.29 18.88 1.54 Safe /Relief 3.86 5.13 1.33 Spray 6.90 11.04 1.60 Note - Braidwood Station Unit 2 Spray nozzle is the source of the bounding stress used to determine the critical crack size.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Table 2-8 Braidwood Station Unit 2 CMTR Data and Margin to Critical Crack Size These additional margins justify larger critical crack sizes for Braidwood Station Unit 2 and therefore more time for PWSCC growth from the detectable leakage size crack to the increased critical crack size as compared to the Industry critical crack size.

Low Susceptibility :

The susceptibility to PWSCC of Alloy 600/82/182 is largely a function of time at temperature when all other variables are constant. Due to the high temperature, the pressurizer is the most highly susceptible location in an operating plant. Since the pressurizers in a PWR operate at saturated conditions, all PWRs that operate at 2250 psi have a pressurizer operating temperature within a few degrees of 653° F, and can therefore be compared directly. Braidwood Station Unit 2 is a relatively young plant compared to the other US PWRs. EPRI MRP prepared a response to the NRC Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles"on PWSCC. This document lists hours of operation for each of the 69 operating PWRs (MRP-48, Table 2-1). The effective full power years (EFPY) data (resorted by EFPY from highest to lowest) from the MRP-48 survey is shown in Table 2-9.

When the Braidwood Station Unit 2 pressurizer nozzles are mitigated in May of 2008 the unit will have less EFPY of operation than Byron Unit 1 when its pressurizer nozzles were volumetrically inspected, post mitigation, in Fall 2006. No recordable indications were reported in the outer 25% of the original 82/182 weld material of the Byron Station Unit 1 pressurizer nozzles. Because the units have the same nozzle design and similar operating histories this provides additional assurance that mitigation of the Braidwood Unit 2 nozzles in May of 2008 is acceptable.

Braidwood Station Unit 2 BR-2 CMTR Data Critical Crack Size Nozzle Yield Stress Ultimate Flow Stress Flow Stress Margin (ksi)

Tensile at 650°F at 650°F Stress of f (ksi) of (ksi) ksi Sure 82/182 50.00 91.00 66.7 61.5 1.08 Safety A 82/182 44.70 87.40 62.7 61.5 1.02 Safety B 82/182 44.70 87.40 62.7 61.5 1.02 Safety C 82/182 44.70 87.40 62.7 61.5 1.02 Relief 82/182 44.70 87.40 62.7 61.5 1.02 Spray 82/182 44.70 87.40 62.7 61.5 1.02 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 Table 2-9 Comparison of EFPY of US PWRs Reported in MRP-48 Rank Unit Name EFPYs as of 2/2001 Rank Unit Name EFPYs as of 2/2001 Rank Unit Name EFPYs as of 2/2001 1

Ginna 23.9 24 North Anna 2 16.7 47 Waterford 3 12.4 2

Point Beach 1

122.9

_ __._-25 Farley 2 16.4 48 Sequoyah 2 12.1 3

Point Beach

.2 MM Cook 1 1&0 49 Byron 1 12.0 4

5 Prairie Island 1 Prairie Island 2 M

M Palisades

[ANO 2 15.9 116 50 51 Vootle 1 Sequoyahl 11.9 111.9 J 6

Kewaunee 21.6 129 Beaver Vane y 1 152 52 Catawba 2 7

Robinson 2 226 130 Crystal River 3 14.9 53 Shearon Harris 8

Oconeel 204 31 Davis-Besse 14.7 54 Bvron 2 9

Oconee 2 20.3 32 St Lude 2 14.7 55 Palo Verde 1 10 Oconee 3 20.1 33 Millstone 2 14.0 56 Palo Verde 2-111.0 11 I

Fort Calhoun 119 34 Summer 13.9 57 Salem 2 10.8 19.5 35 Callaway 13.8 58 Palo Verde 3 10.7 19.4 36 Indian Point 3 116 59

%(ogde yqg~le 2 10.4 14 Turkey Point 3 19.3 © McGuire 1 13.6 60 Braidwood 2 10.3 15 Turkey Point 4 19.0 38 San Onobe 2 1015 61 Beaver VTWeyQ 10.2 16 St. Lucie 1 18.8 39 McGuire 2 13.4 62 Braidwood 1 9.9 17 Cabert Cots 1 183 40 San CKnofre 3 13.3 63 Millstone 3 9.3 18 Farleyl 1&2 41 Cook 2 13.3 64 South Texas 1 9.2 19 ANO 1 18.0 42 Salem 1 13.1 65 South Texas 2 8.9 20 Calvert Cliffs 2 17.9 43 Diablo Canyon 1

13.1 66 Comanche Peak 1 8.9 1 21 1

North Anna 1

l 44 Diablo Canyon 2

12.8 67 Seabrook 8.6 M

Indian Point 2---

45 Wolf Creek 68 Comanche Peak 2 1 6.4 1

Rik T

69 Watts Bar 1 4.3

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Inspection Limitations:

Braidwood Station, Units 1 and 2 Per MRP-48 Table 2-1 Footnote 1 this list was current as of July 31, 2001. In A2R13 (April 2008), Braidwood Station Unit 2 will have a projected conservative EFPY of 17.1.

Braidwood Station's relative position ranking in the table should not change, since all the plants will have gained hours since the table was compiled.

Profiling of the weld and adjoining base metal contour of the six pressurizer welds was performed during A2R11 or A2R12 as recommended in MRP Letter MRP 2003-039.

Plotting of ultrasonic beam propagation on the external and internal profiles of the six pressurizer Alloy 600/82/182 butt welds with the new criteria in MRP-139 and ASME Section XI, Appendix VIII indicated that the 90% examination volume as prescribed in MRP-139 is not achievable.

Recontouring by machining would not result in any significant additional UT coverage, as the coverage limitations are imposed by nozzle tapers and outside diameter reductions that would not be remedied by external machining. Prior to the last Braidwood Station Unit 2 refueling outage (A2R12) there was no qualified inspection technology.

As stated earlier, other methods of NDE (radiography) were considered and evaluated, but their ability to detect PWSCC indications were not successfully demonstrated.

In addition to the MRP recommendation for performing BMV examinations of these welds, Braidwood Station has been performing BMV examinations of these welds on the top of the pressurizer (steam space) every outage in accordance with the Braidwood Station 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."

As stated earlier, Braidwood Station is addressing this issue by preparing to perform preemptive mitigation of all six pressurizer dissimilar welds as well as perform subsequent qualified UT inspection of the completed SWOL. The extension from the MRP-139 inspection/mitigation deadline will be approximately five months. Upon completion of the A21313 outage, all six pressurizer welds will be fully mitigated from PWSCC.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Braidwood Station, Units 1 and 2 In summary, the technical justification outlined above demonstrates that the proposed schedule meets the same objective and intent exhibited by the MRP-139 requirements.

Extending the date of UT inspections for this short period of time is considered an insignificant reduction in margin to pressure boundary leakage at these locations.

Additionally, the following points qualitatively demonstrate that the susceptibility of these welds to result in any safety significant event during the period short of deferral is very low.

1.

Prior bare metal visual and Section XI pressure tests of Braidwood Station Unit 2 pressurizer connections have not detected any evidence of leakage.

2. The operating history of Braidwood Station Unit 2 has established that this unit is among the lowest in the U. S. PWR fleet for PWSCC susceptibility.

3. A significant number of butt welds have already been inspected in the industry, and there is no evidence of widespread PWSCC of Alloy 82/182 butt welds.

4. Braidwood Station has a conservative approach for measuring and monitoring RCS leakage along with an aggressive troubleshooting and investigation procedure should RCS leakage trends increase.

5. A plan is in place to provide an inspectable weld geometry on the subject welds through the application of preemptive SWOL in May 2008.

6. Inspection of the outer 25% of the original 82/182 weld material pressurizer nozzles overlays at Byron Station Unit 1, a sister unit of Braidwood Station Unit 2, in the Fall of 2006 did not identify any recordable indications.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Byron Station, Units 1 and 2 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Byron Station, Units 1 and 2 Pressurizer Alloy 82/182 (there are no Alloy 600 product forms in the Byron Station pressurizers) butt welds at Byron Station Unit 1 have been mitigated. Details concerning the locations mitigated are provided in Table 2-2. Results of completed inspections of the full structural weld overlays (SWOL) were provided in a letter from D M. Hoots (Exelon Generation Company, LLC) to U. S. NRC, "Pressurizer Weld Overlay Examination Results Related to Byron Station Relief Request 13R-08" dated October 26, 2006. Future inspections of pressurizer butt welds at Byron Station Unit 1 will be performed in accordance with industry guidance (i.e., Materials Reliability Program (MRP) 139).

Mitigation of pressurizer Alloy 82/182 butt welds at Byron Station Unit 2 has not yet been performed, but Exelon Generation Company, LLC (EGC) intends to complete mitigation activities on these locations in April 2007. A listing of the locations to be mitigated in the Spring of 2007 for Byron Station Unit 2 are provided in Table 3-3. The listing of all Byron Station pressurizer Alloy 82/182 connections is provided in Table 3-1 below.

1. Previous Inspection Information Table 3-1 Byron Station Alloy 82/182 Connections Ultrasonic examinations on the Byron Station Unit 1 Alloy 82/182 pressurizer welds performed prior to the SWOL mitigation (i.e., Fall 2006) and for all Byron Station Unit 2 Alloy 82/182 pressurizer welds were not performed to Performance Demonstration Initiative (PDI) standards. A listing of all these pre-PDI examinations for the ten steam space connections was provided in a letter from K. R. Jury (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "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,"' dated July 27, 2004.

Unit Pressurizer Connection Listing Identifier 1

sure line connection NOZZLE S 1 RY01 S PN-01-F1 1

safe valve line connection NOZZLE C 1 RY01 S PN-04-F4 1

safe valve line connection NOZZLE B 1 RY01 S PN-05-F5 1

safety valve line connection NOZZLE A 1 RY01 S PN-06-F6 1

relief valve line connection NOZZLE D 1 RY01 S PN-03-F3 1

spray line connection NOZZLE E 1 RY01 S PN-02-F2 2

sure line connection NOZZLE S 2RY01 S PN-01-F1 2

relief valve line connection NOZZLE A 2RY01 S PN-03-F3 2

safe valve line connection NOZZLE B 2RY01 S PN-04-F4 2

safe valve line connection NOZZLE C 2RY01 S PN-05-F5 2

safe valve line connection NOZZLE D 2RY01 S PN-06-F6 2

spray line connection (NOZZLE E)

(2RY01 S) PN-02-F2 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Byron Station, Units 1 and 2 PDI qualified examinations of the Byron Station Unit 1 SWOLs were performed during the Fall 2006 refueling outage and were provided to the NRC in the above listed October 26, 2006, transmittal.

2.

Mitigation Information As described above, all Byron Station Unit 1 pressurizer connections within the scope of MRP-139 have been mitigated by SWOL. A listing is provided below.

Table 3-2 Mitigation Information Byron Station Unit 1 Nozzle MRP-139 Volumetric Inspection Mitigation Pressurizer : 1 RY01 S Requirement Met or to be Met Completed or to be Completed Function /

Susceptible Material Outage Designation Start Date Outage Designation Designation Description (MM/YYYY)

Alloy 82/182 weld 131R14 10/2006 Mitigation by Weld Surge line connection material Overlay PN-01-F1 10/2006 131R14 Alloy 82/182 weld 1311114 10/2006 Mitigation by Weld Spray line connection material Overlay PN-02-F2 10/2006 B1R14 Relief valve line Alloy 82/182 weld 1311314 10/2006 Mitigation by Weld connection material Overlay PN-03-F3 10/2006 1311114 Safety valve line Alloy 82/182 weld 131R14 10/2006 Mitigation by Weld connection material Overlay PN-04-F4 10/2006 131R14 Safety valve line Alloy 82/182 weld 131R14 10/2006 Mitigation by Weld connection material Overlay PN-05-F5 10/2006 1311314 Safety valve line Alloy 82/182 weld 131R14 10/2006 Mitigation by Weld connection material Overlay PN-06-F6 10/2006 BIR14 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Byron Station, Units 1 and 2 Byron Station Unit 2 pressurizer connections within the scope of MRP-139 will be mitigated by installing SWOL in the Spring of 2007. A listing is provided below.

3. RCS Leakage Monitoring Table 3-3 Mitigation Information Byron Station Unit 2 Byron Station Surveillance Procedures 1/2BOSR 4.13.1-1, "Unit 1/2 Reactor Coolant System Water Inventory Balance 72 Hour Surveillance" provides the steps necessary to determine reactor coolant system (RCS) identified and unidentified leakage. The Byron Station surveillance along with EGC procedure ER-AP-331-1003, "RCS Leakage Monitoring and Action Plan," establish the controls and expectations for monitoring RCS leakage well below the Tech Spec required action levels. A summary of the ER-AP-331-1003 leakage levels and corresponding actions is provided in Attachment 5..

Nozzle MRP-139 Volumetric Inspection Mitigation Pressurizer : 2RY01 S Requirement Met or to be Met Completed or to be Completed Function /

Susceptible Material Outage Designation Start Date Outage Designation Designation Description (MM/YYYY)

Alloy 82/182 weld B2R13 4/2007 Mitigation by Weld Surge line connection material Overlay PN-01-Fl 4/2007 132R13 Alloy 821182 weld B2R13 4/2007 Mitigation by Weld Spray line connection material Overlay PN-02-F2 4/2007 B2R13 Relief valve line Alloy 82/182 weld 82R13 4/2007 Mitigation by Weld connection material Overlay PN-03-F3 4/2007 132R13 Safety valve line Alloy 82/182 weld B2R13 4/2007 Mitigation by Weld connection material Overlay PN-04-F4 4/2007 B2R13 Safety valve line Alloy 82/182 weld B2R13 4/2007 Mitigation by Weld connection material Overlay PN-05-F5 4/2007 132R13 Safety valve line Alloy 82/182 weld B2R13 4/2007 Mitigation by Weld connection material Overlay PN-06-F6 4/2007 B2R13 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Byron Station, Units 1 and 2

4. Rationale for Performing Mitigation after December 31, 2007 Byron Station Unit 1 has successfully performed SWOL mitigation on all six Alloy 82/182 pressurizer connections during the Fall 2006 refueling outage. No additional justification is required. Byron Station Unit 2 will complete SWOL mitigation on all six Alloy 82/182 pressurizer connections during the Spring 2007 refueling outage (B2R13) and will be within the mandatory timeframe requirements of MRP-139; therefore no additional justification is required.

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Three Mile Nuclear Station, Unit 1

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Three Mile Island Nuclear Station (TMI) Unit 1 is a two-loop pressurized water reactor with the nuclear steam supply system designed by Babcock and Wilcox Company (B&W). The TMI Unit 1 pressurizer was fabricated by B&W and has five nozzles with Alloy 600/82/182 weld and/or safe end material that require volumetric examination in accordance with Materials Reliability Program (MRP) 139. A description of the Alloy 600/82/182 pressurizer connections and materials that require ultrasonic examination per MRP-139 follows.

1.

Previous Inspection Information Safety Relief Nozzles - No volumetric examinations were performed on the nozzle to safe end welds in the last 10 years.

Pressurizer Spray Nozzle -

Three Mile Nuclear Station, Unit 1 Table 4-1 TMI Unit 1 Alloy 600/82/182 Connections Weld PR-009BM - This weld received a UT and PT examination in October, 2003 as an expanded scope examination because of apparent PWSCC that had been identified in a pressurizer surge nozzle to safe end weld at the hot leg connection. The 2003 examination was PDI qualified. Examination coverage was calculated at 98% with the limitation due to the nozzle taper. Evaluations recently performed for AmerGen Energy have determined that 100% coverage is obtainable for future outages. No inservice degradation was identified.

Weld SP-021 BM - This weld received a UT and PT examination in October, 1999 during normal ISI activities. The UT examination was not PDI qualified. The examination was performed with refracted longitudinal wave transducers. No areas of limited coverage or inservice degradation were identified.

Pressurizer Surge Nozzle (Weld PR-021 BM) - This weld received a UT and PT examination in October, 2003 as an expanded scope examination because of apparent PWSCC that had been identified in a pressurizer surge nozzle to safe end weld at the hot leg connection. The 2003 examination was PDI qualified. Examination coverage was calculated at 100% based on plant specific qualification. This plant specific qualification was based on B&W Owners Group mock-ups that were developed prior to PDI issuing plant specific qualification guidance. No inservice degradation was identified. AmerGen Energy is currently in the process of developing a plant specific mock-up based on PDI guidance to evaluate UT examination coverage and flaw detection capability for this weld.

Pressurizer Connection Listing Number 2 1/z -inch pressure relief nozzles 3

4-inch spray nozzle 1

10-inch surge nozzle 1

Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Pressurizer Bare Metal Visual Examinations - The components identified in Section 2 below received bare metal visual examinations during the October, 2005 refueling outage. No leakage was identified during these examinations.

Additional details are provided in Table 4-2.

2.

Mitigation Information TMI Unit 1 pressurizer connections that require UT examination in accordance with MRP-139 will be mitigated or inspected by the Fall 2007. A listing is provided in Table 4-2.

3. RCS Leakage Monitoring Three Mile Nuclear Station, Unit 1 Table 4-2 Mitigation Information TMI Unit 1 TMI Unit 1 evaluates RCS leakage daily by typically performing two mass balance calculations per day. The Operations RCS leakage surveillance procedure acceptance criteria require investigation/validation of any one-leakage result that is in excess of 0.1 gallons per minute (gpm) from previous results. This verification process shall commence immediately and be concluded within one hour. Diverse means of validating RCS leakage are investigated (i.e. containment radiation monitor, containment sump level, etc.). The RCS System Engineer is notified for further evaluation. The RCS Nozzle MRP-139 Volumetric Inspection Mitigation Completed or to be Requirement Met or to be Met Completed Function /

Susceptible Material Outage Designation Start Date Outage Designation Designation Description (MM/YYYY)

Safety Relief Nozzle /

Alloy 821182 Weld None Planned None Planned Mitigation by PR-007BM Material Replacement 10/2007 T1 R17 Safety Relief Nozzle /

Alloy 82/182 Weld None Planned None Planned Mitigation by PR-006BM Material Replacement 10/2007 T1R17 Relief Valve Nozzle /

Alloy 82/182 Weld None Planned None Planned Mitigation by PR-008BM Material Replacement 10/2007 T1 R17 Spray Nozzle to Safe Alloy 600 Safe End T1 R17 10/2007 Mitigation by Weld End and Alloy 82 Weld Overlay PR-009BM Material T1R19 Spray Safe End to Alloy 600 Safe End T1 R17 10/2007 Mitigation by Weld Elbow Weld and Alloy 82 Weld Overlay SP-021 BM Material Ti R19 Surge Nozzle to Safe Alloy 82/182 Weld T1 R17 10/2007 Mitigation by Weld End Weld Material Overlay PR-021 BM T1 R19 Summary of Inspection and Mitigation Actions for Alloy 82/182 Pressurizer Butt Welds Three Mile Nuclear Station, Unit 1 System Engineer monitors RCS leakage results via a method very similar to the method described in NRC Inspection Chapter 2515 Appendix D Attachment 1 issued on December 2, 2005.

TMI Unit 1 will continue to monitor RCS leak rates on a daily basis in accordance with the Technical Specification. A summary of the leakage levels and corresponding actions is provided in Attachment 5.

If TMI Unit 1 should shut down due to excessive primary system unidentified leakage, and if the leakage cannot be confirmed to originate from a source other than the pressurizer, a bare metal visual examination of Alloy 600/82/182 butt weld locations on the pressurizer will be performed to determine whether the leakage originated at those locations identified in this response.

4.

Rationale for Performing Mitigation after December 31, 2007 TMI Unit 1 will be within the mandatory timeframe requirements of MRP-139; for either inspection or mitigation and therefore no justification is required. Additional mitigation activities (i.e., replacement with resistant materials) will be performed in a later outage.

These activities are beyond the scope of the December 31, 2007 requirements.

Summary of EGC and AmerGen Reactor Coolant System Leakage Monitoring Thresholds and Actions

Expectations Summary of EGC and AmerGen Reactor Coolant System Leakage Monitoring Thresholds and Actions The following is a summary of Reactor Coolant System (RCS) leakage actions, relevant to the pressurizer Alloy 600/82/182 connections, from the EGC and AmerGen standard corporate procedure ER-AP-331-1003, "RCS Leakage Monitoring and Action Plan."

" RCS leak rate should be calculated on a frequency required by the Technical Specifications. This should be done when the unit has been in the most stable condition practical.

If operating with a known unidentified leak, attempts to identify the source of the unidentified leak should be conducted if the unit is placed in Hot Shutdown condition at any time and corrective action consideration given to leakage reduction.

Unidentified Leakage :50.10 gallons per minute (gpm) above Baseline Monitor leakrate as defined in Plant Engineering Performance Monitoring plans including :

o monitor and trend unidentified leakage using Net Unidentified Leakrate and Average of Last Ten Net Unidentified Leakrates.

o monitor and trend Reactor Building/Containment (RB) atmospheric radiation monitor particulate channel for evidence of elevated RCS leakage.

o monitor and trend RB sump level and accumulation rate Unidentified Leakage >0.10 gpm above Baseline and :_0.40 gpm If Average of Net Unidentified Last Ten Leakrates >0.10 gpm but s0.40 gpm, then

" Document unidentified leakage in Issue Report (IR). Notify station management of status of RCS leakage at Plant Health Committee (PHC) promptly - every month.

Perform operability review of increased unidentified leakage per Exelon Nuclear Procedure LS-AA-105.

Maintain a timeline of events to document actions taken, date performed, and responsible organization.

Review unidentified leakage affect on plant operability at Plant Operations Review Committee (PORC) promptly and update in meetings - every month.

Determine location of RCS leakage by reviewing supporting plant parameters.

Visually inspect accessible locations as required.

Perform RCS Leakrate Calculation surveillance collection frequency at least once every day.

Recommend performing remote inspection of inaccessible locations if the source of leakage is not located.

Summary of EGC and AmerGen Reactor Coolant System Leakage Monitoring Thresholds and Actions Unidentified Leakage >0.40 gpm and :50.60 gpm If Average of Net Unidentified Last Ten Leakrates >0.40 gpm but :50.60 gpm, then in addition to the actions above.

Perform RCS Leakrate Calculation surveillance collection frequency at least once every day.

If remote inspection does not identify leakage location(s), then schedule a planned power reduction to acceptable power level to allow detailed inspections of inaccessible locations within three months of entering Action Level.

Unidentified Leakage >0.60 gpm and <1.0 gpm In addition to the actions above.

If supporting evidence suggests increased RCS leakage is inside RB, schedule a planned power reduction to Hot Shutdown.