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| | issue date = 03/18/2008 | | | issue date = 03/18/2008 |
| | title = Transmittal of Technical Specification 5.6.6.1 - Steam Generator Inspection Report | | | title = Transmittal of Technical Specification 5.6.6.1 - Steam Generator Inspection Report |
| | author name = Sena P P | | | author name = Sena P |
| | author affiliation = FirstEnergy Nuclear Operating Co | | | author affiliation = FirstEnergy Nuclear Operating Co |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:FENOC FirstEnergy Nuclear Operating Company Peter P. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 March 18, 2008 L-08-072 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 | | {{#Wiki_filter:FENOC FirstEnergyNuclear OperatingCompany Peter P. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 March 18, 2008 L-08-072 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Technical Specification 5.6.6.1 -Steam Generator Inspection Report In accordance with Beaver Valley Power Station Unit No. 1 Technical Specification (TS) 5.6.6.1, information related to steam generator inspections is provided in Attachment 1.There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager -FENOC Fleet Licensing, at 330-761-6071. | | |
| | Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Technical Specification 5.6.6.1 - Steam Generator Inspection Report In accordance with Beaver Valley Power Station Unit No. 1 Technical Specification (TS) 5.6.6.1, information related to steam generator inspections is provided in Attachment 1. |
| | There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - |
| | FENOC Fleet Licensing, at 330-761-6071. |
| Sincerely, Peter P. Sena III Attachments: | | Sincerely, Peter P. Sena III Attachments: |
| : 1. Technical Specification 5.6.6.1 -Unit 1 SG Tube Inspection Report cc: Mr. S. J. Collins, NRC Region I Administrator Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager Mr. D. J. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP) | | : 1. Technical Specification 5.6.6.1 - Unit 1 SG Tube Inspection Report cc: Mr. S. J. Collins, NRC Region I Administrator Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager Mr. D. J. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP) |
| ATTACHMENT 1 L-08-072 Beaver Valley Power Station Unit No. 1 Technical Specification 5.6.6.1 -Unit 1 SG Tube Inspection Report Page 1 of 7 The following information satisfies the reporting requirement of Beaver Valley Power Station Technical Specification (TS) 5.6.6.1. Refer to the tables on Page 7 of this attachment which define abbreviations, codes and other terms used throughout this report.Information required by Technical Specification 5.6.6.1 Technical Specification 5.6.6.1 requires reporting within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.5.1, Unit 1 Steam Generator (SG) Program. Information required by TS 5.6.6.1, Items a through h is provided below for the maintenance and refueling outage in the fall of 2007 (1 R1 8).a. The scope of inspections performed on each SG Primary side inspection scope involved the following: | | |
| * 100% full length bobbin coil probe examination (Row*3 through Row 47)* 100% bobbin coil probe examination of the hot and cold leg straight sections in Row 1 and 2 up to and including the top support plate* 100% Plus Point probe inspection of the Row 1 and 2 U-bends from the top support plate on the hot leg to the top support plate on the cold leg* 100% Plus Point probe inspection of over-expansions and bulges contained within or just above the hot leg tubesheet (These were previously reported during the preservice examination and are located 10.0" or higher from the hot leg tube end)* 100% Plus Point probe inspection of bobbin coil indications that could possibly represent degradation (I-codes)* Plus Point probe inspection of the top-of-tubesheet periphery (3 tubes deep in any row or column) and the blowdown pipe region (2 tubes deep on either side of the blowdown pipe)* Plus Point probe inspection of any ambiguous bobbin coil signals P "As- left" video inspection of the hot and cold leg channel heads Secondary side inspection scope involved the following: | | ATTACHMENT 1 L-08-072 Beaver Valley Power Station Unit No. 1 Technical Specification 5.6.6.1 - Unit 1 SG Tube Inspection Report Page 1 of 7 The following information satisfies the reporting requirement of Beaver Valley Power Station Technical Specification (TS) 5.6.6.1. Refer to the tables on Page 7 of this attachment which define abbreviations, codes and other terms used throughout this report. |
| * Secondary side visual inspections of the top-of-tubesheet annulus, blowdown pipe region and random in-bundle locations" In-bundle visual inspections of possible loose part locations identified from the analysis of the eddy current data Attachment 1 L-08-072 Page 2 of 7 b. Active degradation mechanisms found No active degradation mechanisms were reported during the 1 R1 8 SG tube examinations. | | Information required by Technical Specification 5.6.6.1 Technical Specification 5.6.6.1 requires reporting within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.5.1, Unit 1 Steam Generator (SG) Program. Information required by TS 5.6.6.1, Items a through h is provided below for the maintenance and refueling outage in the fall of 2007 (1R1 8). |
| : c. Nondestructive examination techniques utilized for each degradation mechanism For 1 R1 8, bobbin coil probes were used as the primary method of degradation detection except in those areas of the tube (e.g. at or below TTS, low row U-bends)where bobbin probes are not qualified. | | : a. The scope of inspections performed on each SG Primary side inspection scope involved the following: |
| In those areas of the tube, Plus Point probes were employed. | | * 100% full length bobbin coil probe examination (Row*3 through Row 47) |
| Plus Point probes were also used to re-examine any ambiguous bobbin coil signals.All examination techniques utilized during the 1 R1 8 SG tube inspections were qualified for detection of the relevant and potential degradation mechanisms listed in the 1 R1 8 degradation assessment. | | * 100% bobbin coil probe examination of the hot and cold leg straight sections in Row 1 and 2 up to and including the top support plate |
| This qualification is in accordance with Appendix H of the EPRI SG Examination Guidelines, Revision 6. Examination techniques used for detection of degradation are tabulated below.Examination Techniques Used for Detection of Degradation Degradation Mode I Tube Location Probe EPRI ETSS*Bobbin 96008.1 General Degradation Free Span Area of Tubing 3 oiuin 2109.1 3 Coil Plus Point 21409.1 PWSCC Tubesheet Region 3 Coil Plus Point 20511.1 Bobbin, 96004.1 Tube Wear AVB Locations Bbi 60.1 Coil Plus Point 21409.1 Bobbin 96001.1 Tube Wear TSP Intersections 3 Coil Plus Point 21998.1 Bobbin 96008.1 Tube Wear Periphery/Blowdown lane 3 Coiuin 998.1 3 Coil Plus Point 21998.1 Bobbin 96010.1 MBM's Free Span Area of Tubing 3 Coil Plus Point 21409.1 Bobbin 96008.1 Volumetric Full Length 3 Coil Plus Point 21998.1-The applicable EPRI Examination Technique Specification Sheets (ETSS) are listed for reference only. Site-specific examination technique sheets are developed prior to each SG inspection in accordance with Appendix H of the EPRI SG Examination Guidelines (Revision 6). | | * 100% Plus Point probe inspection of the Row 1 and 2 U-bends from the top support plate on the hot leg to the top support plate on the cold leg |
| Attachment 1 L-08-072 Page 3 of 7 d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications The following tables provide examination results for tubes with service-induced indications in each SG. All indications listed with I-codes (e.g., distorted support plate signals with indications (DSIs), non-quantifiable indications (NQIs), dents with indications (DNIs), etc,) were re-inspected with the Plus Point Probe. If the Plus Point probe indicated that no degradation/flaws were present, the tube remained in service. If degradation/flaws were reported from the Plus Point examination the tube was removed from service via plugging.SG 1RC-E-1A Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 2 61 NQI 2.11 70 6 06C 47.43 07C TEC 2 61 NDF 6 06C 47.43 06C 07C 8 30 NQI 1.56 39 6 06H 47.00 TEC TEH 8 30 NDF 6 06H 47.00 06H 07H 8 31 NQI 1.71 40 6 07H -1.66 TEC TEH 8 31 NDF 6 07H -1.66 06H 07H 8 32 NQI 1.5 58 6 07H 1.90 TEC TEH 8 32 NDF 2 07H 1.90 07H 07H 12 47 DNI 0.38 150 5 TSC 25.03 TEC TEH 12 47 NDF 6 TSC 25.03 TSC 01C 26 29 DNI 0.61 165 P1 06H -0.02 TEC TEH 26 29 NDF 6 06H -0.02 06H 06H 32 26 DNI 0.28 146 5 02H 28.80 TEC TEH 32 26 NDF 6 02H 28.80 02H 03H 33 25 FSI 0.42 48 3 06C 8.08 TEC TEH 33 25 NDF 6 06C 8.08 05C 07C 33 25 FSI 0.27 29 3 06C 14.92 TEC TEH 33 25 NDF 6 06C 14.92 05C 07C 33 25 FSI 0.11 44 3 06C 22.82 TEC TEH 33 25 NDF 6 06C 22.82 05C 07C 33 25 FSI 0.44 37 3 05C 47.14 TEC TEH 33 25 NDF 6 05C 47.14 05C 07C SG 1RC-E-1B Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 39 52 DNI 0.28 156 P1 AV3 0 TEC TEH 39 52 NDF 2 AV3 0 AV3 AV3 Attachment 1 L-08-072 Page 4 of 7 SG 1RC-E-1C Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 11* 2 DSI 0.23 114 P1 04H 0.34 TEC TEH 11* 2 PCT .0.22 114 29 P4 04H 0.39 0.21 26 04H 04H 43 44 NQI 2.11 65 6 02H 14.99 TEC TEH.43 44 NDF 6 02H 14.99 02H 03H 44 36 DNI 0.31 169 P1 TSC 18.62 TEC TEH 44 36 NDF 6 TSC 18.62 TSH FBC* Row 11, Column 2 was administratively removed from service.
| | * 100% Plus Point probe inspection of over-expansions and bulges contained within or just above the hot leg tubesheet (These were previously reported during the preservice examination and are located 10.0" or higher from the hot leg tube end) |
| Attachment 1 L-08-072 Page 5 of 7 Description of Column Headings and Data Abbreviations Used in Tables in Item d Report Header Definitions Term Definition Row/Col Row & Column of tube location Ind Type of indication (See data analysis acronyms below)Volts Voltage amplitude of an indication Phase Angular rotation of indication
| | * 100% Plus Point probe inspection of bobbin coil indications that could possibly represent degradation (I-codes) |
| %TW Percent through wall of indication Chan Frequency channel from which indication was recorded Loc Location of nearest support structure Inch Indication distance from nearest support structure Length Field measured crack length Angle Field measured crack phase angle From Starting point of examination To End point of examination Examination Extent Definitions (From -To)Term Definition FBC Flow Distribution Baffle -Cold Leg Side TEH Tube End -Hot Leg TSH Hot Leg Tubesheet TEC Tube End Cold Leg TSC Cold Leg Tubesheet 01H 1st Tube Support Plate on Hot Leg Side 01C Ist Tube Support Plate on Cold Leg Side AV1 1st Anti-Vibration Bar (Numbered hot leg to cold leg) | | * Plus Point probe inspection of the top-of-tubesheet periphery (3 tubes deep in any row or column) and the blowdown pipe region (2 tubes deep on either side of the blowdown pipe) |
| Attachment 1 L'08-072 Page 6 of 7 e. Number of tubes plugged during the inspection outage for each active degradation mechanism As stated above, no active degradation mechanisms were reported during 1 R1 8.However, one tube (R1 1 C2 in SG "C") was administratively removed from service-due to a wear indication.
| | * Plus Point probe inspection of any ambiguous bobbin coil signals P "As- left" video inspection of the hot and cold leg channel heads Secondary side inspection scope involved the following: |
| The indication was located at the upper edge of the 4 th hot leg support plate and is believed to be caused by a burr remaining from the broaching process used during fabrication of the support plate.f. Total number and percentage of tubes plugged to date The total number of tubes plugged post 1 R1 8 is 1 tube (0.01%). To date, no tubes have been repaired (sleeved). | | * Secondary side visual inspections of the top-of-tubesheet annulus, blowdown pipe region and random in-bundle locations |
| : g. The results of condition monitoring, including the results of tube pulls and in-situ testing During the 1 R18 SG tube inspections, no indications were observed that exceeded the structural integrity limits for either axial or circumferential degradation (i.e., burst integrity | | " In-bundle visual inspections of possible loose part locations identified from the analysis of the eddy current data L-08-072 Page 2 of 7 |
| > 3 times normal operating primary to secondary pressure differential). | | : b. Active degradation mechanisms found No active degradation mechanisms were reported during the 1R1 8 SG tube examinations. |
| No tubes were found to contain indications that could potentially challenge the tube integrity requirements of NEI 97-06, Revision 2, "Steam Generator Program Guidelines." Based on the 1 R18 inspection results, the SG tubing is expected to meet all operational, structural and leakage integrity requirements at the end of Cycle 19. No, indications were reported during 1 R1 8 that could possibly represent a leakage potential at postulated main steam line break conditions. | | : c. Nondestructive examination techniques utilized for each degradation mechanism For 1R1 8, bobbin coil probes were used as the primary method of degradation detection except in those areas of the tube (e.g. at or below TTS, low row U-bends) where bobbin probes are not qualified. In those areas of the tube, Plus Point probes were employed. Plus Point probes were also used to re-examine any ambiguous bobbin coil signals. |
| As such, in-situ proof or leakage testing was not required during 1 R18 and no tubes or tube sections were removed (pulled).h. The effective plugging percentage for all plugging in each SG Listed below are the plugging percentages for each SG post 1 R1 8. Since no sleeves have been installed to date, the effective plugging percentage is not applicable. | | All examination techniques utilized during the 1R1 8 SG tube inspections were qualified for detection of the relevant and potential degradation mechanisms listed in the 1R1 8 degradation assessment. This qualification is in accordance with Appendix H of the EPRI SG Examination Guidelines, Revision 6. Examination techniques used for detection of degradation are tabulated below. |
| SG Tubes Plugged % Plugged 1RC-E-1A 0 0.00 1 RC-E- B 0 0.00 1RC-E-1C 1 0.01 Attachment 1 L-08-072 Page 7 of 7 Definitions of Abbreviations, Codes and Reporting Terminology 3 Letter Code Definitions Term Bobbin Coil Definitions DNI Dent/Ding with a Possible Indication DSI Distorted Support Signal with Possible Indication FSI Free Span Indication MBM Manufacture's Burnish Mark NQI Non-Quantifiable Indication Term Plus Point Definitions NDF No Degradation Found Additional Definitions Term Definition AVB Anti-Vibration Bar DENT Indentation Observed within Tube Support Plate DING Indentation Observed in Free-Span Region EPRI Electric Power Research Institute NEI Nuclear Energy Institute PCT Percent Through Wall PWSCC. Primary Water Stress Corrosion Cracking TSP Tube Support Plate TTS Top of Tubesheet}} | | Examination Techniques Used for Detection of Degradation Degradation Mode I Tube Location Probe EPRI ETSS* |
| | Bobbin 96008.1 Free Span Area of Tubing 3 oiuin 2109.1 General Degradation 3 Coil Plus Point 21409.1 PWSCC Tubesheet Region 3 Coil Plus Point 20511.1 Tube Wear AVB Locations Bbi Bobbin, 60. |
| | 96004.1 1 Coil Plus Point 21409.1 Bobbin 96001.1 Tube Wear TSP Intersections 3 Coil Plus Point 21998.1 Bobbin 96008.1 Periphery/Blowdown lane 3 Coiuin 998.1 Tube Wear 3 Coil Plus Point 21998.1 Bobbin 96010.1 MBM's Free Span Area of Tubing 3 Coil Plus Point 21409.1 Bobbin 96008.1 Volumetric Full Length 3 Coil Plus Point 21998.1 |
| | -The applicable EPRI Examination Technique Specification Sheets (ETSS) are listed for reference only. Site-specific examination technique sheets are developed prior to each SG inspection in accordance with Appendix H of the EPRI SG Examination Guidelines (Revision 6). |
| | L-08-072 Page 3 of 7 |
| | : d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications The following tables provide examination results for tubes with service-induced indications in each SG. All indications listed with I-codes (e.g., distorted support plate signals with indications (DSIs), non-quantifiable indications (NQIs), dents with indications (DNIs), etc,) were re-inspected with the Plus Point Probe. Ifthe Plus Point probe indicated that no degradation/flaws were present, the tube remained in service. Ifdegradation/flaws were reported from the Plus Point examination the tube was removed from service via plugging. |
| | SG 1RC-E-1A Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 2 61 NQI 2.11 70 6 06C 47.43 07C TEC 2 61 NDF 6 06C 47.43 06C 07C 8 30 NQI 1.56 39 6 06H 47.00 TEC TEH 8 30 NDF 6 06H 47.00 06H 07H 8 31 NQI 1.71 40 6 07H -1.66 TEC TEH 8 31 NDF 6 07H -1.66 06H 07H 8 32 NQI 1.5 58 6 07H 1.90 TEC TEH 8 32 NDF 2 07H 1.90 07H 07H 12 47 DNI 0.38 150 5 TSC 25.03 TEC TEH 12 47 NDF 6 TSC 25.03 TSC 01C 26 29 DNI 0.61 165 P1 06H -0.02 TEC TEH 26 29 NDF 6 06H -0.02 06H 06H 32 26 DNI 0.28 146 5 02H 28.80 TEC TEH 32 26 NDF 6 02H 28.80 02H 03H 33 25 FSI 0.42 48 3 06C 8.08 TEC TEH 33 25 NDF 6 06C 8.08 05C 07C 33 25 FSI 0.27 29 3 06C 14.92 TEC TEH 33 25 NDF 6 06C 14.92 05C 07C 33 25 FSI 0.11 44 3 06C 22.82 TEC TEH 33 25 NDF 6 06C 22.82 05C 07C 33 25 FSI 0.44 37 3 05C 47.14 TEC TEH 33 25 NDF 6 05C 47.14 05C 07C SG 1RC-E-1B Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 39 52 DNI 0.28 156 P1 AV3 0 TEC TEH 39 52 NDF 2 AV3 0 AV3 AV3 L-08-072 Page 4 of 7 SG 1RC-E-1C Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 11* 2 DSI 0.23 114 P1 04H 0.34 TEC TEH 11* 2 PCT .0.22 114 29 P4 04H 0.39 0.21 26 04H 04H 43 44 NQI 2.11 65 6 02H 14.99 TEC TEH |
| | .43 44 NDF 6 02H 14.99 02H 03H 44 36 DNI 0.31 169 P1 TSC 18.62 TEC TEH 44 36 NDF 6 TSC 18.62 TSH FBC |
| | * Row 11, Column 2 was administratively removed from service. |
| | L-08-072 Page 5 of 7 Description of Column Headings and Data Abbreviations Used in Tables in Item d Report Header Definitions Term Definition Row/Col Row & Column of tube location Ind Type of indication (See data analysis acronyms below) |
| | Volts Voltage amplitude of an indication Phase Angular rotation of indication |
| | %TW Percent through wall of indication Chan Frequency channel from which indication was recorded Loc Location of nearest support structure Inch Indication distance from nearest support structure Length Field measured crack length Angle Field measured crack phase angle From Starting point of examination To End point of examination Examination Extent Definitions (From - To) |
| | Term Definition FBC Flow Distribution Baffle - Cold Leg Side TEH Tube End - Hot Leg TSH Hot Leg Tubesheet TEC Tube End Cold Leg TSC Cold Leg Tubesheet 01H 1st Tube Support Plate on Hot Leg Side 01C Ist Tube Support Plate on Cold Leg Side AV1 1st Anti-Vibration Bar (Numbered hot leg to cold leg) |
| | L'08-072 Page 6 of 7 |
| | : e. Number of tubes plugged during the inspection outage for each active degradation mechanism As stated above, no active degradation mechanisms were reported during 1R1 8. |
| | However, one tube (R1 1 C2 in SG "C") was administratively removed from service-due to a wear indication. The indication was located at the upper edge of the 4 th hot leg support plate and is believed to be caused by a burr remaining from the broaching process used during fabrication of the support plate. |
| | : f. Total number and percentage of tubes plugged to date The total number of tubes plugged post 1 R1 8 is 1 tube (0.01%). To date, no tubes have been repaired (sleeved). |
| | : g. The results of condition monitoring, including the results of tube pulls and in-situ testing During the 1R18 SG tube inspections, no indications were observed that exceeded the structural integrity limits for either axial or circumferential degradation (i.e., burst integrity > 3 times normal operating primary to secondary pressure differential). No tubes were found to contain indications that could potentially challenge the tube integrity requirements of NEI 97-06, Revision 2, "Steam Generator Program Guidelines." |
| | Based on the 1 R18 inspection results, the SG tubing is expected to meet all operational, structural and leakage integrity requirements at the end of Cycle 19. No, indications were reported during 1R1 8 that could possibly represent a leakage potential at postulated main steam line break conditions. As such, in-situ proof or leakage testing was not required during 1R18 and no tubes or tube sections were removed (pulled). |
| | : h. The effective plugging percentage for all plugging in each SG Listed below are the plugging percentages for each SG post 1R1 8. Since no sleeves have been installed to date, the effective plugging percentage is not applicable. |
| | SG Tubes Plugged % Plugged 1RC-E-1A 0 0.00 1RC-E- B 0 0.00 1RC-E-1C 1 0.01 L-08-072 Page 7 of 7 Definitions of Abbreviations, Codes and Reporting Terminology 3 Letter Code Definitions Term Bobbin Coil Definitions DNI Dent/Ding with a Possible Indication DSI Distorted Support Signal with Possible Indication FSI Free Span Indication MBM Manufacture's Burnish Mark NQI Non-Quantifiable Indication Term Plus Point Definitions NDF No Degradation Found Additional Definitions Term Definition AVB Anti-Vibration Bar DENT Indentation Observed within Tube Support Plate DING Indentation Observed in Free-Span Region EPRI Electric Power Research Institute NEI Nuclear Energy Institute PCT Percent Through Wall PWSCC. Primary Water Stress Corrosion Cracking TSP Tube Support Plate TTS Top of Tubesheet}} |
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Category:Letter type:L
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Discharge Monitoring Report (NPDES) Permit No. PA0025615 for Second Half of 2022
L-22-281, Discharge Monitoring Report (NPDES) Permit No. PA00256152022-12-16016 December 2022
Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-22-246, Response to Request for Additional Information Regarding a Request to Consolidate Fuel Decay Time Technical Specifications to New LCO2022-12-0707 December 2022
Response to Request for Additional Information Regarding a Request to Consolidate Fuel Decay Time Technical Specifications to New LCO
L-22-217, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152022-11-21021 November 2022
Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-22-226, Emergency Preparedness Plan2022-11-0404 November 2022
Emergency Preparedness Plan
L-22-222, Cycle 29-1 Core Operating Limits Report2022-10-31031 October 2022
Cycle 29-1 Core Operating Limits Report
L-22-228, Independent Spent Fuel Storage Installation Changes, Tests, and Experiments2022-10-26026 October 2022
Independent Spent Fuel Storage Installation Changes, Tests, and Experiments
L-22-200, Response to Request for Additional Information Regarding a 180-Day Steam Generator Tube Inspection Report Fall 2021 Refueling Outage2022-10-21021 October 2022
Response to Request for Additional Information Regarding a 180-Day Steam Generator Tube Inspection Report Fall 2021 Refueling Outage
L-22-232, Withdrawal of a License Amendment Request Associated with Fire Protection Program Changes2022-10-21021 October 2022
Withdrawal of a License Amendment Request Associated with Fire Protection Program Changes
L-22-238, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152022-10-20020 October 2022
Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-22-227, Revised August 2022 Outfall 003 & 004 NPDES Discharge Monitoring Report (Dmr), Revised Daily Effluent Monitoring Report Forms for Outfalls 003 & 004 and Corrective Action Letter from Eurofins2022-10-0303 October 2022
Revised August 2022 Outfall 003 & 004 NPDES Discharge Monitoring Report (Dmr), Revised Daily Effluent Monitoring Report Forms for Outfalls 003 & 004 and Corrective Action Letter from Eurofins
L-22-219, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152022-09-26026 September 2022
Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-22-204, Submittal of Evacuation Time Estimates2022-09-0707 September 2022
Submittal of Evacuation Time Estimates
L-22-137, Request for Fire Protection Program Changes2022-09-0606 September 2022
Request for Fire Protection Program Changes
L-21-238, License Amendment Request for Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident2022-08-31031 August 2022
License Amendment Request for Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident
L-22-188, Response to Request for Additional Information Regarding Steam Generator Inspection Report- Fall 2021 Refueling Outage2022-08-22022 August 2022
Response to Request for Additional Information Regarding Steam Generator Inspection Report- Fall 2021 Refueling Outage
L-22-191, Spent Fuel Storage Cask Registration2022-08-17017 August 2022
Spent Fuel Storage Cask Registration
2023-09-14
[Table view]
Category:Report
MONTHYEARL-23-188, Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-08-0707 August 2023
Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments
L-22-154, Pressure and Temperature Limits Report Revision2022-07-0606 July 2022
Pressure and Temperature Limits Report Revision
L-22-054, Steam Generator Inspection Report- Fall 2021 Refueling Outage2022-02-10010 February 2022
Steam Generator Inspection Report- Fall 2021 Refueling Outage
IR 05000334/20210052021-09-0101 September 2021
Updated Inspection Plan for the Beaver Valley Power Station, Units 1 and 2 (Report 05000334/2021005 and 05000412/2021005)
ML21131A0242021-05-17017 May 2021
Review of Reactor Pressure Vessel Capsule Y Analysis Report
ML21055A0582021-03-10010 March 2021
Review of the Refueling Outage 21 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and F* Reports
L-21-063, Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks2021-02-0101 February 2021
Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks
ML21026A3372021-01-26026 January 2021
Steam Generator Inspection Report Revision - Spring 2020 Refueling Outage
L-20-257, Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic2020-09-22022 September 2020
Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic
ML21127A1652020-07-17017 July 2020
ODCM: Index, Matrix and History of ODCM Changes
L-21-135, ODCM: Liquid Effluents2020-07-17017 July 2020
ODCM: Liquid Effluents
ML21127A1482020-07-17017 July 2020
ODCM: Controls for RETS and REMP Programs
L-20-116, CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency2020-04-0101 April 2020
CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency
ML19280A0382019-10-22022 October 2019
Ti 2515-191 Summary of Findings
ML21127A1562019-07-17017 July 2019
ODCM: Gaseous Effluents
ML19051A1092019-02-20020 February 2019
LTR-SDA-18-054-NP, Revision 0 - Technical Justification to Support the Extended Volumetric Examination Interval for Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds
L-19-053, Emergency Response Data System Data Point Library Update2019-02-20020 February 2019
Emergency Response Data System Data Point Library Update
L-18-037, Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2018-06-18018 June 2018
Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models
L-18-017, Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado.2018-03-15015 March 2018
Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado.
ML21127A1592018-01-22022 January 2018
ODCM: Radiological Environmental Monitoring Program
L-17-277, Pressure and Temperature Limits Reports, Revisions 8 and 92017-10-0404 October 2017
Pressure and Temperature Limits Reports, Revisions 8 and 9
L-17-259, Mitigating Strategies Assessment (MSA) for Flooding2017-09-20020 September 2017
Mitigating Strategies Assessment (MSA) for Flooding
ML17213A0162017-05-11011 May 2017
Enclosure a: Beaver Valley, Units 1, Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1
ML17213A0172017-05-11011 May 2017
Enclosure B: Beaver Valley, Unit 2, Seismic Probabilistic Risk Assessment in Response to 50.54(0 Letter with Regard to NTTF 2.1
ML17024A0362016-12-31031 December 2016
Operating Data Report for 2016
ML16312A3112016-11-0707 November 2016
Spent Fuel Pool Evaluation Supplemental Report. Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Ntif) Review...
L-16-327, Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair2016-10-31031 October 2016
Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair
L-16-093, Report of Facility Changes, Tests and Experiments2016-06-0303 June 2016
Report of Facility Changes, Tests and Experiments
L-16-065, Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22016-04-0606 April 2016
Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2
L-16-006, Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information2016-01-20020 January 2016
Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information
L-15-386, Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time2016-01-0505 January 2016
Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time
L-15-337, Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool..2015-12-21021 December 2015
Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool..
ML16277A1142015-10-28028 October 2015
ERS-MPD-93-007, Rev. 11, BVPS-U1 Gaseous Radioactivity Monitor Emergency Action Levels.
ML15274A3072015-10-0505 October 2015
Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54 Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review
L-15-187, Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2015-09-16016 September 2015
Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models
ML15132A4962015-05-22022 May 2015
Review of the 2014 Steam Generator Tube Inspection Report (TAC No. M4620)
ML16277A1222015-04-20020 April 2015
ERS-LMR-14-001, Rev. 1, Liquid Monitor Alert Emergency Action Level (EAL) Set Points.
ML15119A1042015-03-31031 March 2015
SG-CCOE-14-4-NP, Revision 1, Examination of Steam Generator Tubes Removed, Part 1 of 4
L-15-091, SG-CCOE-14-4-NP, Revision 1, Examination of Steam Generator Tubes Removed, Part 2 of 42015-03-31031 March 2015
SG-CCOE-14-4-NP, Revision 1, Examination of Steam Generator Tubes Removed, Part 2 of 4
ML21127A1582015-02-11011 February 2015
ODCM: Bases for ODCM Controls
ML16277A1152015-01-30030 January 2015
ERS-HHM-87-014, Rev. 8, Unit 1 / Unit 2 ODCM Gaseous Effluent Monitor Setpoints.
ML15027A2352015-01-28028 January 2015
Regulatory Audit in Support of the License Amendment Request to Implement a Risk-Informed, Performance-Based, Fire Protection Program (Tac Nos. MF3301 & MF3302)
L-14-401, First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In2014-12-19019 December 2014
First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In
L-14-352, Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22014-11-0505 November 2014
Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2
ML21127A1442014-10-0707 October 2014
ODCM: Information Related to 40 CFR 190
L-14-259, Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)2014-08-28028 August 2014
Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)
ML14224A5772014-07-31031 July 2014
SG-SGMP-14-17, Revision 1, Beaver Valley, Unit 2, End-of-Cycle 17 Analysis and Prediction for End-of-Cycle 18 Voltage-Based Repair Criteria 90-Day Report.
ML14224A5762014-07-31031 July 2014
SG-CCOE-14-1, Revision 0, Generic Letter 95-05 Tube Intersection Burst Test, Leakage Test, and Morphology Conclusions.
ML14224A5752014-06-25025 June 2014
R17 Steam Generator F* (F Star) Report
ML14009A0892014-02-10010 February 2014
Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Daiichi Nuclear Power Plant Accident
2023-08-07
[Table view]
Category:Miscellaneous
MONTHYEARL-23-188, Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-08-0707 August 2023
Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments
IR 05000334/20210052021-09-0101 September 2021
Updated Inspection Plan for the Beaver Valley Power Station, Units 1 and 2 (Report 05000334/2021005 and 05000412/2021005)
ML21131A0242021-05-17017 May 2021
Review of Reactor Pressure Vessel Capsule Y Analysis Report
ML21055A0582021-03-10010 March 2021
Review of the Refueling Outage 21 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and F* Reports
L-21-063, Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks2021-02-0101 February 2021
Facility Id 04-13361, Unit 2 Emergency Diesel Generator Day Tanks
L-20-257, Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic2020-09-22022 September 2020
Request for One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Pandemic
L-19-053, Emergency Response Data System Data Point Library Update2019-02-20020 February 2019
Emergency Response Data System Data Point Library Update
L-18-037, Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2018-06-18018 June 2018
Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models
L-18-017, Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado.2018-03-15015 March 2018
Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformances Identified in Response to Regulatory Issue Summary 2015-06, Tornado.
ML17024A0362016-12-31031 December 2016
Operating Data Report for 2016
ML16312A3112016-11-0707 November 2016
Spent Fuel Pool Evaluation Supplemental Report. Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (Ntif) Review...
L-16-093, Report of Facility Changes, Tests and Experiments2016-06-0303 June 2016
Report of Facility Changes, Tests and Experiments
L-16-065, Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22016-04-0606 April 2016
Transmittal of 180-Day Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2
L-16-006, Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information2016-01-20020 January 2016
Submittal of Annual Report Pursuant to 10 CFR 75.11, Location Information
L-15-386, Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time2016-01-0505 January 2016
Post Accident Monitoring Report Regarding Inoperability of One of the Source Range Nuclear Excore Instrumentation Channel for a Period in Excess of the 30-Day Restoration Time
L-15-337, Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool..2015-12-21021 December 2015
Completion of Required Action by NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events and NRC Order EA-12-051, Reliable Spent Fuel Pool..
ML15274A3072015-10-0505 October 2015
Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54 Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review
L-15-187, Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2015-09-16016 September 2015
Submittal of 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models
ML15132A4962015-05-22022 May 2015
Review of the 2014 Steam Generator Tube Inspection Report (TAC No. M4620)
ML15027A2352015-01-28028 January 2015
Regulatory Audit in Support of the License Amendment Request to Implement a Risk-Informed, Performance-Based, Fire Protection Program (Tac Nos. MF3301 & MF3302)
L-14-401, First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In2014-12-19019 December 2014
First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In
L-14-352, Steam Generator Tube Inspection Report - Technical Specification 5.6.6.22014-11-0505 November 2014
Steam Generator Tube Inspection Report - Technical Specification 5.6.6.2
L-14-259, Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)2014-08-28028 August 2014
Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)
ML14009A0892014-02-10010 February 2014
Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Daiichi Nuclear Power Plant Accident
ML13284A0242013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Appendix B, Seismic Walkdown Checklists (Swcs), Part 3 of 5
ML13284A0272013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Appendix B, Seismic Walkdown Checklists (Swcs), Continued, Part 4 of 5
ML13284A0262013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Appendix a, Resumes and Qualifications, Part 2 of 5
ML13284A0252013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3, Seismic Walkdown Report Revision 1, Part 1 of 5
ML13284A0232013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix a, Resumes and Qualifications, Part 2 of 5
ML13284A0192013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix B Seismic Walkdown Checklists (Swcs), Part 3 of 5
ML13284A0222013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Part 1 of 5
ML13284A0202013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix B Seismic Walkdown Checklists (Swcs) Continued. Part 4 of 5
ML13284A0212013-09-0404 September 2013
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Revision 1, Appendix C Area Walk-by Checklists (Awcs), Part 5 of 5
L-13-036, Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedules2013-05-28028 May 2013
Proposed Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedules
L-13-095, Discharge Monitoring Report (NPDES) Permit No. PA00256152013-02-22022 February 2013
Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-12-437, Containment Liner Random and Non-Random Examinations Report2012-12-18018 December 2012
Containment Liner Random and Non-Random Examinations Report
L-12-347, FENOC Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2012-11-27027 November 2012
FENOC Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident
ML13008A0422012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklits (Swcs), Battery 2N. Pages 1 - 82
ML13008A0412012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report. Cover Through A-40
ML13135A2382012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B - Seismic Walkdown Checklists (Swcs), Sheet 1 of 513 Through Sheet 301 of 513
ML13008A0432012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), Purge Valve Isolation. Pages 83 - 161
ML13008A0442012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), Bus F1, Low Voltage Switchgear. Pages 162 - 240
ML13008A0452012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), SGI Level Switch SP9B6. Pages 241 - 318
ML13008A0462012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B Seismic Walkdown Checklists (Swcs), Relief Valves SW 3963. Pages 319 - End
ML13135A2412012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C - Area Walk-by Checklists (Awcs) and Appendix D - Component List for Anchorage Configuration Check
ML13135A2402012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B - Seismic Walkdown Checklists (Swcs), Sheet 302 of 513 Through Sheet 513 of 513
ML13135A2362012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report
ML13135A2372012-10-31031 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix a - Resumes and Qualifications
ML13008A0402012-10-23023 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C Area Walk-by Checklists (Awcs), Prim. Aux. Bldg 735'6. Pages 135 - 191
ML13008A0392012-10-23023 October 2012
Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C Area Walk-by Checklists (Awcs), Emer. Swgr. Room Train A. Pages 66 - 134
2023-08-07
[Table view] |
Text
FENOC FirstEnergyNuclear OperatingCompany Peter P. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 March 18, 2008 L-08-072 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Technical Specification 5.6.6.1 - Steam Generator Inspection Report In accordance with Beaver Valley Power Station Unit No. 1 Technical Specification (TS) 5.6.6.1, information related to steam generator inspections is provided in Attachment 1.
There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager -
FENOC Fleet Licensing, at 330-761-6071.
Sincerely, Peter P. Sena III Attachments:
- 1. Technical Specification 5.6.6.1 - Unit 1 SG Tube Inspection Report cc: Mr. S. J. Collins, NRC Region I Administrator Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager Mr. D. J. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)
ATTACHMENT 1 L-08-072 Beaver Valley Power Station Unit No. 1 Technical Specification 5.6.6.1 - Unit 1 SG Tube Inspection Report Page 1 of 7 The following information satisfies the reporting requirement of Beaver Valley Power Station Technical Specification (TS) 5.6.6.1. Refer to the tables on Page 7 of this attachment which define abbreviations, codes and other terms used throughout this report.
Information required by Technical Specification 5.6.6.1 Technical Specification 5.6.6.1 requires reporting within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.5.1, Unit 1 Steam Generator (SG) Program. Information required by TS 5.6.6.1, Items a through h is provided below for the maintenance and refueling outage in the fall of 2007 (1R1 8).
- a. The scope of inspections performed on each SG Primary side inspection scope involved the following:
- 100% full length bobbin coil probe examination (Row*3 through Row 47)
- 100% bobbin coil probe examination of the hot and cold leg straight sections in Row 1 and 2 up to and including the top support plate
- 100% Plus Point probe inspection of the Row 1 and 2 U-bends from the top support plate on the hot leg to the top support plate on the cold leg
- 100% Plus Point probe inspection of over-expansions and bulges contained within or just above the hot leg tubesheet (These were previously reported during the preservice examination and are located 10.0" or higher from the hot leg tube end)
- 100% Plus Point probe inspection of bobbin coil indications that could possibly represent degradation (I-codes)
- Plus Point probe inspection of the top-of-tubesheet periphery (3 tubes deep in any row or column) and the blowdown pipe region (2 tubes deep on either side of the blowdown pipe)
- Plus Point probe inspection of any ambiguous bobbin coil signals P "As- left" video inspection of the hot and cold leg channel heads Secondary side inspection scope involved the following:
- Secondary side visual inspections of the top-of-tubesheet annulus, blowdown pipe region and random in-bundle locations
" In-bundle visual inspections of possible loose part locations identified from the analysis of the eddy current data L-08-072 Page 2 of 7
- b. Active degradation mechanisms found No active degradation mechanisms were reported during the 1R1 8 SG tube examinations.
- c. Nondestructive examination techniques utilized for each degradation mechanism For 1R1 8, bobbin coil probes were used as the primary method of degradation detection except in those areas of the tube (e.g. at or below TTS, low row U-bends) where bobbin probes are not qualified. In those areas of the tube, Plus Point probes were employed. Plus Point probes were also used to re-examine any ambiguous bobbin coil signals.
All examination techniques utilized during the 1R1 8 SG tube inspections were qualified for detection of the relevant and potential degradation mechanisms listed in the 1R1 8 degradation assessment. This qualification is in accordance with Appendix H of the EPRI SG Examination Guidelines, Revision 6. Examination techniques used for detection of degradation are tabulated below.
Examination Techniques Used for Detection of Degradation Degradation Mode I Tube Location Probe EPRI ETSS*
Bobbin 96008.1 Free Span Area of Tubing 3 oiuin 2109.1 General Degradation 3 Coil Plus Point 21409.1 PWSCC Tubesheet Region 3 Coil Plus Point 20511.1 Tube Wear AVB Locations Bbi Bobbin, 60.
96004.1 1 Coil Plus Point 21409.1 Bobbin 96001.1 Tube Wear TSP Intersections 3 Coil Plus Point 21998.1 Bobbin 96008.1 Periphery/Blowdown lane 3 Coiuin 998.1 Tube Wear 3 Coil Plus Point 21998.1 Bobbin 96010.1 MBM's Free Span Area of Tubing 3 Coil Plus Point 21409.1 Bobbin 96008.1 Volumetric Full Length 3 Coil Plus Point 21998.1
-The applicable EPRI Examination Technique Specification Sheets (ETSS) are listed for reference only. Site-specific examination technique sheets are developed prior to each SG inspection in accordance with Appendix H of the EPRI SG Examination Guidelines (Revision 6).
L-08-072 Page 3 of 7
- d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications The following tables provide examination results for tubes with service-induced indications in each SG. All indications listed with I-codes (e.g., distorted support plate signals with indications (DSIs), non-quantifiable indications (NQIs), dents with indications (DNIs), etc,) were re-inspected with the Plus Point Probe. Ifthe Plus Point probe indicated that no degradation/flaws were present, the tube remained in service. Ifdegradation/flaws were reported from the Plus Point examination the tube was removed from service via plugging.
SG 1RC-E-1A Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 2 61 NQI 2.11 70 6 06C 47.43 07C TEC 2 61 NDF 6 06C 47.43 06C 07C 8 30 NQI 1.56 39 6 06H 47.00 TEC TEH 8 30 NDF 6 06H 47.00 06H 07H 8 31 NQI 1.71 40 6 07H -1.66 TEC TEH 8 31 NDF 6 07H -1.66 06H 07H 8 32 NQI 1.5 58 6 07H 1.90 TEC TEH 8 32 NDF 2 07H 1.90 07H 07H 12 47 DNI 0.38 150 5 TSC 25.03 TEC TEH 12 47 NDF 6 TSC 25.03 TSC 01C 26 29 DNI 0.61 165 P1 06H -0.02 TEC TEH 26 29 NDF 6 06H -0.02 06H 06H 32 26 DNI 0.28 146 5 02H 28.80 TEC TEH 32 26 NDF 6 02H 28.80 02H 03H 33 25 FSI 0.42 48 3 06C 8.08 TEC TEH 33 25 NDF 6 06C 8.08 05C 07C 33 25 FSI 0.27 29 3 06C 14.92 TEC TEH 33 25 NDF 6 06C 14.92 05C 07C 33 25 FSI 0.11 44 3 06C 22.82 TEC TEH 33 25 NDF 6 06C 22.82 05C 07C 33 25 FSI 0.44 37 3 05C 47.14 TEC TEH 33 25 NDF 6 05C 47.14 05C 07C SG 1RC-E-1B Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 39 52 DNI 0.28 156 P1 AV3 0 TEC TEH 39 52 NDF 2 AV3 0 AV3 AV3 L-08-072 Page 4 of 7 SG 1RC-E-1C Row Col Ind Volts Phase %TW Chan Loc Inch Length Angle From To 11* 2 DSI 0.23 114 P1 04H 0.34 TEC TEH 11* 2 PCT .0.22 114 29 P4 04H 0.39 0.21 26 04H 04H 43 44 NQI 2.11 65 6 02H 14.99 TEC TEH
.43 44 NDF 6 02H 14.99 02H 03H 44 36 DNI 0.31 169 P1 TSC 18.62 TEC TEH 44 36 NDF 6 TSC 18.62 TSH FBC
- Row 11, Column 2 was administratively removed from service.
L-08-072 Page 5 of 7 Description of Column Headings and Data Abbreviations Used in Tables in Item d Report Header Definitions Term Definition Row/Col Row & Column of tube location Ind Type of indication (See data analysis acronyms below)
Volts Voltage amplitude of an indication Phase Angular rotation of indication
%TW Percent through wall of indication Chan Frequency channel from which indication was recorded Loc Location of nearest support structure Inch Indication distance from nearest support structure Length Field measured crack length Angle Field measured crack phase angle From Starting point of examination To End point of examination Examination Extent Definitions (From - To)
Term Definition FBC Flow Distribution Baffle - Cold Leg Side TEH Tube End - Hot Leg TSH Hot Leg Tubesheet TEC Tube End Cold Leg TSC Cold Leg Tubesheet 01H 1st Tube Support Plate on Hot Leg Side 01C Ist Tube Support Plate on Cold Leg Side AV1 1st Anti-Vibration Bar (Numbered hot leg to cold leg)
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- e. Number of tubes plugged during the inspection outage for each active degradation mechanism As stated above, no active degradation mechanisms were reported during 1R1 8.
However, one tube (R1 1 C2 in SG "C") was administratively removed from service-due to a wear indication. The indication was located at the upper edge of the 4 th hot leg support plate and is believed to be caused by a burr remaining from the broaching process used during fabrication of the support plate.
- f. Total number and percentage of tubes plugged to date The total number of tubes plugged post 1 R1 8 is 1 tube (0.01%). To date, no tubes have been repaired (sleeved).
- g. The results of condition monitoring, including the results of tube pulls and in-situ testing During the 1R18 SG tube inspections, no indications were observed that exceeded the structural integrity limits for either axial or circumferential degradation (i.e., burst integrity > 3 times normal operating primary to secondary pressure differential). No tubes were found to contain indications that could potentially challenge the tube integrity requirements of NEI 97-06, Revision 2, "Steam Generator Program Guidelines."
Based on the 1 R18 inspection results, the SG tubing is expected to meet all operational, structural and leakage integrity requirements at the end of Cycle 19. No, indications were reported during 1R1 8 that could possibly represent a leakage potential at postulated main steam line break conditions. As such, in-situ proof or leakage testing was not required during 1R18 and no tubes or tube sections were removed (pulled).
- h. The effective plugging percentage for all plugging in each SG Listed below are the plugging percentages for each SG post 1R1 8. Since no sleeves have been installed to date, the effective plugging percentage is not applicable.
SG Tubes Plugged % Plugged 1RC-E-1A 0 0.00 1RC-E- B 0 0.00 1RC-E-1C 1 0.01 L-08-072 Page 7 of 7 Definitions of Abbreviations, Codes and Reporting Terminology 3 Letter Code Definitions Term Bobbin Coil Definitions DNI Dent/Ding with a Possible Indication DSI Distorted Support Signal with Possible Indication FSI Free Span Indication MBM Manufacture's Burnish Mark NQI Non-Quantifiable Indication Term Plus Point Definitions NDF No Degradation Found Additional Definitions Term Definition AVB Anti-Vibration Bar DENT Indentation Observed within Tube Support Plate DING Indentation Observed in Free-Span Region EPRI Electric Power Research Institute NEI Nuclear Energy Institute PCT Percent Through Wall PWSCC. Primary Water Stress Corrosion Cracking TSP Tube Support Plate TTS Top of Tubesheet