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| | number = ML071030023 | | | number = ML071030023 |
| | issue date = 05/10/2007 | | | issue date = 05/10/2007 |
| | title = 10/18/2006 Summary of Conference Call with PSEG Nuclear LLC to Discuss the Fall 2006 Steam Generator Tube Inspections at Salem Nuclear Generating Station, Unit No. 2 | | | title = Summary of Conference Call with PSEG Nuclear LLC to Discuss the Fall 2006 Steam Generator Tube Inspections at Salem Nuclear Generating Station, Unit No. 2 |
| | author name = Ennis R B | | | author name = Ennis R |
| | author affiliation = NRC/NRR/ADRO/DORL/LPLB | | | author affiliation = NRC/NRR/ADRO/DORL/LPLB |
| | addressee name = Levis W | | | addressee name = Levis W |
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| | docket = 05000311 | | | docket = 05000311 |
| | license number = DPR-075 | | | license number = DPR-075 |
| | contact person = Ennis R B, NRR/DLPM, 415-1420 | | | contact person = Ennis R, NRR/DLPM, 415-1420 |
| | case reference number = TAC MD3326 | | | case reference number = TAC MD3326 |
| | package number = ML071310308 | | | package number = ML071310308 |
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| =Text= | | =Text= |
| {{#Wiki_filter:May 10, 2007Mr. William LevisSenior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038 | | {{#Wiki_filter:May 10, 2007 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038 |
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| ==SUBJECT:== | | ==SUBJECT:== |
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| ==SUMMARY== | | ==SUMMARY== |
| OF CONFERENCE CALL WITH PSEG NUCLEAR LLC TODISCUSS THE FALL 2006 STEAM GENERATOR TUBE INSPECTIONS AT SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 (TAC NO. MD3326) | | OF CONFERENCE CALL WITH PSEG NUCLEAR LLC TO DISCUSS THE FALL 2006 STEAM GENERATOR TUBE INSPECTIONS AT SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 (TAC NO. MD3326) |
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| ==Dear Mr. Levis:== | | ==Dear Mr. Levis:== |
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| On October 18, 2006, the Nuclear Regulatory Commission (NRC) staff participated in aconference call with PSEG Nuclear LLC (PSEG) to discuss the 2006 steam generator (SG) tube inspection activities taking place during refueling outage 2R15 at Salem Nuclear Generating Station, Unit No. 2. To facilitate the discussion, PSEG provided a summary of the 2R15 SG base scope tube inspection and expansion plans, and a table showing the most significant tube indications identified as of October 18, 2006. Enclosed is a brief summary of the conference call as well as the information provided by PSEG. This completes the NRC staff efforts for TAC No. MD3326. | | On October 18, 2006, the Nuclear Regulatory Commission (NRC) staff participated in a conference call with PSEG Nuclear LLC (PSEG) to discuss the 2006 steam generator (SG) tube inspection activities taking place during refueling outage 2R15 at Salem Nuclear Generating Station, Unit No. 2. To facilitate the discussion, PSEG provided a summary of the 2R15 SG base scope tube inspection and expansion plans, and a table showing the most significant tube indications identified as of October 18, 2006. Enclosed is a brief summary of the conference call as well as the information provided by PSEG. This completes the NRC staff efforts for TAC No. MD3326. |
| | | If you have any questions regarding this matter, I may be reached at 301-415-1420. |
| If you have any questions regarding this matter, I may be reached at 301-415-1420.Sincerely,/ra/Richard B. Ennis, Senior Project ManagerPlant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-311 | | Sincerely, |
| | /ra/ |
| | Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-311 |
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| ==Enclosures:== | | ==Enclosures:== |
| : 1. Summary of Conference Call | | : 1. Summary of Conference Call |
| : 2. Summary of 2R15 SG Base Scope Tube Inspection and Expansion Plans | | : 2. Summary of 2R15 SG Base Scope Tube Inspection and Expansion Plans |
| : 3. Table of Tube Indicationscc w/encls: See next page | | : 3. Table of Tube Indications cc w/encls: See next page |
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| .Package: ML071310308 Letter and Enclosure 1: ML071030023 Enclosures 2 and 3: ML071310403*By memo dated 11/7/06OFFICELPL1-2/PMLPL1-1/LACSGB/BC*LPL1-2/BC NAMEREnnisSLittleAHiserHChernoff DATE4/18/074/18/200711/07/065/10/07 Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:
| | Package: ML071310308 Letter and Enclosure 1: ML071030023 Enclosures 2 and 3: ML071310403 *By memo dated 11/7/06 OFFICE LPL1-2/PM LPL1-1/LA CSGB/BC* LPL1-2/BC NAME REnnis SLittle AHiser HChernoff DATE 4/18/07 4/18/2007 11/07/06 5/10/07 Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc: |
| Mr. Dennis WinchesterVice President - Nuclear Assessment PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. Thomas P. JoyceSite Vice President - Salem PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. George H. GellrichPlant Support Manager PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. Carl J. FrickerPlant Manager - Salem PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038Mr. James MallonManager - Licensing 200 Exelon Way, KSA 3-E Kennett Square, PA 19348Mr. Steven MannonManager - Regulatory Assurance P.O. Box 236 Hancocks Bridge, NJ 08038Jeffrie J. Keenan, EsquirePSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038Township ClerkLower Alloways Creek Township Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038Mr. Paul Bauldauf, P.E., Asst. DirectorRadiation Protection Programs NJ Department of Environmental Protection and Energy | | Mr. Dennis Winchester Mr. Paul Bauldauf, P.E., Asst. Director Vice President - Nuclear Assessment Radiation Protection Programs PSEG Nuclear NJ Department of Environmental P.O. Box 236 Protection and Energy Hancocks Bridge, NJ 08038 CN 415 Trenton, NJ 08625-0415 Mr. Thomas P. Joyce Site Vice President - Salem Mr. Brian Beam PSEG Nuclear Board of Public Utilities P.O. Box 236 2 Gateway Center, Tenth Floor Hancocks Bridge, NJ 08038 Newark, NJ 07102 Mr. George H. Gellrich Regional Administrator, Region I Plant Support Manager U.S. Nuclear Regulatory Commission PSEG Nuclear 475 Allendale Road P.O. Box 236 King of Prussia, PA 19406 Hancocks Bridge, NJ 08038 Senior Resident Inspector Mr. Carl J. Fricker Salem Nuclear Generating Station Plant Manager - Salem U.S. Nuclear Regulatory Commission PSEG Nuclear - N21 Drawer 0509 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Mr. James Mallon Manager - Licensing 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Mr. Steven Mannon Manager - Regulatory Assurance P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Township Clerk Lower Alloways Creek Township Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 |
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| CN 415 Trenton, NJ 08625-0415Mr. Brian BeamBoard of Public Utilities 2 Gateway Center, Tenth Floor Newark, NJ 07102Regional Administrator, Region IU.S. Nuclear Regulatory Commission
| | ==SUMMARY== |
| | | OF CONFERENCE CALL WITH PSEG NUCLEAR LLC FALL 2006 STEAM GENERATOR TUBE INSPECTIONS SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 DOCKET NO. 50-311 On October 18, 2006, the Nuclear Regulatory Commission (NRC) staff participated in a conference call with PSEG Nuclear LLC (PSEG or the licensee) to discuss the 2006 steam generator (SG) tube inspection activities taking place during refueling outage 2R15 at Salem Nuclear Generating Station, Unit No. 2 (Salem Unit 2). |
| 475 Allendale Road King of Prussia, PA 19406Senior Resident InspectorSalem Nuclear Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038
| | Salem Unit 2 has four Westinghouse Model 51 SGs. Each SG contains approximately 3400 mill-annealed Alloy 600 tubes. Each tube has a nominal outside diameter of 0.875-inch and a nominal wall thickness of 0.050-inch. The tubes were explosively expanded (WEXTEX) at both ends for the full length of the tubesheet and are supported by a number of drilled-hole carbon steel tube supports. The row 1 tubes were preventatively plugged. The SGs are scheduled to be replaced during the next refueling outage in spring 2008 (2R16). |
| | To facilitate the discussion, PSEG provided a summary of the 2R15 SG base scope tube inspection and expansion plans, and a table showing the most significant tube indications identified as of October 18, 2006 (Enclosures 2 and 3). In addition to the written material provided by the licensee, the following additional clarifying information was discussed during the conference call. |
| | The licensee took two exceptions to the Pressurized Water Reactor (PWR) SG Examination Guidelines. The first exception was to calibrate the rotating coil with a 20% inside diameter (ID) axial and circumferential notch rather than with a 40% ID notch. An analysis was performed using data from a sister plant that has both the 20% and 40% throughwall notches in its calibration standard to demonstrate that the technique used at Salem Unit 2 is essentially equivalent to that recommended in the guidelines. The second exception was that analysts were not required to review overcalls. These two exceptions have been in place since 2R13. |
| | At the time of the conference call, the licensee had acquired 50% to 60% of the data, and had analyzed approximately 32% to 50%. Bobbin coil inspections were almost completed and the licensee had completed about 15% of rotating pancake coil (RPC) inspections in the tubesheet region of one SG. The RPC inspections of the U-bend region of the row 9 and 10 tubes were almost complete, and no indications had been detected. At the time of the call, there were a lower number of bobbin indications (I-codes) than had been observed in previous inspections. |
| | Most of the I-codes identified are manufacturing burnish marks (MBMs). During the past two inspections, 100% of the MBMs were inspected with a rotating probe. The licensee performed a historical review of the bobbin data associated with these MBMs to determine if there was any significant voltage change, but none was identified. In addition, a historical review of the RPC data for these MBMs was performed to confirm none of the indications are flaw-like. |
| | Enclosure 1 |
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| ==SUMMARY==
| | A list of the most significant indications identified as of October 18, 2006, is shown in . In this table, there are two cold leg thinning indications: one with a 58% |
| OF CONFERENCE CALL WITH PSEG NUCLEAR LLCFALL 2006 STEAM GENERATOR TUBE INSPECTIONSSALEM NUCLEAR GENERATING STATION, UNIT NO. 2DOCKET NO. 50-311On October 18, 2006, the Nuclear Regulatory Commission (NRC) staff participated in aconference call with PSEG Nuclear LLC (PSEG or the licensee) to discuss the 2006 steam generator (SG) tube inspection activities taking place during refueling outage 2R15 at Salem Nuclear Generating Station, Unit No. 2 (Salem Unit 2). Salem Unit 2 has four Westinghouse Model 51 SGs. Each SG contains approximately3400 mill-annealed Alloy 600 tubes. Each tube has a nominal outside diameter of 0.875-inch and a nominal wall thickness of 0.050-inch. The tubes were explosively expanded (WEXTEX) at both ends for the full length of the tubesheet and are supported by a number of drilled-hole carbon steel tube supports. The row 1 tubes were preventatively plugged. The SGs are scheduled to be replaced during the next refueling outage in spring 2008 (2R16).To facilitate the discussion, PSEG provided a summary of the 2R15 SG base scope tubeinspection and expansion plans, and a table showing the most significant tube indications identified as of October 18, 2006 (Enclosures 2 and 3). In addition to the written material provided by the licensee, the following additional clarifying information was discussed during the conference call. The licensee took two exceptions to the Pressurized Water Reactor (PWR) SG ExaminationGuidelines. The first exception was to calibrate the rotating coil with a 20% inside diameter (ID) axial and circumferential notch rather than with a 40% ID notch. An analysis was performed using data from a sister plant that has both the 20% and 40% throughwall notches in its calibration standard to demonstrate that the technique used at Salem Unit 2 is essentially equivalent to that recommended in the guidelines. The second exception was that analysts were not required to review overcalls. These two exceptions have been in place since 2R13.At the time of the conference call, the licensee had acquired 50% to 60% of the data, and hadanalyzed approximately 32% to 50%. Bobbin coil inspections were almost completed and the licensee had completed about 15% of rotating pancake coil (RPC) inspections in the tubesheet region of one SG. The RPC inspections of the U-bend region of the row 9 and 10 tubes were almost complete, and no indications had been detected. At the time of the call, there were a lower number of bobbin indications (I-codes) than had been observed in previous inspections.
| | throughwall depth and a voltage of 3.75 volts and the other one with a 64% throughwall depth and a voltage of 0.95 volts. A historical review of these indications showed that, during the previous outage, the 58% throughwall indication measured 39% throughwall and had a voltage of 2.3 volts, while the 64% throughwall indication measured 34% throughwall and had a voltage of 1.15 volts. The licensee indicated that it is not uncommon to have such changes in voltage and depth given the low voltages for these indications. Both indications have adequate tube integrity. |
| Most of the I-codes identified are manufacturing burnish marks (MBMs). During the past two inspections, 100% of the MBMs were inspected with a rotating probe. The licensee performed a historical review of the bobbin data associated with these MBMs to determine if there was any significant voltage change, but none was identified. In addition, a historical review of the RPC data for these MBMs was performed to confirm none of the indications are flaw-like. A list of the most significant indications identified as of October 18, 2006, is shown inEnclosure 3. In this table, there are two cold leg thinning indications: one with a 58%
| | The licensee plans to plug all wear indications caused by loose parts in addition to any stress-corrosion cracking (SCC) degradation. |
| throughwall depth and a voltage of 3.75 volts and the other one with a 64% throughwall depth and a voltage of 0.95 volts. A historical review of these indications showed that, during the previous outage, the 58% throughwall indication measured 39% throughwall and had a voltage of 2.3 volts, while the 64% throughwall indication measured 34% throughwall and had a voltage of 1.15 volts. The licensee indicated that it is not uncommon to have such changes in voltage and depth given the low voltages for these indications. Both indications have adequate tube integrity. The licensee plans to plug all wear indications caused by loose parts in addition to any stress-corrosion cracking (SCC) degradation.At the time of this call, no in-situ pressure tests or tube pulls were planned, however, ifnecessary, Electric Power Research Institute (EPRI) in-situ guidelines will be followed.At the time of the conference call, inspections to detect loose parts were ongoing. Eddy currentinspections identified some possible loose part indications. These indications were scheduledto be investigated visually. In addition, the licensee identified the presence of one part by eddy current that had been previously left in service. No wear has been associated with any of the possible loose part indications identified.Historically, the licensee has detected primary water stress-corrosion cracking (PWSCC) nearthe top of the tubesheet, PWSCC and outside diameter stress corrosion cracking (ODSCC) at the tube support plate (TSP) elevations, ODSCC at the top of the tubesheet, ODSCC at the TSP elevations (only a few indications primarily at the 4 th hot leg TSP and not associated withdents), and SCC (typically PWSCC) at dents (typically at the 1 st hot leg tube support: five weredetected at the 1 st hot leg tube support during the last outage). No degradation has beenidentified in dings.The data quality is monitored per revision 6 of the EPRI PWR SG Examination Guidelines.The data quality is similar to past inspections.On October 20, 2006, another conference call was conducted with the licensee to discuss anaxially-oriented ODSCC indication found at a 6.83-volt ding in SG 24. The indication was 1.55 inches above the top of the tubesheet on the hot leg side of the SG. The indication was reported to be 0.26 inch long with a voltage of 0.47 volts. The licensee was expanding the scope of their inspection consistent with the EPRI PWR SG Examination Guidelines.Based on the information provided, the NRC staff did not identify any technical issues thatwarranted follow-up action at this time.}} | | At the time of this call, no in-situ pressure tests or tube pulls were planned, however, if necessary, Electric Power Research Institute (EPRI) in-situ guidelines will be followed. |
| | At the time of the conference call, inspections to detect loose parts were ongoing. Eddy current inspections identified some possible loose part indications. These indications were scheduled to be investigated visually. In addition, the licensee identified the presence of one part by eddy current that had been previously left in service. No wear has been associated with any of the possible loose part indications identified. |
| | Historically, the licensee has detected primary water stress-corrosion cracking (PWSCC) near the top of the tubesheet, PWSCC and outside diameter stress corrosion cracking (ODSCC) at the tube support plate (TSP) elevations, ODSCC at the top of the tubesheet, ODSCC at the TSP elevations (only a few indications primarily at the 4th hot leg TSP and not associated with dents), and SCC (typically PWSCC) at dents (typically at the 1st hot leg tube support: five were detected at the 1st hot leg tube support during the last outage). No degradation has been identified in dings. |
| | The data quality is monitored per revision 6 of the EPRI PWR SG Examination Guidelines. |
| | The data quality is similar to past inspections. |
| | On October 20, 2006, another conference call was conducted with the licensee to discuss an axially-oriented ODSCC indication found at a 6.83-volt ding in SG 24. The indication was 1.55 inches above the top of the tubesheet on the hot leg side of the SG. The indication was reported to be 0.26 inch long with a voltage of 0.47 volts. The licensee was expanding the scope of their inspection consistent with the EPRI PWR SG Examination Guidelines. |
| | Based on the information provided, the NRC staff did not identify any technical issues that warranted follow-up action at this time.}} |
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MONTHYEARML0710300232007-05-10010 May 2007 Summary of Conference Call with PSEG Nuclear LLC to Discuss the Fall 2006 Steam Generator Tube Inspections at Salem Nuclear Generating Station, Unit No. 2 Project stage: Other 2007-05-10
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[Table view] Category:Meeting Summary
MONTHYEARML24155A1992024-06-0303 June 2024 Annual Assessment Meeting for Pa/Md/Ny/Nj Nuclear Power Plants - NRC Meeting Summary ML23276A6242023-10-12012 October 2023 Summary of the Closed Meeting with PSEG Nuclear LLC and Westinghouse Electric Company to Discuss Cobalt-60 Production at Salem Nuclear Generating Station ML23179A0012023-06-28028 June 2023 6/7/2023 - Annual Assessment Meeting for Pa/Md/Ny/Nj Nuclear Power Plants - Meeting Summary ML23146A1012023-05-31031 May 2023 Summary of May 25, 2023, Pre-Submittal Public Meeting with PSEG Nuclear LLC to Discuss Reduction of the Site'S Exclusion Area Boundary ML22187A1462022-07-0606 July 2022 Joint Annual Assessment Meeting for Pa/Md/Nj/Ny Nuclear Power Plants - Meeting Summary ML22153A4492022-06-14014 June 2022 Summary of Pre-Submittal Public Meeting with PSEG Nuclear LLC to Extend Emergency Diesel Generator Allowed Outage Time ML21201A0702021-08-0505 August 2021 Summary of July 20, 2021, Meeting with the National Marine Fisheries Service Regarding Incidental Take of Kemps Ridley Sea Turtles at Salem Nuclear Generating Station, Units 1 and 2 ML21064A0962021-03-0909 March 2021 Summary of 3/4/2021 Pre-Submittal Public Meeting with PSEG Nuclear LLC to Discuss Adoption of ASME Section XI Code Cases N-878, N-879, and N-880 at Hope Creek Generating Station and Salem Nuclear Generating Station ML20204A8422020-07-28028 July 2020 Summary of July 20, 2020, Pre-Submittal Public Meeting with PSEG Nuclear LLC to Discuss Proposed Relief Request for Pressurizer Weld Interval Extension ML20125A1652020-05-0404 May 2020 Summary of April 23, 2020, Teleconference with NMFS Regarding Incidental Take of Atlantic Sturgeon at Salem Nuclear Generating Station ML20113F0892020-04-27027 April 2020 Summary of April 15, 2020, Pre-Submittal Meeting with PSEG Nuclear LLC to Discuss Proposed License Amendment on Leak Before Break Analyses ML20108F0652020-04-17017 April 2020 Summary of April 15, 2020, Teleconference with PSEG Nuclear, LLC Regarding Incidental Take of Atlantic Sturgeon at Salem ML19109A1282019-04-16016 April 2019 and Hope Creek Generating Station: Summary of Public Webinar Meeting on April 10, 2019 ML18156A4242018-06-0505 June 2018 Annual Assessment Meeting, 5/30/18 Meeting Summary ML18044A9942018-02-26026 February 2018 Summary of Meeting with PSEG Nuclear LLC Regarding Planned Submittal of License Amendment Request to Revise Technical Specifications Regarding Extending the Allowed Outage Time for the Inverters ML17332A6032017-12-0404 December 2017 Summary of October 27, 2017, Conference Call with PSEG Nuclear LLC Fall 2017 Steam Generator Inspections ML17206A0282017-08-0707 August 2017 Summary of 7/19/17 Meeting with PSEG Planned Submittal of License Amendment Request to Revise Technical Specifications for Reactor Trip and Engineered Safety Feature Actuation System Instrumentation for Salem, Units 1 and 2 (CAC Nos. MF9900 ML17102A7772017-04-10010 April 2017 Summary of Annual Assessment Meeting with PSEG ML16347A3932017-01-0505 January 2017 Summary of Teleconference with PSEG Nuclear LLC Regarding Planned License Amendment Request Regarding Change to the Salem Nuclear Generating Station-Hope Creek Generating Station Emergency Action Levels (MF8875, MF8876, and MF8877) ML16123A1322016-05-17017 May 2016 Summary of Public Pre-Application Meeting Held on March 30, 2016, with PSEG LLC to Discuss License Amendment Request Containment Fan Coil Unit Allowed Outage Time at Salem Nuclear Generating Station, Unit Nos. 1 and 2 ML16096A2222016-04-0505 April 2016 March 30, 2016, Summary of Annual Assessment Meeting with PSEG - Salem Nuclear Generating Station & Hope Creek Generating Station ML16035A4802016-02-12012 February 2016 Jsc Meeting Summary Attachment ML15202A2432015-07-29029 July 2015 June 24, 2015, Summary of Pre-Submittal Meeting with PSEG Nuclear LLC on Upgrades and Maintenance to the Chilled Water System at Salem Nuclear Generating Station, Unit Nos. 1 and 2 ML15104A1492015-04-14014 April 2015 Hope, Summary of Annual Assessment Meeting with PSEG ML14167A1592014-09-0202 September 2014 Summary of Category 1 Public Meeting with Salem Nuclear Generating Station, Units 1 and 2 to Discuss Seismic Hazard Reevaluations Associated with Implementation of Japan Lessons-Learned Near-Term Task Force Recommendation 2.1 Seismic ML14100A0242014-04-10010 April 2014 Salem and Hope Creek - Summary of Annual Assessment Meeting with PSEG ML13109A2352013-05-10010 May 2013 Summary of Conference Call Regarding the Fall 2012 Steam Generator Tube Inspections at Salem Nuclear Generating Station, Unit No. 2 ML13099A3242013-04-0909 April 2013 Salem & Hope Creek - Summary of Annual Assessment Meeting with PSEG ML13078A4482013-04-0505 April 2013 Summary of February 28, 2013, Public Meeting with PSEG Final Supplemental Response to Generic Letter 2004-02 for Salem Nuclear Generating Station, Units 1 and 2 ML12144A3312012-05-23023 May 2012 Summary of Meeting at the PSEG to Discuss the Nrc'S Assessment of Pseg'S Safety Performance at Salem & Hope Creek for 2011 ML1129700652011-10-21021 October 2011 Summary of PSEG Drop-in Visit, 9-28-2011 and Slides ML1125200382011-09-19019 September 2011 Summary of, Meeting with PSEG Nuclear LLC to Discuss Proposed License Amendment Regarding Reactor Coolant System Activity for Salem Nuclear Generating Station, Unit Nos. 1 and 2 ML11139A3462011-05-26026 May 2011 Summary of Telephone Conference Call Held on May 10, 2011, Between the NRC and Nuclear, LLC, Concerning a Draft Response to a Draft RAI Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, (Salem) License Renewal Application ML1114002132011-05-20020 May 2011 Summary of Annual Assessment Meeting for Salem and Hope Creek ML1107608982011-04-0505 April 2011 January 12, 2011, Summary of Telephone Conference Call Held Between the U.S. Nuclear Regulatory Commission and PSEG Nuclear, Llx, Concerning Audit of the Westems Software Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, LR ML1107606042011-04-0505 April 2011 January 25, 2011, Summary of Telephone Conference Call Held Btw the USNRC and PSEG Nuclear, LLC, Concerning Information on Buried Piping Inspections at the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1107007082011-04-0505 April 2011 February 17, 2011, Summary of Telephone Conference Call Held Between NRC and PSEG, Concerning Information on the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1036307352011-02-0909 February 2011 Summary of Telephone Conference Call Held on September 16, 2010, Between the U.S. Nuclear Regulatory Commission and PSEG Nuclear LLC, Concerning Questions Pertaining to the Salem Nuclear Generating Station, Units 1 & 2, License Renewal App ML1035005662011-02-0909 February 2011 Summary of Telephone Conference Call Held on 11/18/10 Between the USNRC and PSEG Nuclear, LLC, Concerning Draft Response to the Salem Nuclear Generating Station, Units 1 and 2, Request for Additional Information, and Concerning NRC Open Ite ML1100501562011-01-26026 January 2011 Summary of Telephone Conference Call Held on September 27, 2010, Between the U.S. Nuclear Regulatory Commission and PSEG Nuclear, LLC, Concerning Questions Pertaining to the Salem Nuclear Generating Station, Units 1 and 2 ML1035606562011-01-26026 January 2011 12/13/10 Summary of Telephone Conference Call Held on 12/13/10 Btw the USNRC and PSEG Nuclear, LLC, Concerning Draft Requests for Additional Information Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Appl ML1035401842011-01-0707 January 2011 10/08/10 Summary of Telephone Conference Call Held Btw the USNRC and PSEG Nuclear, LLC, Concerning Questions Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1032102502010-12-13013 December 2010 10/22/10 Summary of Telephone Conference Call Held Btw the USNRC and PSEG Nuclear LLC Concerning Questions Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1026400612010-10-14014 October 2010 09/09/2010-Summary of Telephone Conference Call Between the USNRC and PSEG Nuclear, LLC, Concerning Questions Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1024600952010-10-14014 October 2010 Summary of Telephone Conference Call Held Between Us NRC and PSEG Nuclear, LLC Concerning Questions Pertaining to the Salem Nuclear Generation Station Units 1&2 LRA ML1023200572010-09-0303 September 2010 8/16/10 Summary of Telephone Conference Call Held Between the USNRC and PSEG Nuclear LLC Concerning a D-RAI Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1022900442010-08-31031 August 2010 Summary of Conference Call Regarding Spring 2010 Steam Generator Tube Inspections at Salem Nuclear Generating Station, Unit No. 1 ML1022305122010-08-18018 August 2010 08/04/10 Summary of Telephone Conference Call Held on Between NRC and PSEG Nuclear, LLC, Concerning Drais Pertaining to the Salem Nuclear Generating Station, Units 1 and 2, License Renewal Application ML1022200122010-08-13013 August 2010 Summary of Telephone Conf Call Held on 7/29/10, Between the USNRC and PSEG Nuclear LLC, Concerning Follow-Up Questions Pertaining to the Salem Nuclear Generating Station, Units 1 & 2, and Hope Creek Generating Station License Renewal Enviro ML1018004772010-07-14014 July 2010 Summary of Category 1 Public Meeting with PSEG Nuclear LLC, Regarding Response to Generic Letter 2004-02 for Salem Nuclear Generating Station, Units 1 and 2 2024-06-03
[Table view] |
Text
May 10, 2007 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
SUMMARY
OF CONFERENCE CALL WITH PSEG NUCLEAR LLC TO DISCUSS THE FALL 2006 STEAM GENERATOR TUBE INSPECTIONS AT SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 (TAC NO. MD3326)
Dear Mr. Levis:
On October 18, 2006, the Nuclear Regulatory Commission (NRC) staff participated in a conference call with PSEG Nuclear LLC (PSEG) to discuss the 2006 steam generator (SG) tube inspection activities taking place during refueling outage 2R15 at Salem Nuclear Generating Station, Unit No. 2. To facilitate the discussion, PSEG provided a summary of the 2R15 SG base scope tube inspection and expansion plans, and a table showing the most significant tube indications identified as of October 18, 2006. Enclosed is a brief summary of the conference call as well as the information provided by PSEG. This completes the NRC staff efforts for TAC No. MD3326.
If you have any questions regarding this matter, I may be reached at 301-415-1420.
Sincerely,
/ra/
Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-311
Enclosures:
- 1. Summary of Conference Call
- 2. Summary of 2R15 SG Base Scope Tube Inspection and Expansion Plans
- 3. Table of Tube Indications cc w/encls: See next page
Package: ML071310308 Letter and Enclosure 1: ML071030023 Enclosures 2 and 3: ML071310403 *By memo dated 11/7/06 OFFICE LPL1-2/PM LPL1-1/LA CSGB/BC* LPL1-2/BC NAME REnnis SLittle AHiser HChernoff DATE 4/18/07 4/18/2007 11/07/06 5/10/07 Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:
Mr. Dennis Winchester Mr. Paul Bauldauf, P.E., Asst. Director Vice President - Nuclear Assessment Radiation Protection Programs PSEG Nuclear NJ Department of Environmental P.O. Box 236 Protection and Energy Hancocks Bridge, NJ 08038 CN 415 Trenton, NJ 08625-0415 Mr. Thomas P. Joyce Site Vice President - Salem Mr. Brian Beam PSEG Nuclear Board of Public Utilities P.O. Box 236 2 Gateway Center, Tenth Floor Hancocks Bridge, NJ 08038 Newark, NJ 07102 Mr. George H. Gellrich Regional Administrator, Region I Plant Support Manager U.S. Nuclear Regulatory Commission PSEG Nuclear 475 Allendale Road P.O. Box 236 King of Prussia, PA 19406 Hancocks Bridge, NJ 08038 Senior Resident Inspector Mr. Carl J. Fricker Salem Nuclear Generating Station Plant Manager - Salem U.S. Nuclear Regulatory Commission PSEG Nuclear - N21 Drawer 0509 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Mr. James Mallon Manager - Licensing 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Mr. Steven Mannon Manager - Regulatory Assurance P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Township Clerk Lower Alloways Creek Township Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038
SUMMARY
OF CONFERENCE CALL WITH PSEG NUCLEAR LLC FALL 2006 STEAM GENERATOR TUBE INSPECTIONS SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 DOCKET NO. 50-311 On October 18, 2006, the Nuclear Regulatory Commission (NRC) staff participated in a conference call with PSEG Nuclear LLC (PSEG or the licensee) to discuss the 2006 steam generator (SG) tube inspection activities taking place during refueling outage 2R15 at Salem Nuclear Generating Station, Unit No. 2 (Salem Unit 2).
Salem Unit 2 has four Westinghouse Model 51 SGs. Each SG contains approximately 3400 mill-annealed Alloy 600 tubes. Each tube has a nominal outside diameter of 0.875-inch and a nominal wall thickness of 0.050-inch. The tubes were explosively expanded (WEXTEX) at both ends for the full length of the tubesheet and are supported by a number of drilled-hole carbon steel tube supports. The row 1 tubes were preventatively plugged. The SGs are scheduled to be replaced during the next refueling outage in spring 2008 (2R16).
To facilitate the discussion, PSEG provided a summary of the 2R15 SG base scope tube inspection and expansion plans, and a table showing the most significant tube indications identified as of October 18, 2006 (Enclosures 2 and 3). In addition to the written material provided by the licensee, the following additional clarifying information was discussed during the conference call.
The licensee took two exceptions to the Pressurized Water Reactor (PWR) SG Examination Guidelines. The first exception was to calibrate the rotating coil with a 20% inside diameter (ID) axial and circumferential notch rather than with a 40% ID notch. An analysis was performed using data from a sister plant that has both the 20% and 40% throughwall notches in its calibration standard to demonstrate that the technique used at Salem Unit 2 is essentially equivalent to that recommended in the guidelines. The second exception was that analysts were not required to review overcalls. These two exceptions have been in place since 2R13.
At the time of the conference call, the licensee had acquired 50% to 60% of the data, and had analyzed approximately 32% to 50%. Bobbin coil inspections were almost completed and the licensee had completed about 15% of rotating pancake coil (RPC) inspections in the tubesheet region of one SG. The RPC inspections of the U-bend region of the row 9 and 10 tubes were almost complete, and no indications had been detected. At the time of the call, there were a lower number of bobbin indications (I-codes) than had been observed in previous inspections.
Most of the I-codes identified are manufacturing burnish marks (MBMs). During the past two inspections, 100% of the MBMs were inspected with a rotating probe. The licensee performed a historical review of the bobbin data associated with these MBMs to determine if there was any significant voltage change, but none was identified. In addition, a historical review of the RPC data for these MBMs was performed to confirm none of the indications are flaw-like.
Enclosure 1
A list of the most significant indications identified as of October 18, 2006, is shown in . In this table, there are two cold leg thinning indications: one with a 58%
throughwall depth and a voltage of 3.75 volts and the other one with a 64% throughwall depth and a voltage of 0.95 volts. A historical review of these indications showed that, during the previous outage, the 58% throughwall indication measured 39% throughwall and had a voltage of 2.3 volts, while the 64% throughwall indication measured 34% throughwall and had a voltage of 1.15 volts. The licensee indicated that it is not uncommon to have such changes in voltage and depth given the low voltages for these indications. Both indications have adequate tube integrity.
The licensee plans to plug all wear indications caused by loose parts in addition to any stress-corrosion cracking (SCC) degradation.
At the time of this call, no in-situ pressure tests or tube pulls were planned, however, if necessary, Electric Power Research Institute (EPRI) in-situ guidelines will be followed.
At the time of the conference call, inspections to detect loose parts were ongoing. Eddy current inspections identified some possible loose part indications. These indications were scheduled to be investigated visually. In addition, the licensee identified the presence of one part by eddy current that had been previously left in service. No wear has been associated with any of the possible loose part indications identified.
Historically, the licensee has detected primary water stress-corrosion cracking (PWSCC) near the top of the tubesheet, PWSCC and outside diameter stress corrosion cracking (ODSCC) at the tube support plate (TSP) elevations, ODSCC at the top of the tubesheet, ODSCC at the TSP elevations (only a few indications primarily at the 4th hot leg TSP and not associated with dents), and SCC (typically PWSCC) at dents (typically at the 1st hot leg tube support: five were detected at the 1st hot leg tube support during the last outage). No degradation has been identified in dings.
The data quality is monitored per revision 6 of the EPRI PWR SG Examination Guidelines.
The data quality is similar to past inspections.
On October 20, 2006, another conference call was conducted with the licensee to discuss an axially-oriented ODSCC indication found at a 6.83-volt ding in SG 24. The indication was 1.55 inches above the top of the tubesheet on the hot leg side of the SG. The indication was reported to be 0.26 inch long with a voltage of 0.47 volts. The licensee was expanding the scope of their inspection consistent with the EPRI PWR SG Examination Guidelines.
Based on the information provided, the NRC staff did not identify any technical issues that warranted follow-up action at this time.