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Category:Report
MONTHYEARRA-24-0173, Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld2024-06-28028 June 2024 Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023 Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023 Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022 Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary) RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-21-0312, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2021-11-22022 November 2021 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) IR 05000261/20210052021-08-25025 August 2021 Updated Inspection Plan for H. B. Robinson Steam Electric Plant, Unit 2 (Report 05000261/2021005) RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 RNP-RA/18-0024, Report of Changes Pursuant to 10 CFR 50.59(d)(2)2018-04-0202 April 2018 Report of Changes Pursuant to 10 CFR 50.59(d)(2) RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017 Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions) ML16281A5102016-12-15015 December 2016 Staff Assessment of the Reactor Vessel Internals Aging Management Program Plans RNP-RA/16-0087, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-31031 October 2016 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals RNP-RA/16-0078, Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors2016-10-0505 October 2016 Technical Specifications Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable High Range Containment Area Radiation Monitors RA-16-0024, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P2016-10-0303 October 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RNP-RA/16-0038, Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-1742016-05-25025 May 2016 Transition Report, Revision 1, Transition to National Fire Protection Association Standard 805, with Attachments a, I, J, L, M, and V. Pages 22-174 RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P ML15253A4102016-03-11011 March 2016 Enclosure 2 Screening Analysis Report RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. RA-15-0047, Annual Report of Changes Pursuant to 10 CFR 50.462015-11-17017 November 2015 Annual Report of Changes Pursuant to 10 CFR 50.46 ML15280A1992015-10-19019 October 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review RNP-RA/15-0053, Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04)2015-08-19019 August 2015 Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12-04) RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RNP-RA/15-0018, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel2015-02-26026 February 2015 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Power Range Neutron Flux Channel RNP-RA/14-0037, Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System2014-04-0808 April 2014 Response to NRC Request for Additional Information Regarding License Amendment Request to Modify Technical Specification 3.4.12, Low Temperature Overpressure Protection System RNP-RA/14-0012, Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3....2014-03-12012 March 2014 Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3.... RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014 Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident ML13365A2912014-02-19019 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14027A0632014-01-24024 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for H. B. Robinson Steam Electric Plant, Unit 2, TAC No.: MF0720 ML13267A2122013-09-30030 September 2013 Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition ML13270A1762013-09-24024 September 2013 Redacted - Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch (Srxb) Support of Region II Inspection of H. B. Robinson Treatment of Voids in Systems That Are Important to Safety RNP-RA/13-0079, Technical Specifications (TS) Section 5.8.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Wide Range Containment Pressure Transmitter2013-08-21021 August 2013 Technical Specifications (TS) Section 5.8.6, Post Accident Monitoring (PAM) Instrumentation Report, 14-Day Report for the Inoperability of the Wide Range Containment Pressure Transmitter RNP-RA/13-0066, Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for the Pressurizer Safety Valve Indication2013-06-24024 June 2013 Technical Specifications Section 5.6.6 Post Accident Monitoring Instrumentation Report for the Pressurizer Safety Valve Indication RNP-RA/13-0037, Independent Spent Fuel Storage Installation, Annual Individual Exposure Monitoring Report for 20122013-04-25025 April 2013 Independent Spent Fuel Storage Installation, Annual Individual Exposure Monitoring Report for 2012 ML12331A1752012-11-26026 November 2012 Draft Bypass Fiber Quantity Test Plan ML12278A3992012-08-31031 August 2012 WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant ML1210707162012-05-17017 May 2012 Letter Report on the Evaluation of Cables from the HEAF Fire Event at the H.B. Robinson Steam Electric Plant ML12068A1332012-02-23023 February 2012 Calculation RNP-M/MECH-1815, Revision 1, Evaluation of Emergency Diesel Generator Starting Capability at 150 PSIG RNP-RA/11-0100, Annual Report of Changes to or Errors Discovered in an Acceptable Loss-of-Coolant Accident Evaluation Model Application for the Emergency Core Cooling System2011-11-23023 November 2011 Annual Report of Changes to or Errors Discovered in an Acceptable Loss-of-Coolant Accident Evaluation Model Application for the Emergency Core Cooling System ML1124113592011-09-23023 September 2011 Final Precursor Analysis: Electrical Fault Causes Fire and Subsequent Reactor Trip with a Loss of Reactor Coolant Pump Seal Injection and Cooling ML1124113582011-09-23023 September 2011 Final Precursor Analysis: Concurrent Unavailabilities - EDG B Inoperable Due to Failed Output Breaker and EDG a Unavailable Due to Testing and Maintenance ML1128005282010-12-29029 December 2010 NRC 2011 Hb Robinson ML1019304172010-05-0606 May 2010 Tritium Database Report RNP-RA/09-0081, WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.2009-07-31031 July 2009 WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant. RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009 Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 2024-06-28
[Table view] Category:Technical
MONTHYEARRA-24-0173, Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld2024-06-28028 June 2024 Renewed License Number DPR-23 Request Review of White Papers to Determine Fracture Toughness-based Reference Temperature of Heat Number W5214 Weld RA-23-0313, Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation2023-12-14014 December 2023 Baffle-Former Bolt (Bfb) Subsequent Inspection Interval Evaluation RA-23-0141, Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule2023-07-12012 July 2023 Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule RA-23-0120, Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.22023-05-31031 May 2023 Supplemental Information Regarding Addition of Feedwater Isolation on Steam Generator Level High-High to Technical Specification 3.3.2 RA-23-0080, Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube2023-04-0505 April 2023 Refueling Outage 32 Steam Generator Tube Inspection Report Supplement Pursuant to Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube RA-22-0302, Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range)2022-11-0101 November 2022 Technical Specifications (TS) Section 5.6.6, Post Accident Monitoring Instrumentation Report for Inoperable Containment Sump Water Level (Wide Range) RA-22-0239, Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary)2022-08-0909 August 2022 Technical Specifications Section 5.6.6. Post Accident Monitoring Instrumentation Report for Inoperable Safety Valve Position (Primary) RA-22-0017, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-01-0606 January 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections RA-17-0040, Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions)2017-08-15015 August 2017 Providing Methodology Reports DPC-NE-1008-P, Revision 0, DPC-NF-2010, Revision 3, and DPC-NE-2011-P, Revision 2. (Non-Proprietary Versions) ML16280A2002016-10-0505 October 2016 Response to Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals RNP-RA/16-0073, Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits2016-09-14014 September 2016 Response to Request for Additional Information Regarding License Amendment Request to Revise Reactor Coolant System Pressure and Temperature Limits RA-16-0023, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P2016-05-0404 May 2016 Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-1008-P RA-15-0042, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis.2015-11-19019 November 2015 Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis. RA-15-0031, Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors.2015-08-19019 August 2015 Application to Revise Technical Specifications for Methodology Report DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/Simulate-3 for Westinghouse Reactors. RNP-RA/14-0011, Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2014-02-27027 February 2014 Revision to Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident ML13365A2912014-02-19019 February 2014 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14027A0632014-01-24024 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for H. B. Robinson Steam Electric Plant, Unit 2, TAC No.: MF0720 ML13267A2122013-09-30030 September 2013 Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition ML13270A1762013-09-24024 September 2013 Redacted - Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch (Srxb) Support of Region II Inspection of H. B. Robinson Treatment of Voids in Systems That Are Important to Safety ML12331A1752012-11-26026 November 2012 Draft Bypass Fiber Quantity Test Plan ML12278A3992012-08-31031 August 2012 WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant ML1210707162012-05-17017 May 2012 Letter Report on the Evaluation of Cables from the HEAF Fire Event at the H.B. Robinson Steam Electric Plant ML12068A1332012-02-23023 February 2012 Calculation RNP-M/MECH-1815, Revision 1, Evaluation of Emergency Diesel Generator Starting Capability at 150 PSIG RNP-RA/09-0081, WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant.2009-07-31031 July 2009 WCAP-17077-NP, Revision 0, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant. RNP-RA/09-0054, Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation2009-06-19019 June 2009 Attachments VI & VII to Serial: RNP-RA/09-0054 - Request for Technical Specifications Change to Section 3.3.1 Reactor Protection System Instrumentation ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 RNP-RA/06-0081, Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP2006-08-31031 August 2006 Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet, WCAP-16627-NP RNP-RA/06-0027, ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis.2006-03-31031 March 2006 ANP-2512(NP), Loss of Forced Reactor Coolant Flow Analysis. ML0507004082004-02-20020 February 2004 EMF-3030(NP), Revision 0, Robinson Nuclear Plant, Realistic Large Break LOCA Analysis, February 2004, Non-Proprietary Version RNP-RA/03-0075, Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-62003-07-31031 July 2003 Technical Basis for RPV Head CRDM Nozzle Inspection Interval H.B. Robinson Steam Electric Plant, Unit No. 2, References 9-1 Through E-6 RNP-RA/03-0031, Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 5042003-04-28028 April 2003 Response to Request for Additional Information Re Application for Renewal of Operating License, Attachment III, Pages 356 - 504 ML0311207052003-04-18018 April 2003 Review of 90-day Steam Generator Tube Inservice Inspection Report for a Refueling Outage in 2001 ML0305202692003-02-15015 February 2003 Follow-up Report, Reference Event #39516 RNP-RA/03-0012, Request for Relief Pertaining to Examination Coverage Less than Essentially 100 Percent2003-02-11011 February 2003 Request for Relief Pertaining to Examination Coverage Less than Essentially 100 Percent RNP-RA/02-0172, Steam Generator Tube Plugging During Refueling Outage 212002-11-11011 November 2002 Steam Generator Tube Plugging During Refueling Outage 21 RNP-RA/02-0164, Part 17 of 17, CD-ROM Providing the Review Tool Supporting the Application for Renewal of Operating License2002-11-0606 November 2002 Part 17 of 17, CD-ROM Providing the Review Tool Supporting the Application for Renewal of Operating License ML0230402682002-09-19019 September 2002 Part 4 of 4 - Westinghouse Technology Manual, Course Outline for R-104P and Course Manual ML0221103432002-01-31031 January 2002 Caldon, Inc Engineering Report: ER-267N, Bounding Uncertainty Analysis for Thermal Power Determination at CP&L Robinson Nuclear Power Station Using the LEFM Check Plus System ML18288A3691990-10-31031 October 1990 ANF-88-054(NP)(A), PDC-3: Advanced Nuclear Fuels Corp Power Distribution Control for PWRs & Application of PDC-3 to Hb Robinson Unit 2. NRC Generic Letter 1979-451979-09-25025 September 1979 NRC Generic Letter 1979-045: Transmittal of Reports Regarding Foreign Reactor Operation Experiences 2024-06-28
[Table view] |
Text
Operated for the U.S. Department of Energy by Sandia Corporation Carlos Lopez P.O. Box 5800 Principal Member of the Technical Staff Albuquerque, NM 87185-0748 Structural and Thermal Analysis Department 6233 Phone: (505) 845-9545 Fax: (505) 844-2829 Internet: carlope@sandia.gov May 17, 2012 Gabriel Taylor (Gabriel.Taylor@nrc.gov)
Mail Stop 10 F 13 USNRC/RES/PRAB Washington, DC 20555
Dear Mr. Taylor,
Subject:
Letter report on the evaluation of cables from the HEAF fire event at the H. B. Robinson Steam Electric Plant Please find attached the evaluation of some electrical cable segments that were affected by the high energy arcing fault incident that occurred on March 28, 2010, at the H. B. Robinson Steam Electric Plant. This effort was sponsored under NRC job code N6982. Please feel free to contact me if you have any questions.
Sincerely, Carlos Lopez
Attachment:
Evaluation of Cables from the March 2010 HEAF Fire at H. B. Robinson Steam Electric Plant Exceptional Service in the National Interest
Evaluation
of
Cables
from
the
March
2010
HEAF
Fire
at
H.
B.
Robinson
Steam
Electric
Plant
(by Jason Brown and Carlos Lopez, Sandia National Laboratories*)
Background
On March 28, 2010, at the H. B. Robinson Steam Electric Plant, cable insulation failure on the 4 kV supply to Bus 5 led to an arc flash within a rigid metal conduit (see Figure 1 and Figure 2).
It is believed that the first fault was in the switchgear where the cables exited the conduit mouth.
The initial arc flash event also caused internal damage to the Unit Auxiliary Transformer (UAT) and a subsequent fire within the conduit (see Figure 3). When the fault occurred, the circuit breaker did not trip on over current as anticipated and remained closed throughout the event. A defective fuse disabled the breaker trip control circuit, which caused the fault to persist on buses 4 and 5. After the UAT failed internally and tripped on fault protection, the fault was then transferred from the UAT to the startup transformer (SUT) and continued for several seconds before the breaker was actuated. This cleared the fault and ended the first electrical fault event.
Figure 1: Photo of the conduit damage above Bus 5 After the recovery procedures were conducted, the plant operators attempted to reset the electrical generator lockout relays. This action resulted in the SUT being reconnected to the uncleared fault on the 4 kV Bus 5. A second fault lasting several seconds occurred in the back of a switchgear cubicle before the breaker tripped. During this second arc flash, both safety-related 125 Vdc battery buses developed electrical grounds that were likely caused by arc flash and fire damage. The damage caused by this arc may be observed in Figure 4.
Figure 2: Photo of the conduit damage above Bus 5 (different view)
Figure 3: Photo within Bus 5 displaying the junction between the conduit and the cabinet
Figure 4: Damage caused by the second arc Cable Shipment In an attempt to gain a better understanding of the arcing events and subsequent fires at Robinson, the US Nuclear Regulatory Commission (NRC) collected samples of cables and shipped them to Sandia National Laboratories. The cables arrived in a 1.2 m x 1.2 m x 1.2 m (4 ft X 4 ft X 4 ft) wooden box; however, only the lower portion was occupied. Padding was not used during the shipment, although the longer sections of cable were wrapped with tape or zip ties as seen in Figure 5. Since the cables were packaged in this manner, additional damage to the jacketing or the insulation materials may have occurred during transit.
Figure 5: Cable shipment with the front panel removed After removing the front panel of the shipping box, the cables were separated into individual samples (see Figure 6). Some of the samples were labeled with the tray and location along the tray (see Figure 7) and some were not labeled (see Figure 8). After the cables were removed from the shipping container, two loose tags (Figure 9) were found at the bottom of the box, but placing these tags with the appropriate samples appears to be possible, as seen in Figure 10.
Figure 6: All cables separated out from box Figure 7: Cable Item 5 with tray location labels
Figure 8: Cable Item 9 sample without labels Figure 9: Tags separated from the cable samples
Figure 10: Possible location for separated tags The cables varied in conductor count, gauge, and insulation/jacketing materials. The cables shipped to SNL are listed in Table 1. Availability of cable markings on each sample varied (e.g.,
gauge and manufacturer) and some cables were too damaged to distinguish any present writing.
Also, there were samples that did not include any jacket markings. In such instances, fields for the gauge, manufacturer, and voltage rating were left as Unknown. For the cable location in Table 1, the labels fixed to the cable by staff at Robinson were used to determine the location of the cable sample in proximity to the fire event. If there were no labels attached to the cable, the Origin was designated as Unspecified.
Table 1: Cable inventory from Robinson Item Conductor/ Length Voltage Manufacturer Origin Notes Number Gauge (ft) Rating 1 1C/Unknown Unknown 1 Unspecified Unknown Unshielded 2 1C/350 MCM Okonite 3 Unspecified 5000 V Shielded 3 1C/350 MCM Okonite 3 Unspecified 5000 V Shielded 4 1C/350 MCM Okonite 3 Unspecified 5000 V Shielded Silicon 5 2C/Unknown Unknown 16 Tray 1 Unknown Rubber Bundle of 6 3C/Unknown Unknown 10 (ea.) Tray 3 Unknown three 7 1C/750 MCM Okonite 20 Tray 2 5000 V Shielded 8 1C/750 MCM Okonite 15 Unspecified 5000 V Shielded Silicon 9 2C/Unknown Unknown 16 Unspecified Unknown Rubber 10 1C/750 MCM Okonite 20 Tray 4 5000 V Shielded
In general, each sample displayed different levels of damage severity. This may be observed in Figure 11, Figure 12, and Figure 13. Some samples displayed evidence of corrosion of the metal wrap beneath the jacket while other cables showed little damage. This difference is elaborated upon in the subsequent section.
Figure 11: Close up of exposed metal wrap and cable damage for Item 2, 3, and 4 Figure 12: Close up of Cable Item 1 Figure 13: Indication of limited cable damage
Fire
Scene
Reconstruction
As depicted in Figure 14, the tray directly above Bus 5 is denoted as Tray 1 with horizontal locations of A, B, and C from left to right. This tray is open on the top and bottom. Tray 2 is located directly above Tray 1 and is labeled similarly with A, B, and C from left to right. From the photo information package, the bottom of the tray is open and the top is covered. Given the information from the labels fixed to the samples, some of the cables were removed from these two locations. Other cables within the shipment were labeled as coming from Tray 3 and Tray 4; however, these tray locations were not depicted or defined in any of the photos provided by the NRC.
It may be observed that both trays above Bus 5 were loaded with varying amounts of cables that were not well described in the information package. Additionally, it was not clear where the shipped samples of cable were located within the tray. As an example, it is unclear if these cables were located along the rail, in the center of the tray, or top/middle/bottom layer of cables.
When comparing Figure 15, Figure 16, and Figure 17, the level of damage on the cables exposed directly to the fire appears to be much greater than the sample that was shipped to SNL. This implies that the cable shipped to SNL for analysis was not from the area of greatest heat impact.
When tested for continuity, the cable taken from Tray 1 did not display insulation degradation between conductors. Continuity measurements were made using a Fluke 87V True-RMS multimeter.
Figure 14: Labels of the cable location for Bus 5
Figure 15: Damage of cables at location B in Tray 1 Figure 16: Cables located in Tray 1 and above Bus 5
Figure 17: Cable located within Tray 1 at location B which was received from Robinson displays very limited damage Tray 2 is located above Tray 1 and, as shown in Figure 18 and Figure 19, has a covered top and open bottom. In the latter picture, the level of damage may be observed, albeit limitedly. The cable from this tray is 750 MCM in size and contains a metallic, sub-jacket wrap. The cable shipped to SNL (shown in Figure 20) displayed less damage than depicted in the photograph of the cables within the tray. Electrical continuity was checked between the copper strands and the wrap and it was determined that there was no measurable shorting between the two. These continuity measurements were also made using a Fluke 87V True-RMS multimeter.
Figure 18: Photo of the tray cover attached to Tray 2 Figure 19: Photo of cables located within Tray 2 at an undefined horizontal location Figure 20: Close up of the received cable found in Tray 2
General
Observations
and
Conclusions
All the cables supplied to SNL varied in terms of fire damage, as some displayed more thermal impact than others. Without detailed knowledge of origin, including specific tray and location within the tray, scene reconstruction was only limitedly useful. To gain some perspective on the electrical integrity, each sample was tested for continuity between adjacent conductors or the metal wrap. Although all the results indicated that the insulation maintained electrical isolation between the conductors and metal wrap, it is not certain that the samples shipped to SNL were the most damaged from each tray location. As indicated in the prior section of this report, the level of damage depicted in the photographs was not well represented by the samples received.
Two of the cable samples were located within Tray 3 and Tray 4. The information package provided by NRC, however, did not identify the position of these trays in the context of the fire scene. Because of this, additional details are necessary to properly place the cable samples from these locations and to quantify the damage from the fires. This information would also be useful to identify differences between the photos from the incident to the cables received from Robinson as well as the loading for each tray.
There were several samples of cable that were unlabeled and not described in the information package, specifically Items 1 through 4, 8, and 9. Without any description, these samples were not helpful in reconstructing the fire scene.
It was not possible to estimate the zone of influence of the HEAF event using the cables and information provided. It would be possible to use some of these cables to run limited Penlight exposure tests to gain insights on the thermal impacts received at designated cable locations.
However, since there is a lack of information, such as the location of the cable within the tray bundle, this exercise may not be as useful as anticipated. Additionally, such tests would be outside of the scope of work for this task.
- Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energys National Nuclear Security Administration under contract DE-AC04-94AL85000.