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| {{Adams | | {{Adams |
| | number = ML111590912 | | | number = ML111960460 |
| | issue date = 06/08/2011 | | | issue date = 07/15/2011 |
| | title = IR 05000348-11-012, 05000364-11-012, on 11/10/2010 - 05/24/2011, Joseph M. Farley Nuclear Plant, Unit 1, NRC Inspection Report and Preliminary White Finding | | | title = Summary of Public Meeting with Joseph M. Farley Nuclear Plant, to Discuss Safety Significance of Preliminary White Finding Associated with One Apparent Violation Documented in NRC Inspection Report 05000348-11-012 and 05000364-11-012 |
| | author name = Croteau R P | | | author name = Shaeffer S |
| | author affiliation = NRC/RGN-II/DRP | | | author affiliation = NRC/RGN-II/DRP/RPB2 |
| | addressee name = Stinson L M | | | addressee name = Stinson L |
| | addressee affiliation = Southern Nuclear Operating Co, Inc | | | addressee affiliation = Southern Nuclear Operating Co, Inc |
| | docket = 05000348, 05000364 | | | docket = 05000348, 05000364 |
| | license number = NPF-002, NPF-008 | | | license number = NPF-002, NPF-008 |
| | contact person = | | | contact person = |
| | case reference number = EA-11-112 | | | case reference number = IR-11-012 |
| | document report number = IR-11-012
| | | document type = Letter, Meeting Summary, Slides and Viewgraphs |
| | document type = Inspection Report, Letter, Enforcement Action | | | page count = 46 |
| | page count = 13 | |
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| {{#Wiki_filter: Enclosure(s) transmitted herewith contains(s) SUNSI. When separated from enclosure(s) this transmittal document is decontrolled. June 8, 2011 EA-11-112 Mr. L. Michael Stinson Vice President - Farley Southern Nuclear Operating Company, Inc. 7388 North State Highway 95 Columbia, AL 36319 | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION uly 15, 2011 |
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| SUBJECT: JOSEPH M. FARLEY NUCLEAR PLANT, NRC INSPECTION REPORT 05000348/2011-012 AND 05000364/2011-012 AND PRELIMINARY WHITE FINDING | | ==SUBJECT:== |
| | PUBLIC MEETING SUMMARY - JOSEPH M. FARLEY NUCLEAR PLANT - |
| | DOCKET NOS. 50-348 AND 50-364 |
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| ==Dear Mr. Stinson:== | | ==Dear Mr. Stinson:== |
| This letter transmits a finding for the Joseph M. Farley Nuclear Plant, Unit 2, which has preliminarily been determined to be White, i.e., a finding with low to moderate increased safety significance that may require additional NRC inspections. As described in the enclosed inspection report, the finding involves the failure to maintain the configuration of the 1A RCP oil lift pump system in accordance with plant design and drawings. This resulted in an electrical short on November 10, 2010, that caused a fire on the Unit 1 main control room (MCR) 1A reactor coolant pump (RCP) board handswitch. The finding affected both units due to the common control room. For Unit 1, the risk was preliminarily determined to be of very low safety significance (Green), since that unit was shut down at the time. However, the predominant risk factor for the event was associated with Unit 2 since it was operating at the time. Final disposition of the issue for Unit 1 will be made in conjunction with the final significance determination for Unit 2. NRC Inspection Report Number 05000348, 364/2010005 (ADAMS Accession Number ML110280059), dated January 27, 2011, provides additional details regarding the staff's review of this matter. Subsequent in-office and on-site inspections were completed on May 24, 2011, as documented in the enclosed inspection report. This issue was discussed on June 8, 2011, with Mr. T. Youngblood and other members of your staff. | | This refers to the Regulatory Conference conducted on July 12, 2011, in Atlanta, GA. The purpose of the Regulatory Conference was to provide opportunities to discuss the safety significance of the preliminary White finding associated with one Apparent Violation that was documented in NRC Inspection Report 05000348/2011012 and 364/2011012 (ML111590912). |
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| This finding was assessed based on the best available information, including appropriate assumptions, using the applicable Significance Determination Process (SDP). The NRC took into consideration the testing that your staff performed to recreate the fire conditions when performing the risk assessment. The finding was preliminarily determined to be of low to moderate safety significance (White) for Unit 2. The final resolution of this finding will convey SNC 2 the increment in the importance to safety by assigning the corresponding color, i.e., White, a finding with low to moderate increased importance to safety that may require additional NRC inspections. The fire event, caused by the mis-wiring of the 1A RCP oil lift pump pressure switch, was considered to be potentially challenging (i.e. if not suppressed, the fire could have eventually led to a MCR evacuation). The basis for assuming the fire could be potentially challenging was based upon the subjective classification criteria of NUREG-6850, Fire PRA Methodology for Nuclear Power Plants, Appendix C, Section C.2.3.2, Subjective Classification Criteria. Specifically, this was considered to be a fire requiring active intervention to prevent spread. The SDP analysis is included as Enclosure 2. Although the Unit 2 finding has potential safety significance, it does not present an immediate safety concern because you implemented corrective actions that included, but were not limited to, correcting the mis-wiring on the 1A RCP oil lift pump pressure switch and replacing the damaged 1A RCP handswitch on the MCR board. The finding is also an apparent violation of Technical Specification 5.4.1 as discussed in the enclosed inspection report, and is being considered for escalated enforcement action in accordance with the Enforcement Policy, which can be found on the NRC's Web site at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html. In addition, this finding is considered to have a cross-cutting aspect related to the Work Practices component of the Human Performance area in that personnel proceeded with work despite uncertainty and unexpected circumstances [H.4(a)]. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, we intend to complete our risk evaluations using the best available information and issue our final determination of safety significance within 90 days of this letter. The SDP encourages an open dialogue between the staff and the licensee; however, the dialogue should not impact the timeliness of the staff's final determination. Before we make a final decision on this matter, we are providing you with an opportunity to (1) attend a Regulatory Conference where you can present to the NRC your perspective on the facts and assumptions the NRC used to arrive at the finding and assess its significance, or (2) submit your position on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held within 30 days of the receipt of this letter and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. If a Regulatory Conference is held, it will be open for public observation. If you decide to submit only a written response, such submittal should be sent to the NRC within 30 days of your receipt of this letter. If you decline to request a Regulatory Conference or submit a written response, you relinquish your right to appeal the final SDP determination, in that by not doing either, you fail to meet the appeal requirements stated in the Prerequisite and Limitation sections of Attachment 2 of IMC 0609. Please contact Mr. Scott Shaeffer at (404) 997-4521 within 10 business days from the issue date of this letter to notify the NRC of your intentions. If we have not heard from you within 10 business days, we will continue with our significance determination and enforcement decision. The final resolution of this matter will be conveyed in separate correspondence. SNC 3 Since the NRC has not made a final determination in this matter, no Notice of Violation is being issued for this inspection finding at this time. In addition, please be advised that the number and characterization of the apparent violation may change as a result of further NRC review. Additionally, if you disagree with the cross-cutting aspect assigned to the finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Joseph M. Farley Nuclear Plant. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, Enclosure 1, and your response (if you choose to provide one), will be made available electronically for public inspection in the NRC Public Document Room or from ADAMS, accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. However, because of the security-related information contained in Enclosure 2, and in accordance with 10 CFR 2.390, a copy of Enclosure 2 will not be available for public inspection. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
| | The findings dealt with the failure to maintain the configuration of the 1A reactor coolant pump (RCP) oil lift pump system in accordance with plant design and drawings. This resulted in an electrical short on November 10, 2010 that caused a fire on the Unit 1 main control room (MCR) |
| | 1A RCP board handswitch. This conference also addressed whether enforcement action is warranted for the associated Apparent Violation. |
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| Sincerely,/RA/ Richard P. Croteau, Director Division of Reactor Projects Docket No.: 50-348, 50-364 License No.: NPF-2, NPF-8 Enclosure(s): 1. NRC Inspection Report 05000348/2011012, 05000364/2011012 2. SDP Phase 3 Summary ( ) SNC 3 Since the NRC has not made a final determination in this matter, no Notice of Violation is being issued for this inspection finding at this time. In addition, please be advised that the number and characterization of the apparent violation may change as a result of further NRC review. Additionally, if you disagree with the cross-cutting aspect assigned to the finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Joseph M. Farley Nuclear Plant. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, Enclosure 1, and your response (if you choose to provide one), will be made available electronically for public inspection in the NRC Public Document Room or from ADAMS, accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. However, because of the security-related information contained in Enclosure 2, and in accordance with 10 CFR 2.390, a copy of Enclosure 2 will not be available for public inspection. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
| | In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter will be available electronically for public inspection in the NRC Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). |
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| Sincerely,/RA/ Richard P. Croteau, Director Division of Reactor Projects Docket No.: 50-348, 50-364 License No.: NPF-2, NPF-8 Enclosure(s): 1. NRC Inspection Report 05000348/2011012, 05000364/2011012 2. SDP Phase 3 Summary ( )
| | ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). |
| X PUBLICLY AVAILABLE G NON-PUBLICLY AVAILABLE G SENSITIVE X NON-SENSITIVE ADAMS: G Yes ACCESSION NUMBER:______ML111590912________ OFFICE RII:DRP RII:DRP RII:DRP RII:DRS RII:EICS RII:DRP SIGNATURE SDR /RA/ Via email Via email Vai email Via email SMS /RA/ NAME SRose ECrowe JSowa WRogers CEvans SShaeffer DATE 06/08/2011 06/08/2011 06/08/2011 06/08/2011 06/08/2011 06/07/2011 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICIAL RECORD COPY DOCUMENT NAME: G:\DRPII\RPB2\FARLEY\REPORTS\2011IIR\2011-012\PUBLIC 2011-012 RCP HANDSWITCH FIRE CHOICE LETTER.DOCX SNC 4 cc w/encl: B. D. McKinney, Jr. Regulatory Response Manager Southern Nuclear Operating Company, Inc.
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| Electronic Mail Distribution M. J. Ajluni Nuclear Licensing Director Southern Nuclear Operating Company, Inc.
| | Should you have any questions concerning this meeting, please contact me at (404) 997-4521. |
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| Electronic Mail Distribution T. D. Honeycutt Regulatory Response Supervisor Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Todd L. Youngblood Plant Manager Joseph M. Farley Nuclear Plant Electronic Mail Distribution L. P. Hill Licensing Supervisor Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Jeffrey T. Gasser Chief Nuclear Officer Southern Nuclear Operating Company, Inc. Electronic Mail Distribution R. L. Gladney Licensing Engineer Southern Nuclear Operating Company, Inc. Electronic Mail Distribution N. J. Stringfellow Licensing Manager Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Paula Marino Vice President Engineering Southern Nuclear Operating Company, Inc. Electronic Mail Distribution J. L. Pemberton SVP & General Counsel-Ops & SNC Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Chris Clark Commissioner Georgia Department of Natural Resources Electronic Mail Distribution John G. Horn Site Support Manager Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Ted V. Jackson Emergency Response and Radiation Program Manager Environmental Protection Division Georgia Department of Natural Resources Electronic Mail Distribution Tom W. Pelham Performance Improvement Supervisor Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Cynthia A. Sanders Radioactive Materials Program Manager Environmental Protection Division Georgia Department of Natural Resources Electronic Mail Distribution James C. Hardeman Environmental Radiation Program Manager Environmental Protection Division Georgia Department of Natural Resources Electronic Mail Distribution William D. Oldfield Principal Licensing Engineer Joseph M. Farley Nuclear Plant Electronic Mail Distribution (cc w/encl 1 continued next page) SNC 5 (cc w/encl 1 continued) Mr. Mark Culver Chairman Houston County Commission P. O. Box 6406 Dothan, AL 36302 James A. Sommerville Program Coordination Branch Chief Environmental Protection Division Georgia Department of Natural Resources Electronic Mail Distribution James L. McNees, CHP Director Office of Radiation Control Alabama Dept. of Public Health P. O. Box 303017 Montgomery, AL 36130-3017 State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 30317 Montgomery, AL 36130-3017 L. L. Crumpton Administrative Assistant, Sr. Southern Nuclear Operating Company, Inc.
| | Sincerely, |
| | /RA/ |
| | Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos.: 50-348, 50-364 License Nos.: NPF-2, NPF-8 |
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| Electronic Mail Distribution F. Allen Barnes Director Environmental Protection Division Georgia Department of Natural Resources Electronic Mail Distribution SNC 5 Letter to L. Michael Stinson from Richard P. Croteau dated June 8, 2011
| | ===Enclosures:=== |
| | 1. List of Attendees 2. NRC Agenda 3. SNC Powerpoint Presentation |
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| SUBJECT: JOSEPH M. FARLEY NUCLEAR PLANT, NRC INSPECTION REPORT 05000348/2011-012 AND 05000364/2011-012 AND PRELIMINARY WHITE FINDING Distribution w/encl: RidsNrrPMFarley Resource C. Evans, RII L. Douglas, RII OE Mail RIDSNRRDIRS PUBLIC Enclosure 1 U.S. NUCLEAR REGULATORY COMMISSION REGION II Docket No.: 50-348, 50-364 License No..: NPF-2, NPF-8 Report No.: 05000348/2011-012 AND 05000364/2011-012 Licensee: Southern Nuclear Operating Company, Inc.
| | REGION II== |
| | 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 FARLEY REGULATORY CONFERENCE Atlanta, GA July 12, 2011 Name (Print) Title and Organization |
| | _____________J. Munday_____________ Division Director NRC RII/DRS |
| | _____________R. Nease ______________ _Branch Chief NRC RII/DRS/EB2_______ |
| | _____________G. McCoy______________ _Branch Chief NRC RII/DRP/RPB5______ |
| | _____________L. Suggs ______________ _Inspector NRC RII/DRS/EB2__________ |
| | _____________R. Fanner______________ _Inspector NRC RII/DRS/EB2__________ |
| | _____________G. Wiseman____________ _Senior Inspector NRC RII/DRS/EB2____ |
| | _____________D. Chung______________ NRC NRR/DRA/PRA Operational Support |
| | _____________E. Crowe______________ _Senior Resident Farley NRC RII/DRP___ |
| | _____________W. Rogers_____________ _SRA NRC RII/DRP/RPB7_____________ |
| | _____________W. Jones______________ _Deputy Division Director NRC RII/DRP_ |
| | _____________R. Croteau_____________ _Division Director NRC RII/DRP _______ |
| | _____________L. Wert________________ Deputy Regional Administrator NRC RII_ |
| | _____________S. Sparks______________ Senior Enforcement Specialist NRC RII__ |
| | _____________S. Shaeffer____________ _Branch Chief NRC RII/DRP/RPB2______ |
| | _____________S. Rose_______________ Sr. Project Engineer NRC RII/DRP/RPB2 |
| | _____________T. Lighty______________ _Project Engineer NRC RII/DRP/RPB2__ |
| | Enclosure 1 |
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| Facility: Joseph M. Farley Nuclear Plant, Unit 1 Location: Columbia, AL Dates: November 10, 2010 - May 24, 2011 Inspectors: W. Rogers, Senior Reactor Analyst (Section 4OA5) E. Crowe, Senior Resident Inspector (Section 4OA5) J. Sowa, Resident Inspector (Section 4OA5)
| | UNITED STATES NUCLEAR REGULATORY COMMISSION |
| Approved by: Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects Enclosure 1 SUMMARY OF FINDINGS IR 05000348/2011012. 05000364/2011012; 11/10/2010 - 5/24/2011; Joseph M. Farley Nuclear Plant; Unit 1; Other Activities.
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| The report transmits the results of the NRC's preliminary assessment of the 1A RCP handswitch fire. One self-revealing finding and Apparent Violation with potentially low to moderate safety significance (White) was identified. The significance of most findings is indicated by their color (great than Green, or Green, White, Yellow, Red); the significance was determined using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP); the cross-cutting aspect was determined using IMC 0310, 'Components Within The Cross-Cutting Areas;' and that findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. Cornerstone: Initiating Events * TBD. A self-revealing finding and apparent violation of Technical Specification 5.4, Procedures, was identified for failing to maintain the configuration of the 1A RCP oil lift pump system in accordance with plant design and drawings. The licensee incorrectly re-landed electrical wiring following maintenance to the 1A RCP oil lift pump pressure switch. This issue revealed itself upon the discovery of a flame on the 1A RCP handswitch in the Unit 1 main control room (MCR). The licensee's failure to maintain the configuration of the 1A RCP oil lift pump system in accordance with plant design and drawings is a performance deficiency. Work was completed, by skill of the craft, without inclusion into an amendment to the existing calibration work order, and resulted in the incoming electrical feeds for the 125 vDC and 130 vAC circuits being cross-connected and causing a fire on the MCR board when the 1A RCP handswitch was taken to start. The finding is more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events cornerstone to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, a fire occurred in the MCR for Units 1 and 2 as a result of the mis-wiring causing an electrical short in the 1A RCP handswitch. This finding was assessed using the Phase 1 screening worksheets of Appendix 4 and Appendix F of MC 0609, and warranted a review by a regional Senior Risk Analyst because a fire in the MCR had actually occurred. The regional Senior Risk Analysts determined the significance of this finding is preliminarily White. The finding does not represent an immediate safety concern because the wiring has been returned to the original plant design. The finding was assigned a cross-cutting aspect in the Work Practices component of the Human Performance area in that personnel did proceed in the face of uncertainty or unexpected circumstances. [H.4(a)] 3 Enclosure 1 4. OTHER ACTIVITIES 4OA5 Other Mis-wiring of the 1A RCP oil lift pump pressure switch results in flame on the 1A RCP handswitch. a. Inspection Scope The inspectors conducted an in-office review and significance evaluation of the events which led to the 1A RCP handswitch fire. b. Findings Introduction A self-revealing finding and apparent violation (AV) of TS 5.4, Procedures, was identified for failing to maintain the configuration of the 1A RCP oil lift pump system in accordance with plant design and drawings. The licensee incorrectly re-installed electrical wiring following maintenance to the 1A RCP oil lift pump pressure switch. This issue revealed itself upon the discovery of a flame on the 1A RCP handswitch in the Unit 1 MCR board. Description On November 10, 2010, with Unit 1 shutdown for the refueling outage (RFO), the licensee attempted to start the 1A RCP. The control room operator noticed the amber light for the switch position miss-match indication illuminate, heard an audible noise coming from the 1A RCP handswitch and noticed the presence of smoke coming from the handswitch. The control room operator and control room supervisor investigated the smoke by removing the light array from the handswitch and opening the door to the back of the MCR board panel. Each individual noticed a flame of approximately one inch in height emanating from the handswitch. The fire was extinguished by personnel blowing out the flame, after multiple attempts. The estimated fire duration was approximately one minute. Leading up to this event, the licensee had recently completed calibration of the 1A RCP oil lift pump pressure switch during the RFO. During the calibration, workers noticed damage to wiring on the oil lift pump pressure switch. The workers involved in the calibration replaced the damage conduit in the containment building and later discussed the replacement of the conduit with their supervisor. Work was completed by skill of the craft without inclusion into the calibration or other work order. The conduit replacement resulted in the removal of the pressure switch's electrical wiring from its 125 vDC and 130 vAC circuits. The licensee's event review determined that during the re-installation of the electrical wiring for the oil pressure switch, the incoming electrical feeds for the 125 vDC and 130 vAC circuits were swapped resulting in the AC and DC circuits being cross-connected creating the path for an electrical short when the associated control room 1A RCP handswitch was taken to the start position. 4 Enclosure 1 The inspectors reviewed licensee work order (WO) 1063205801 which included station procedure FNP-0-IMP-425.3, Pressure Actuated Switches (Generic). The inspectors discovered that FNP-0-IMP-425.3, section 7.4, controls the switch replacement and that step 7.4.3 requires the technician to "make note of lead locations on the pressure switch terminals then disconnect and remove the field leads." The licensee utilizes skill of the craft to ensure proper landing of electrical leads at the station. The licensee's planners also include a generic data sheet in work orders that lift and land leads to electrical components. This generic data sheet was not used since it is not formally required, but serves as an aid to the technician. The inspectors also reviewed statements from individuals involved in the lifting and landing of the wires. The individuals indicated that they "wiggled" wires on one end of the conduit as their means to locate that wire at the other end of the conduit. The licensee did not amend the original work order or re-plan the work activity in order to effect repair to the damaged conduit. The licensee entered this event into its corrective action program (CAP) as CR 201011613.
| | ==REGION II== |
| | 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 FARLEY REGULATORY CONFERENCE Atlanta, GA July 12, 2011 (Via Teleconference) |
| | Name (Print) Title and Organization |
| | _____________M. Ashley___________________ NRC HQ/NRR |
| | _____________S. Meng Wong_______________ _NRC HQ/DRA____________________________ |
| | _____________A. Klein_____________________ _NRC HQ/DRA/AFPB______________________ |
| | _____________S. Lee______________________ _NRC HQ/DRA____________________________ |
| | _____________J. Hyslop___________________ _NRC HQ/DRA/FRB_______________________ |
| | _____________N. Coleman_________________ _NRC HQ/OE_____________________________ |
| | _____________J. Circle____________________ _NRC HQ/NRR/DRA/APOB_________________ |
| | _____________R. Gallucci__________________ _NRC HQ/NRR/DRA/APLA__________________ |
| | _____________B. Martin___________________ _NRC HQ/NRR/DORL______________________ |
| | _____________D. Harrison_________________ _NRC HQ/DRA____________________________ |
| | ________________________________________ _________________________________________ |
| | ________________________________________ _________________________________________ |
| | ________________________________________ _________________________________________ |
| | ________________________________________ _________________________________________ |
| | ________________________________________ _________________________________________ |
| | ________________________________________ _________________________________________ |
| | Enclosure 1 |
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| Analysis The licensee's failure to maintain the configuration of the 1A RCP oil lift pump system in accordance with plant design and drawings is a performance deficiency. Work was incorrectly completed by skill of the craft without replanning the calibration work order. Conduct of the work directly resulted in the incoming electrical feeds for the 125 vDC and 130 vAC circuits being cross-connected. The finding is more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events cornerstone to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. An electrical short in the 1A RCP handswitch resulted in a fire occurring in the MCR for Units 1 and 2. The NRC staff determined the fire to be potentially challenging (i.e. if not suppressed, the fire could have eventually lead to a MCR evacuation). The determination that the fire was potentially challenging was based upon the subjective classification criteria of NUREG-6850, Fire PRA Methodology For Nuclear Power Plants, Appendix C, Section C.2.3.2, Subjective Classification Criteria. Specifically, this was considered to be a fire requiring active intervention to prevent spread. This finding was assessed using the Phase 1 screening worksheets of Appendix 4 and Appendix F of MC 0609, and warranted a review by a regional Senior Risk Analyst because a fire in the MCR had actually occurred. The regional Senior Risk Analysts determined the significance of this finding is preliminarily White. The finding does not represent an immediate safety concern because the wiring has been returned to the original plant design. The finding was assigned a cross-cutting aspect in the Work Practices component of the Human Performance area in that personnel did proceed in the face of uncertainty or unexpected circumstances. [H.4(a)] Enforcement TS 5.4.1 a, states in part that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A states, in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, on November 10, 2010, the licensee failed to 5 Enclosure 1 properly preplan and perform maintenance that affected safety-related equipment. Specifically, maintenance and repair activities were performed on the 1A RCP oil lift pump pressure switch wiring in the Unit 1 containment using skill of the craft techniques and without an approved amendment to WO 1063205801. During maintenance repair activities for the 125 vDC and 130 vAC circuits on the 1A RCP oil lift pump pressure switch, the proper wiring configuration was not maintained or accomplished in accordance with Drawing D-177249, Elementary Diagram Reactor Coolant Pump Bearing Lift Oil Pumps, Version 2.0. During subsequent activities to return Unit 1 to power operation on November 10, 2010, licensed operators attempted to start the 1A RCP. As a result of the mis-wiring, when the 1A RCP control handswitch was taken to the 'start' position, flame and smoke emanated from the handswitch, from the top and under the MCR board, thereby presenting a challenge to safety-related equipment inside and adjacent to the MCR board. The licensee has returned the wiring to its original plant design and the licensee has entered this issue into their corrective action program as CR 2010116613. URI 05000348/2010005-03, opened in NRC Inspection Report Number 05000348/2010005 and 0500364/2010005 is closed. Pending final significance determination, this finding is identified as AV 05000348,364/2011012-01, Flame Detected on the 1A RCP Handswitch. 4OA6 Meetings, Including Exit On June 8, 2011, the NRC presented the inspection results to Mr. T. Youngblood who acknowledged the findings. ATTACHMENT: SUPPLEMENTAL INFORMATION Attachment SUPPLEMENTAL INFORMATION LIST OF REPORT ITEMS Opened 05000348, 364/2011012-01 AV Flame Detected on the 1A RCP Handswitch (Section 4OA5)
| | Enclosure 1 Enclosure 2 |
| Closed 05000348/2010005-03 URI Flame Detected on the 1A RCP Handswitch
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| | REGULATORY CONFERENCE AGENDA SOUTHERN NUCLEAR OPERATING COMPANY JULY 12, 2011 NRC REGION II, ATLANTA, GEORGIA I. OPENING REMARKS AND INTRODUCTION II. NRC REGULATORY AND ENFORCEMENT POLICY III. STATEMENT OF ISSUE AND APPARENT VIOLATION IV. SOUTHERN NUCLEAR OPERATING COMPANY V. TRANSITION TO CLOSED SESSION (If Required) |
| | a. PUBLIC QUESTIONS b. NRC REMARKS/BREAK c. SOUTHERN NUCLEAR OPERATING COMPANY VI. BREAK/NRC CAUCUS VII. NRC FOLLOW UP QUESTIONS VIII. CLOSING REMARKS IX. PUBLIC QUESTIONS (If Not Previously Performed) |
| | Enclosure 2 |
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| | Farley Nuclear Plant NRC Regulatory Conference July 12, 2011 Mark J Ajluni, PE Nuclear Licensing Director John D Lattner, PE Principal Engineer - Fire Protection Ken McElroy RIE Program Manager Enclosure 3 |
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| | Agenda |
| | * Introductions |
| | * Objectives |
| | * Presentation |
| | * Wrap-Up and Conclusion Enclosure 3 |
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| | Objectives 1. To provide new information about the event 2. To prove the handswitch fire that occurred on Nov. 10, 2010, in the Unit 1 control room was a non-challenging fire per the guidance of NUREG/CR 6850 3. To demonstrate that the event is not risk significant and should be characterized as green Enclosure 3 |
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| | The Event |
| | * Calibration of 1A RCP oil lift system |
| | * Broken conduit found |
| | * Repairs made under PM processes versus CM |
| | * Operators later attempted start of the 1A RCP and detected burning inside the handswitch |
| | * Operator responded by opening panel and blowing on the switch box enclosure |
| | * At the same time the circuit opened and the burning stopped Enclosure 3 |
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| | Enclosure 3 |
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| | Handswitch Internals Spade Terminals Spade Terminal Region Contacts Normally Open Enclosure 3 |
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| | Handswitch in Closed Position Contacts Closed Enclosure 3 |
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| | Handswitch Spring Returns to Open Arcing Enclosure 3 |
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| | Handswitch Damage Enclosure 3 |
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| | Testing Results Because we are dealing with a specific wiring arrangement and switch all seven SNC tests of handswitch fault indicate the same location for damage and the repeatable nature of failure for this particular fault. |
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| | Enclosure 3 |
| | |
| | Handswitch Construction |
| | * Fire resistive construction of switch block |
| | * Does not melt or drip plastic |
| | * Fire resistive construction of switch housing |
| | * Objective of testing was to confirm how the switch responds to shorted conditions Enclosure 3 |
| | |
| | Main Control Board Handswitch Enclosure 3 |
| | |
| | Main Control Board Handswitch Enclosure 3 |
| | |
| | Main Control Board Handswitch Enclosure 3 |
| | |
| | Operator Intervention |
| | * The operator action had no impact on putting out the fire |
| | * Switch is well placed in fire retardant enclosure making direct air flow to the point of combustion impossible |
| | * The Condition Report assumes what the operator initially believed at the time, that he blew out the fire |
| | * Switch will self-extinguish when current path becomes interrupted |
| | * Operator action was inconsequential, testing shows the fire will self-extinguish anyway Enclosure 3 |
| | |
| | Main Control Board Enclosure 3 |
| | |
| | No Potential for Fire Spread |
| | * Handswitch fire self-extinguishes |
| | * Confirmed by testing |
| | * Handswitch fire not of sufficient duration or intensity to ignite secondary combustibles |
| | * Test thermocouples |
| | * CHRISTIFIRE test results |
| | * Fire resistive properties of cables and handswitch materials |
| | * No hot gas layer formed within panel |
| | * Conclusion: No potential for fire spread Enclosure 3 |
| | |
| | NUREG/CR-6850 Appendix C Objective Classification Criteria If any one of the following exist, the event is challenging NUREG/CR-6850 Objective Criteria FNP Event A hose stream, multiple portable fire No automatic or manual suppression used extinguishers, and/or a fixed fire suppression system (either manually or automatically actuated) were used to suppress the fire One or more components outside the No evidence of any collateral damage boundaries of the fire ignition source outside of the fire ignition source were affected Combustible materials outside the No ignition of secondary combustibles boundaries of the fire ignition source were ignited Enclosure 3 |
| | |
| | NUREG/CR-6850 Appendix C Objective Classification Criteria If any two of the following exist, the event is challenging NUREG/CR-6850 Objective Criteria FNP Event Actuation of an automatic detection Event did not produce sufficient smoke to system actuate the main control room detection system. An ionization smoke detector is located on the ceiling directly behind Section C of the MCB A plant trip was experienced Event did not cause a plant trip. The unit was in cold shutdown. |
| | |
| | A reported loss of greater than $5,000 Physical damage limited to the handswitch (< $1,000) |
| | A burning duration or suppression time of Burning duration was less than two 10 minutes or longer minutes Enclosure 3 |
| | |
| | NUREG/CR-6850 Appendix C Subjective Classification Criteria If any of the following exist, the event is Challenging NUREG/CR-6850 Subjective Criteria FNP Event It is apparent that active intervention was Operator blowing on the switch enclosure needed to prevent potential spread was insignificant intervention. Test data demonstrates the handswitch fire self-extinguishes There are indications that heat was No evidence of heat damage to any generated of sufficient intensity and components outside of the fire ignition duration to affect components outside source the fire ignition source There are indications that flames or heat No ignition of secondary combustibles was generated of sufficient intensity and duration to cause the ignition of secondary combustibles outside the fire ignition source Substantial smoke was generated Insignificant amount of smoke Enclosure 3 |
| | |
| | NUREG/CR-6850 Appendix C Subjective Classification Criteria If any of the following exist, the event is Challenging NUREG/CR-6850 Subjective Criteria FNP Event It is apparent that active intervention was Operator blowing on the switch enclosure needed to prevent potential spread was insignificant intervention. Test data demonstrates the handswitch fire self-extinguishes There are indications that heat was No evidence of heat damage to any generated of sufficient intensity and components outside of the fire ignition duration to affect components outside source the fire ignition source There are indications that flames or heat No ignition of secondary combustibles was generated of sufficient intensity and duration to cause the ignition of secondary combustibles outside the fire ignition source Substantial smoke was generated Insignificant amount of smoke Enclosure 3 |
| | |
| | NUREG/CR-6850 Appendix C Objective Classification Criteria If any two of the following exist, the event is challenging NUREG/CR-6850 Objective Criteria FNP Event Actuation of an automatic detection Event did not produce sufficient smoke to system actuate the main control room detection system. An ionization smoke detector is located on the ceiling directly behind Section C of the MCB A plant trip was experienced Event did not cause a plant trip. The unit was in cold shutdown. |
| | |
| | A reported loss of greater than $5,000 Physical damage limited to the handswitch (< $1,000) |
| | A burning duration or suppression time of Burning duration was less than two 10 minutes or longer minutes Enclosure 3 |
| | |
| | Non-Challenging Control Room Fires From Fire Events Database Involving Intervention Fire Description Type of Intervention Incident No 374 SDV high level RPS relay burned Control room personnel extinguished the burning relay 425 A relay burned due to its old age Portable CO2 extinguisher used 815 A relay burned up in the primary Portable CO2 extinguisher containment isolation panel used 2224 Defective insulation on windings led to Portable fire extinguisher fault within current protection relay used 2266 Small fire found in control panel Operator blew out flame transformer Enclosure 3 |
| | |
| | Conclusion: Non-Challenging |
| | * None of the objective or subjective criteria was met |
| | * Event is non-challenging Enclosure 3 |
| | |
| | SNC Risk Significance Determination Enclosure 3 |
| | |
| | SDP Factors HRA Control Room Abandonment Dominated Unit 1 MCR Panel (Unit 2) |
| | Suppression Non-Fire Ignition Will Fire before Fire Severity Shutdown propagation to Suppression Probability Propagate? Factor outside MCR cable bundle Probability No Risk Not Significant Unit 2 Enclosure 3 |
| | |
| | Phase III Summary Fire Propagation Suppress before MCR Shutdown CCDP Ignition Probability propagation to Abandonment outside Probability cable bundle Probability MCR NRC 1.0 0.5 1.0** 1.1E-3 (upper) 9.9E-3 5.5E-6 (white) |
| | 5.6E-4 (lower) 2.8E-6 (white) |
| | SNC 1.0 0.5* 1.0** (upper) 6.3E-5 3.1E-3 1.0E-7 (green) |
| | 0.01 (realistic) 6.4E-4 1.0E-8 (green) |
| | ** SNC Position - fire not challenging |
| | ***Considered in MCR abandonment probability Enclosure 3 |
| | |
| | MCR Abandonment - HVAC |
| | * HVAC Operation |
| | - HVAC allows Main Control Room (MCR) to remain manned for larger fires |
| | - Larger fires are less likely to occur |
| | - Risk of Abandonment reduced by a factor of 20 |
| | * Both NRC and SNC used NUREG/CR-6850 methods Enclosure 3 |
| | |
| | MCR Abandonment (cont.) |
| | |
| | * NRC assumed HVAC inoperable |
| | * SNC determined HVAC was operating normally and would continue to operate |
| | * HVAC controls/circuitry physically separated from fire location |
| | * Operators and Fire Brigade would not trip HVAC Enclosure 3 |
| | |
| | Farley Common MCR Fire Location HVAC Controls Enclosure 3 |
| | |
| | Shutdown from Outside MCR NRC Value SNC Value Comments 9.9E-3 3.1E-3 *Dominated by loss of aux feedwater due to human error |
| | *Equipment failure of AFW negligible |
| | *Difference is due to modeling of human error Enclosure 3 |
| | |
| | SDP Summary |
| | * Even assuming a challenging fire and an event probability of 1.0, the CCDP is green |
| | - HVAC was a operating normally making the likelihood of control room evacuation less |
| | - SNC performed a more realistic HRA reducing the calculated risk Enclosure 3 |
| | |
| | Corrective Actions Enclosure 3 |
| | |
| | Root Cause of the Event |
| | * No written work package generated for the replacement of the flex conduit. We stepped out of process mixing CM with PM |
| | * No lift sheet |
| | * No pre-job brief |
| | * Management has not been successful in getting Maintenance to internalize Human Performance tool usage 100% of the time. |
| | |
| | * No discussion of human performance tools when problem identified |
| | * Skill-of-the-craft is accepted behavior Enclosure 3 |
| | |
| | Corrective Actions |
| | * Revised fleet procedures to define allowable work scope for minor maintenance, tool pouch work, and CR initiation. |
| | |
| | * Maintenance personnel were trained using dynamic training methods. |
| | |
| | * Implemented leadership action plan for Maintenance Superintendents. Accomplished with the assistance of leadership expert. |
| | |
| | Enclosure 3 |
| | |
| | Conclusion and Wrap-Up |
| | * The event is non-challenging as it does not meet the NUREG / CR 6850 criteria for challenging |
| | * Tests revealed that the handswitch consistently fails at the same location, resulting in a self-extinguishing condition |
| | * Target cables will not catch fire because heat rates are low. |
| | |
| | * Modeling control room HVAC results in green risk. |
| | |
| | * The event is not risk significant |
| | * The event should be characterized as green Enclosure 3 |
| }} | | }} |
|
---|
Category:Letter
MONTHYEARIR 05000348/20240022024-08-12012 August 2024 – Integrated Inspection Report 05000348-2024002 and 05000364-2024002 and Exercise of Enforcement Discretion ML24215A3772024-08-0707 August 2024 2024 Farley Requal Inspection Corporate Notification Letter ML24201A2032024-08-0101 August 2024 Request for Withholding Information from Public Disclosure for Joseph M. Farley Nuclear Plant, Units 1 and 2, License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions ML24200A1392024-07-25025 July 2024 Regulatory Audit in Support of Review of the License Amendment Request to Revise Technical Specification 3.6.5, Containment Air Temperature ML24199A2072024-07-24024 July 2024 TS 3-6-5 LAR - Individual Notice Transmittal Letter and NSHC - Rev 1 (1) - Letter NL-24-0261, 10 CFR 50.46 ECCS Evaluation Model Annual Report for 20232024-07-19019 July 2024 10 CFR 50.46 ECCS Evaluation Model Annual Report for 2023 NL-24-0281, License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions2024-07-18018 July 2024 License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions NL-24-0227, Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2024-07-0303 July 2024 Proposed Inservice Inspection Alternative GEN-ISI-AL T-2024-03 for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML24184A0452024-07-0202 July 2024 Notification of Inspection and Request for Information for NRC Problem Identification and Resolution Inspection ML24178A3762024-07-0101 July 2024 Nonacceptance of Requested Licensing Action License Amendment Request to Revise Technical Specification 3.6.5, Containment Air Temperature, Using Risk-Informed Process for Evaluations NL-24-0234, Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown2024-06-28028 June 2024 Application to Revise Technical Specifications to Adopt TSTF-589, Eliminate Automatic Diesel Generator Start During Shutdown NL-24-0230, Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-3602024-06-28028 June 2024 Plants Units 1&2, License Amendment Request for Adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, DC Electrical Rewrite – Update to TSTF-360 NL-24-0143, Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in2024-06-27027 June 2024 Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in NL-24-0201, Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)2024-06-18018 June 2024 Proposed Inservice Inspection Alternative GEN-ISI-ALT-2024-002 for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) NL-24-0238, RIPE License Amendment Request to Change Containment Air Temperature Actions, Supplemental Information2024-06-14014 June 2024 RIPE License Amendment Request to Change Containment Air Temperature Actions, Supplemental Information NL-24-0229, Notice of Intent to Pursue Subsequent License Renewal2024-06-0707 June 2024 Notice of Intent to Pursue Subsequent License Renewal ML24149A0492024-06-0404 June 2024 SNC Fleet - Regulatory Audit in Support of Review of the License Amendment Request to Revise TS 1.1, Use and Application Definitions, and Add New Technical Specification 5.5.21 and 5.5.17, Online Monitoring Program, ML24143A0412024-05-28028 May 2024 Supplemental Information Needed for Using the Risk-Informed Process for Evaluation for the LAR to Change Technical Specification 3.6.5, Containment Air Temperature, Actions NL-24-0202, SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations2024-05-24024 May 2024 SNC Response to Regulatory Issue Summary 2024-01: Preparation and Scheduling of Operator Licensing Examinations NL-24-0191, Annual Radiological Environmental Operating Reports for 20232024-05-10010 May 2024 Annual Radiological Environmental Operating Reports for 2023 NL-24-0064, Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.52024-05-0303 May 2024 Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5 NL-24-0188, Cycle 33 Core Operating Limits Report2024-05-0101 May 2024 Cycle 33 Core Operating Limits Report NL-24-0165, Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 20232024-04-25025 April 2024 Annual Non-Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports for 2023 IR 05000348/20240012024-04-24024 April 2024 Integrated Inspection Report 05000348/2024001 and 05000364/2024001 NL-24-0137, RIPE License Amendment Request to Change Containment Air Temperature Actions2024-04-19019 April 2024 RIPE License Amendment Request to Change Containment Air Temperature Actions IR 05000348/20240102024-04-16016 April 2024 NRC Inspection Report 05000348/2024010 and 05000364/2024010 NL-24-0134, Renewed National Pollutant Discharge Elimination System (NPDES) Permit2024-04-12012 April 2024 Renewed National Pollutant Discharge Elimination System (NPDES) Permit IR 05000348/20240112024-04-12012 April 2024 Fire Protection Team Inspection Report 05000348/2024011 and 05000364/2024011 NL-24-0088, Application to Revise Technical Specifications Regarding Change to Containment Spray Nozzle Test Frequency2024-04-0505 April 2024 Application to Revise Technical Specifications Regarding Change to Containment Spray Nozzle Test Frequency NL-24-0115, Response to Request for Additional Information Exemption Requests for Physical.2024-04-0404 April 2024 Response to Request for Additional Information Exemption Requests for Physical. NL-24-0116, Nuclear Property Insurance Coverage as of April 1, 2024 and Licensee Guarantees of Payment of Deferred.2024-03-29029 March 2024 Nuclear Property Insurance Coverage as of April 1, 2024 and Licensee Guarantees of Payment of Deferred. ML24061A1692024-03-22022 March 2024 PM Reassignment Letter Farley March 2024 NL-24-0092, Inservice Inspection Program Owners Activity Report (OAR-1) for Outage 2R292024-03-15015 March 2024 Inservice Inspection Program Owners Activity Report (OAR-1) for Outage 2R29 IR 05000348/20230062024-02-28028 February 2024 Annual Assessment Letter for Joseph M. Farley Nuclear Plant, Units 1 and 2 - Report 05000348/2023006 and 05000364/2023006 NL-24-0003, Inservice Inspection Program Owners Activity Report (OAR) Supplement for Outages 1 R30 and 1 R312024-02-16016 February 2024 Inservice Inspection Program Owners Activity Report (OAR) Supplement for Outages 1 R30 and 1 R31 NL-24-0042, Response to Request for Additional Information Exemption Requests for Physical Barriers2024-02-13013 February 2024 Response to Request for Additional Information Exemption Requests for Physical Barriers IR 05000348/20230042024-02-0909 February 2024 Integrated Inspection Report 05000348/2023004 and 05000364/2023004 and Exercise of Enforcement Discretion NL-24-0033, Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers2024-02-0505 February 2024 Response to Question for Withholding Information from Public Disclosure Exemption Requests for Physical Barriers ML24010A0032024-01-30030 January 2024 Exemption from Select Requirements of 10 CFR Part 73 - Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting ML24026A0532024-01-30030 January 2024 Notification of an NRC Fire Protection Team Inspection- (Report 05000348/2024011, 05000364/2024011) and Request for Information IR 05000348/20230402024-01-24024 January 2024 95001 Supplemental Inspection Report 05000348/2023040 and Follow-Up Assessment Letter ML23341A2042024-01-12012 January 2024 Request for Additional Information Exemption Requests for Physical Barriers (EPID L-2023-LLE-0018 & L-2023-LLE-0021) NL-24-0011, Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis2024-01-11011 January 2024 Report of Changes to Emergency Plan and Summary of 50.54(q) Analysis ML23345A1312024-01-0303 January 2024 Withholding Letter - SNC Fleet - Physical Barriers Exemption (L-2023-LLE-0018 and L-2023-LLE-0021) ML23346A2222023-12-22022 December 2023 Transmittal of Dam Inspection Report - Public NL-23-0901, 30-Day 10 CFR 21 Notification - Framatome Supplied Siemens Medium Voltage (Mv) Circuit Breakers2023-12-15015 December 2023 30-Day 10 CFR 21 Notification - Framatome Supplied Siemens Medium Voltage (Mv) Circuit Breakers NL-23-0908, Cycle 30 Core Operating Limits Report2023-12-13013 December 2023 Cycle 30 Core Operating Limits Report NL-23-0877, Request for Exemption from Security Event Notification Implementation2023-11-29029 November 2023 Request for Exemption from Security Event Notification Implementation ML23228A1432023-11-22022 November 2023 Issuance of Amendment Nos. 249 and 246 to Revise TS 3.6.3, Surveillance Requirement 3.6.3.5 to Eliminate Event-Based Testing of Containment Purge Valves with Resilient Seals NL-23-0825, Reply to Notice of Violation EA-23-080 and Readiness for 95001 Inspection2023-11-14014 November 2023 Reply to Notice of Violation EA-23-080 and Readiness for 95001 Inspection 2024-08-07
[Table view] Category:Meeting Summary
MONTHYEARML24213A2462024-08-0505 August 2024 Summary of Closed Meeting SNC Fleet - Physical Barriers Exemption ML24198A0772024-07-18018 July 2024 Summary of Observation Public Meeting Held with SNC Regarding Previous LAR to Revise TS 3.6.5 Completion Time Increase Using Risk-Informed Process for Evaluation to Increase for Farley Nuclear Plant, Units 1 and 2 ML24156A1772024-06-11011 June 2024 Summary of June 6, 2024, Closed Meeting Held with Southern Nuclear Operating Company, Inc., Regarding the Physical Barrier Exemptions for Joseph M. Farley Nuclear Plant, Units 1 and 2, Edwin I. Hatch Nuclear Plant, Units 1 and 2, and Vogtle ML24138A0092024-05-17017 May 2024 Public Meeting Summary - 2023 Annual Assessment Meeting for Joseph M. Farley, Units 1 and 2 Docket No. 50-348, and 50-364 -Meeting Number 20240474 ML24093A0612024-04-18018 April 2024 April 1, 2024, Summary of Observation Pre-Submittal Meeting Held with Southern Nuclear Operating Company, Inc., Proposed Alternative Requests to the Extend Examination Intervals for Joseph M. Farley Nuclear Plant and Vogtle ML24096A1832024-04-12012 April 2024 Summary of April 3, 2024, Observation Pre-Submittal Meeting Held with Southern Nuclear Operating Company, Inc., Regarding Proposed License Amendment Request to Revise Technical Specification 3.6.5 ML24019A1932024-01-30030 January 2024 Summary of January 8, 2024, Observation Pre-Submittal Meeting Held with Southern Nuclear Operating Company Regarding a Proposed LAR for Vogtle Electric Generating Plant, Units 1, 2, 3, & 4; Hatch, Units 1 & 2; Farley, Units 1 & 2 ML23356A0282023-12-27027 December 2023 December 11, 2023, Pre-Application Meeting with SNC Regarding a Proposed Fleet LAR to Revise TS 5.3.1 to Relocate Training Qualification Requirements from TS to Their Qatr/Ndqam ML23311A1302023-11-17017 November 2023 Summary of Nov. 6, 2023, - Closed Meeting Held with Southern Nuclear Operating Company, Inc., Regarding Physical Barriers Exemptions for Hatch Units 1 & 2, Farley Units 1 & 2 and Vogtle Units 1, 2, 3 & 4 (EPID L-2023-LLE-0018 & L-2023-LLE-0 ML23146A1672023-05-31031 May 2023 Public Meeting Summary2022 Annual Assessment Meeting for Joseph M. Farley, Units 1 and 2 Docket No. 50?348, and 50?364 Meeting Number 20230610 ML23094A1092023-04-14014 April 2023 Summary of Conference Call Held with SNC, Regarding Regulatory Issue Summary 2000-11, Supplement 2, NRC Emergency Telecommunications System, for Joseph M. Farley Nuclear Plant, Units 1 and 2, and SNC Emergency Operations Facility ML22172A2142022-07-11011 July 2022 June 16, 2022, Summary of Public Meeting with Southern Nuclear Operating Company, Inc., a Proposed License Amendment Request to Modify the Technical Specifications Regarding Pressurizer Safety Valve Setpoints at the Joseph M. Farley Nuclear ML22108A1382022-04-18018 April 2022 Annual Assessment Public Meeting Summary ML21313A1172021-11-18018 November 2021 Summary of Public Meeting with Southern Nuclear Operating Company, Inc., Regarding a Proposed License Amendment Request to Change the Technical Specification Related to Technical Specification 5.5.16 for Joseph M. Farley Nuclear Plant ML21288A3312021-11-0404 November 2021 Summary of October 14, 2021, Public Meeting with SNC, Inc., Regarding Proposed LAR to Modify the TSs to Increase Containment Peak Pressure to Address Several Nuclear Safety Advisory Letters L-2021-LRM-0097 ML21176A1712021-07-0606 July 2021 Summary of June 24, 2021, Public Meeting Regarding a Proposed LAR to Permanently Remove the Encapsulation Vessels Around Containment Sump Recirculation Valves ML21124A1232021-05-0404 May 2021 April 27, 2021, Public Meeting Summary Joseph M. Farley Nuclear Plant Units 1 and 2, Edwin I. Hatch Nuclear Plant Units 1 and 2, Docket No. 50-348, 50-364, 50-321, 50-366, Meeting Number ML21117A030 ML21102A3112021-04-19019 April 2021 Corrected - 03/31/2021 Summary of Public Meeting with Southern Nuclear Operating Company, Inc., Regarding a Proposed Licensing Action Request to Change the Quality Assurance Topical Report ML21092A0082021-04-0808 April 2021 Summary of March 31, 2021, Public Meeting with Southern Nuclear Operating Company, Inc., Regarding a Proposed Licensing Action Request to Change the Technical Specification Related to Heat Flux Hot Channel Factor ML21063A3282021-03-12012 March 2021 Public Mtg Summary March 4, 2021 with Southern Nuclear Operating Company, Inc., a Proposed LAR to Change the TS Related to Heat Flux Hot Channel Factor for Joseph M. Farley Nuclear Plant, Units 1 & 2, and Vogtle Electric Generating Plant, U ML21047A2622021-02-19019 February 2021 Summary of Public Meeting with Southern Nuclear Operating Company, Inc., Regarding a Proposed License Amendment Request to Change the Core Power Distribution Monitoring System ML20318A0752020-11-16016 November 2020 Summary of November 12, 2020, Public Meeting with Southern Nuclear Operating Company, Inc., Regarding a License Amendment Request for the Emergency Plan ML20217L3692020-08-13013 August 2020 Summary of Meeting Joseph M. Farley Nuclear Plant, Units 1 and 2; Edwin I. Hatch, Units 1 and 2; and Vogtle Electric Generating Plant, Units 1, 2, 3, and 4 a LAR for the Emergency Plan (EPID Nos. L-2020-LLA0150 and L-2020-LLA-015 ML20154K7242020-06-0202 June 2020 EOC Meeting Summary ML19149A4222019-05-29029 May 2019 Summary of Public Meeting Concerning Annual Assessment of the Farley Nuclear Plant, Units 1 and 2, Docket Nos. 05000348 and 05000361 ML19059A3672019-03-0707 March 2019 March 5, 2019, Meeting Summary Regarding Pre-Application Meeting to Discuss Proposed Amendment to Revise Technical Specification 3.3.1 and 3.3.2 ML19038A3462019-02-15015 February 2019 Summary of Meeting with the Nuclear Regulatory Commission and Southern Nuclear Operating Company, Inc. Regarding Pre-Submittal of a License Amendment Request for Proposed Changes to Emergency Response Organization Staffing and Augmentation ML18318A2532018-11-26026 November 2018 Summary of Pre-Application Meeting to Discuss the Joseph M. Farley Nuclear Plant, Units 1 and 2, Spent Fuel Pool Criticality Analyses and Measurement Uncertainty Recapture Amendments ML18306A3132018-11-0707 November 2018 Summary of October 16, 2018, Meeting with Southern Nuclear Operating Company, Inc., to Discuss the Joseph M. Farley Unit 1 & 2, Electrical Distribution System as It Relates to the Risk-Informed Completion Times Amendment ML18284A0222018-10-16016 October 2018 August 23, 2018, Summary of Pre-Application Meeting with Southern Nuclear Operating Company, Inc. to Discuss TSTF-51 and TSTF-471 ML18142B0452018-05-24024 May 2018 Summary of May 21, 2018, Pre-submittal Public Teleconference with Southern Nuclear Operating Company, Inc. to Discuss the Upcoming License Amendment Request to Remove the Dedicated Shift Technical Advisor ML18110A3302018-04-20020 April 2018 Public Meeting Summary - Farley Nuclear Plant Docket Nos. 50-348, 50-364 ML17276A0862018-03-16016 March 2018 Summary of March 12, 2018, Pre-Application Meeting with Southern Nuclear Company, Inc., to Discuss a Future Risk-Informed Completion Times Amendment ML17104A1382017-04-14014 April 2017 Summary of Meeting to Discuss the Annual Assessment of Farley Nuclear Plant, Units 1 and 2 ML16237A0612016-10-0404 October 2016 Summary of Pre-Application Meeting with Southern Nuclear Operating Company to Discuss an Alternative Source Term Submittal ML16126A0812016-05-0505 May 2016 Summary of Public Meeting with Joseph M. Farley, Units 1 and 2 ML15125A0872015-05-26026 May 2015 May 12, 2015 Summary of Meeting with Southern Nuclear Operating Company to Discuss a Future Fleet Standard Emergency Plan Submittal ML15120A5942015-04-30030 April 2015 Summary of Public Meeting to Discuss 2014 Annual Performance Assessment of the Joseph M. Farley Nuclear Plant ML15103A0602015-04-30030 April 2015 April 22, 2015, Summary of Meeting with Southern Nuclear Operating Company to Discuss a Future Fleet Standard Emergency Plan Submittal ML15092A2432015-04-22022 April 2015 Summary of Telephone Conference Call Held on November 19, 2014, Between the U.S. Nuclear Regulatory Commission and DTE Electric Concerning Set 7 - November 19, 2014 ML15092A1142015-04-10010 April 2015 March 4, 2015, Summary of Meeting Between Representatives of the U.S. Army Corps of Engineers, Nuclear Regulatory Commission, and Southern Nuclear Operating Company, Inc. to Discuss Flooding Analysis Associated with Joseph M. Farley Nuclear ML15043A4592015-02-24024 February 2015 Summary of Site Audit to Support the Review of a License Amendment Request for Joseph M. Farley Nuclear Plant, Units 1 and 2, to Transition to the National Fire Protection Association Standard 805 Fire Protection Licensing Basis ML14197A5932014-08-0606 August 2014 June 26, 2014, Summary of Category 1 Public Meeting with Southern Nuclear Operating Company to Discuss Seismic Hazard Reevaluations Associated with Implementation of Japan Lessons-Learned Near-Term Task Force Recommendation 2.1 ML14133A0392014-06-0303 June 2014 Summary of Closed Meeting Between Representatives of the Army Corps of Engineers, Nuclear Regulatory Commission, and Southern Nuclear Company to Discuss Dam Failure for the Joseph Farley Nuclear Plant ML14134A0312014-05-22022 May 2014 Summary of May 16, 2014, Meeting with Southern Nuclear Operating Company to Discuss the Joseph M. Farley Flooding Hazard Reevaluation Extension Request ML14122A0372014-05-0808 May 2014 April 23, 2014 Summary of Meeting with Southern Nuclear Operating Company to Discuss Grid Design Criteria ML14108A3762014-05-0808 May 2014 Summary of Meeting with Southern Nuclear Operating Company to Discuss Deletion of the Residual Heat Removal Autoclosure Interlock ML14115A4052014-04-25025 April 2014 2013 End of Cycle Farley Public Meeting Summary Memo ML13225A2932013-08-12012 August 2013 7/23-24/2013, Summary of Meeting 2013 U.S. Nuclear Regulatory Commission (NRC) Initial Exam Writers' Workshop ML13009A2932013-01-16016 January 2013 Summary of Meeting with Tennessee Valley Authority Regarding Bellefonte Nuclear Plant, Unit 1, Fuel and Safety Analyses Methodologies 2024-08-05
[Table view] Category:Slides and Viewgraphs
MONTHYEARML24192A2262024-07-11011 July 2024 Mtg to Exchange Technical Information Regarding Farley, Units 1 & 2, Proposed License Amendment Request to Increase TS 3.6.5 Containment Temperature Limit Completion Using Risk-Informed Process for Evaluation (EPID L-2024-LRM-0086) - Slides ML24190A1082024-07-0808 July 2024 NRC Slides-07/11/2024 Mtg to Exchange Technical Information Regarding Joseph M. Farley Nuclear Plant, Units 1 & 2, Proposed License Amendment Request to Increase TS 3.6.5 Containment Temperature Limit Completion Using Risk-Informed Process ML24134A1062024-05-13013 May 2024 2023 Annual Assessment Meeting Presentation Slides ML24088A1662024-04-0303 April 2024 SNC Slides - Pre-Submittal Meeting on April 3, 2024 - Farley LAR Changing TS 3.6.5 Containment Air Temperature Using RIPE ML23361A0812024-01-0808 January 2024 SNC Slides - Pre-Submittal Public Meeting on January 8, 2024 Related to Fleet QATR ML23139A2462023-05-19019 May 2023 Annual Assessment Public Meeting Slides (May 23, 2023) ML22080A0192022-03-18018 March 2022 Annual Assessment Meeting Presentation ML21306A0242021-11-30030 November 2021 SNC Pre-Submittal Meeting for LAR to Relocate Augmented Piping Inspection Program Details from TSs to a Licensee Controlled Document for Farley Nuclear Plant, Units 1 & 2; and Vogtle Electric Generating Plant, Units 1 & 2 (EPID L-2021-LRM-0 ML21280A0012021-10-31031 October 2021 Pre-Submittal Meeting Presentation from SNC to Discuss a Future LAR Regarding Change TS to Increase Containment Peak Pressure to Address Several NSALs ML21173A0632021-06-24024 June 2021 Pre-Application SNC Presentation Slides - Removal of Encapsulations from Recirculation Suction Valves ML21083A2872021-03-24024 March 2021 Pre-Submittal Meeting - Quality Assurance Topical Report Revision to Allow Application of Remote Assessment Techniques During Exigent Conditions ML21057A0182021-02-26026 February 2021 Pre-Submittal Meeting Revise Technical Specification 3.2.1 Heat Flux Hot Channel Factor (Fq(Z)) (EPID L-2021-LRM-0023) (Slides) ML21036A1172021-02-0505 February 2021 SNC Slides - Pre-Submittal Meeting on February 16, 2021 - Beacon LAR ML20318A1032020-11-12012 November 2020 SNC ERO Staffing LAR RAI Responses ML20101K6582020-04-10010 April 2020 Annual Assessment ML20022A0262020-01-22022 January 2020 NEI Presentation Status Update for Ropte Short Term Action Plan: Removed Planned Unavailability for Mitigating Systems Performance Index ML19149A4222019-05-29029 May 2019 Summary of Public Meeting Concerning Annual Assessment of the Farley Nuclear Plant, Units 1 and 2, Docket Nos. 05000348 and 05000361 ML19057A4762019-03-0707 March 2019 Presentation Slides Related to Pre-Application Public Meeting Regarding to Technical Specification 3.3.1 and 3.3.2 (L-2019-LRM-0010) ML19025A0242019-01-25025 January 2019 SNC License Amendment Request for Proposed Changes to Emergency Response Organization Staffing and Augmentation NL-18-1412, Presentation Material for the Measurement Uncertainty Recovery and Spent Fuel Pool Criticality Analysis Pre-Submittal Meeting2018-11-0808 November 2018 Presentation Material for the Measurement Uncertainty Recovery and Spent Fuel Pool Criticality Analysis Pre-Submittal Meeting ML18290B0942018-10-16016 October 2018 October 16, 2018, Public Meeting Presentation Slides Related to July 27, 2018, License Amendment Request to Implement NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b. ML18065A0202018-03-12012 March 2018 March 12, 2018, Public Meeting Presentation ML18065A0222018-03-0606 March 2018 March 12, 2018, Public Meeting - Draft 4b Application ML18065A0252018-03-0606 March 2018 Joseph M Farley March 12, 2018, Public Meeting - Draft 4b Application Enclosures NL-16-0388, September 22, 2016, Public Meeting Presentation Slides2016-09-22022 September 2016 September 22, 2016, Public Meeting Presentation Slides ML16260A1162016-09-22022 September 2016 September 22, 2016, Public Meeting Presentation Slides ML15106A2042015-04-22022 April 2015 April 22, 2015, Meeting with Southern Nuclear Company, Slide Presentation ML14115A4052014-04-25025 April 2014 2013 End of Cycle Farley Public Meeting Summary Memo ML14120A2252014-04-23023 April 2014 April 23, 2014, Category 1 Public Meeting, J.M. Farley Presentation Slides, Residual Heat Removal Autoclosure Interlock Deletion. ML14108A0052014-04-23023 April 2014 April 23, 2014, Public Meeting Presentation Slides, Southern Nuclear Company, SNC Degraded Grid Design Criteria. ML14108A0042014-04-18018 April 2014 April 23, 2014, Category 1 Public Meeting, J.M. Farley Presentation Slides, Residual Heat Removal Autoclosure Interlock Deletion ML13079A1502013-03-20020 March 2013 Summary of Public Meeting with Joseph M. Farley ML13078A2452013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 1 (3/19) ML13078A2462013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 2 (3/19) ML13078A2472013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 2 (3/20) ML13078A2482013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 3 (3/19) ML13078A2502013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 4 ML13078A2512013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 5 (3/19) ML13078A2522013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 6 ML13078A2532013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 8 ML13078A2542013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 9 ML13078A2552013-03-19019 March 2013 03/21/2013 Licensee Meeting Slides Re Joseph M. Farley Nuclear Plant NFPA 805 Fire Modeling LAR Submittal - Session 10 IR 05000348/20110122011-07-15015 July 2011 Summary of Public Meeting with Joseph M. Farley Nuclear Plant, to Discuss Safety Significance of Preliminary White Finding Associated with One Apparent Violation Documented in NRC Inspection Report 05000348-11-012 and 05000364-11-012 ML1016705482010-06-0808 June 2010 Farley Nuclear Plant, NRC Screening Criteria Response ML1009805922010-04-0808 April 2010 Summary of Public Meeting to Discuss Nrc'S Reactor Oversight Process and Nrc'S Annual Assessment of Plant Safety Performance for Period of January 1, 2009 - December 31, 2009 - Joseph M. Farley Nuclear Plant ML0909210352009-03-19019 March 2009 Annual Assessment Meeting Presentation ML0819603642008-07-10010 July 2008 07/102008-Joseph M. Farley Regulatory Performance Meeting, Results of the NRC 95002 Supplemental Inspection ML0814901062008-05-22022 May 2008 Slides - Joseph M. Farley Nuclear Plant, Annual Public Assessment Meeting, ML0729806702007-10-24024 October 2007 Meeting Presentation on GEH Chemical Effects Testing Update ML0732413392007-09-19019 September 2007 Email from D. Decker, NRR to K. Cotton, NRR Et Al. Final Copy Farley Aitnrc Slides.Ppt 2024-07-08
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Inspection Report - Farley - 2011012 |
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UNITED STATES NUCLEAR REGULATORY COMMISSION uly 15, 2011
SUBJECT:
PUBLIC MEETING SUMMARY - JOSEPH M. FARLEY NUCLEAR PLANT -
DOCKET NOS. 50-348 AND 50-364
This refers to the Regulatory Conference conducted on July 12, 2011, in Atlanta, GA. The purpose of the Regulatory Conference was to provide opportunities to discuss the safety significance of the preliminary White finding associated with one Apparent Violation that was documented in NRC Inspection Report 05000348/2011012 and 364/2011012 (ML111590912).
The findings dealt with the failure to maintain the configuration of the 1A reactor coolant pump (RCP) oil lift pump system in accordance with plant design and drawings. This resulted in an electrical short on November 10, 2010 that caused a fire on the Unit 1 main control room (MCR)
1A RCP board handswitch. This conference also addressed whether enforcement action is warranted for the associated Apparent Violation.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter will be available electronically for public inspection in the NRC Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Should you have any questions concerning this meeting, please contact me at (404) 997-4521.
Sincerely,
/RA/
Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos.: 50-348, 50-364 License Nos.: NPF-2, NPF-8
Enclosures:
1. List of Attendees 2. NRC Agenda 3. SNC Powerpoint Presentation
REGION II==
245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 FARLEY REGULATORY CONFERENCE Atlanta, GA July 12, 2011 Name (Print) Title and Organization
_____________J. Munday_____________ Division Director NRC RII/DRS
_____________R. Nease ______________ _Branch Chief NRC RII/DRS/EB2_______
_____________G. McCoy______________ _Branch Chief NRC RII/DRP/RPB5______
_____________L. Suggs ______________ _Inspector NRC RII/DRS/EB2__________
_____________R. Fanner______________ _Inspector NRC RII/DRS/EB2__________
_____________G. Wiseman____________ _Senior Inspector NRC RII/DRS/EB2____
_____________D. Chung______________ NRC NRR/DRA/PRA Operational Support
_____________E. Crowe______________ _Senior Resident Farley NRC RII/DRP___
_____________W. Rogers_____________ _SRA NRC RII/DRP/RPB7_____________
_____________W. Jones______________ _Deputy Division Director NRC RII/DRP_
_____________R. Croteau_____________ _Division Director NRC RII/DRP _______
_____________L. Wert________________ Deputy Regional Administrator NRC RII_
_____________S. Sparks______________ Senior Enforcement Specialist NRC RII__
_____________S. Shaeffer____________ _Branch Chief NRC RII/DRP/RPB2______
_____________S. Rose_______________ Sr. Project Engineer NRC RII/DRP/RPB2
_____________T. Lighty______________ _Project Engineer NRC RII/DRP/RPB2__
Enclosure 1
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 FARLEY REGULATORY CONFERENCE Atlanta, GA July 12, 2011 (Via Teleconference)
Name (Print) Title and Organization
_____________M. Ashley___________________ NRC HQ/NRR
_____________S. Meng Wong_______________ _NRC HQ/DRA____________________________
_____________A. Klein_____________________ _NRC HQ/DRA/AFPB______________________
_____________S. Lee______________________ _NRC HQ/DRA____________________________
_____________J. Hyslop___________________ _NRC HQ/DRA/FRB_______________________
_____________N. Coleman_________________ _NRC HQ/OE_____________________________
_____________J. Circle____________________ _NRC HQ/NRR/DRA/APOB_________________
_____________R. Gallucci__________________ _NRC HQ/NRR/DRA/APLA__________________
_____________B. Martin___________________ _NRC HQ/NRR/DORL______________________
_____________D. Harrison_________________ _NRC HQ/DRA____________________________
________________________________________ _________________________________________
________________________________________ _________________________________________
________________________________________ _________________________________________
________________________________________ _________________________________________
________________________________________ _________________________________________
________________________________________ _________________________________________
Enclosure 1
Enclosure 1 Enclosure 2
REGULATORY CONFERENCE AGENDA SOUTHERN NUCLEAR OPERATING COMPANY JULY 12, 2011 NRC REGION II, ATLANTA, GEORGIA I. OPENING REMARKS AND INTRODUCTION II. NRC REGULATORY AND ENFORCEMENT POLICY III. STATEMENT OF ISSUE AND APPARENT VIOLATION IV. SOUTHERN NUCLEAR OPERATING COMPANY V. TRANSITION TO CLOSED SESSION (If Required)
a. PUBLIC QUESTIONS b. NRC REMARKS/BREAK c. SOUTHERN NUCLEAR OPERATING COMPANY VI. BREAK/NRC CAUCUS VII. NRC FOLLOW UP QUESTIONS VIII. CLOSING REMARKS IX. PUBLIC QUESTIONS (If Not Previously Performed)
Enclosure 2
Farley Nuclear Plant NRC Regulatory Conference July 12, 2011 Mark J Ajluni, PE Nuclear Licensing Director John D Lattner, PE Principal Engineer - Fire Protection Ken McElroy RIE Program Manager Enclosure 3
Agenda
- Wrap-Up and Conclusion Enclosure 3
Objectives 1. To provide new information about the event 2. To prove the handswitch fire that occurred on Nov. 10, 2010, in the Unit 1 control room was a non-challenging fire per the guidance of NUREG/CR 6850 3. To demonstrate that the event is not risk significant and should be characterized as green Enclosure 3
The Event
- Calibration of 1A RCP oil lift system
- Repairs made under PM processes versus CM
- Operators later attempted start of the 1A RCP and detected burning inside the handswitch
- Operator responded by opening panel and blowing on the switch box enclosure
- At the same time the circuit opened and the burning stopped Enclosure 3
Enclosure 3
Handswitch Internals Spade Terminals Spade Terminal Region Contacts Normally Open Enclosure 3
Handswitch in Closed Position Contacts Closed Enclosure 3
Handswitch Spring Returns to Open Arcing Enclosure 3
Handswitch Damage Enclosure 3
Testing Results Because we are dealing with a specific wiring arrangement and switch all seven SNC tests of handswitch fault indicate the same location for damage and the repeatable nature of failure for this particular fault.
Enclosure 3
Handswitch Construction
- Fire resistive construction of switch block
- Does not melt or drip plastic
- Fire resistive construction of switch housing
- Objective of testing was to confirm how the switch responds to shorted conditions Enclosure 3
Main Control Board Handswitch Enclosure 3
Main Control Board Handswitch Enclosure 3
Main Control Board Handswitch Enclosure 3
Operator Intervention
- The operator action had no impact on putting out the fire
- Switch is well placed in fire retardant enclosure making direct air flow to the point of combustion impossible
- The Condition Report assumes what the operator initially believed at the time, that he blew out the fire
- Switch will self-extinguish when current path becomes interrupted
- Operator action was inconsequential, testing shows the fire will self-extinguish anyway Enclosure 3
Main Control Board Enclosure 3
No Potential for Fire Spread
- Handswitch fire self-extinguishes
- Handswitch fire not of sufficient duration or intensity to ignite secondary combustibles
- Fire resistive properties of cables and handswitch materials
- No hot gas layer formed within panel
- Conclusion: No potential for fire spread Enclosure 3
NUREG/CR-6850 Appendix C Objective Classification Criteria If any one of the following exist, the event is challenging NUREG/CR-6850 Objective Criteria FNP Event A hose stream, multiple portable fire No automatic or manual suppression used extinguishers, and/or a fixed fire suppression system (either manually or automatically actuated) were used to suppress the fire One or more components outside the No evidence of any collateral damage boundaries of the fire ignition source outside of the fire ignition source were affected Combustible materials outside the No ignition of secondary combustibles boundaries of the fire ignition source were ignited Enclosure 3
NUREG/CR-6850 Appendix C Objective Classification Criteria If any two of the following exist, the event is challenging NUREG/CR-6850 Objective Criteria FNP Event Actuation of an automatic detection Event did not produce sufficient smoke to system actuate the main control room detection system. An ionization smoke detector is located on the ceiling directly behind Section C of the MCB A plant trip was experienced Event did not cause a plant trip. The unit was in cold shutdown.
A reported loss of greater than $5,000 Physical damage limited to the handswitch (< $1,000)
A burning duration or suppression time of Burning duration was less than two 10 minutes or longer minutes Enclosure 3
NUREG/CR-6850 Appendix C Subjective Classification Criteria If any of the following exist, the event is Challenging NUREG/CR-6850 Subjective Criteria FNP Event It is apparent that active intervention was Operator blowing on the switch enclosure needed to prevent potential spread was insignificant intervention. Test data demonstrates the handswitch fire self-extinguishes There are indications that heat was No evidence of heat damage to any generated of sufficient intensity and components outside of the fire ignition duration to affect components outside source the fire ignition source There are indications that flames or heat No ignition of secondary combustibles was generated of sufficient intensity and duration to cause the ignition of secondary combustibles outside the fire ignition source Substantial smoke was generated Insignificant amount of smoke Enclosure 3
NUREG/CR-6850 Appendix C Subjective Classification Criteria If any of the following exist, the event is Challenging NUREG/CR-6850 Subjective Criteria FNP Event It is apparent that active intervention was Operator blowing on the switch enclosure needed to prevent potential spread was insignificant intervention. Test data demonstrates the handswitch fire self-extinguishes There are indications that heat was No evidence of heat damage to any generated of sufficient intensity and components outside of the fire ignition duration to affect components outside source the fire ignition source There are indications that flames or heat No ignition of secondary combustibles was generated of sufficient intensity and duration to cause the ignition of secondary combustibles outside the fire ignition source Substantial smoke was generated Insignificant amount of smoke Enclosure 3
NUREG/CR-6850 Appendix C Objective Classification Criteria If any two of the following exist, the event is challenging NUREG/CR-6850 Objective Criteria FNP Event Actuation of an automatic detection Event did not produce sufficient smoke to system actuate the main control room detection system. An ionization smoke detector is located on the ceiling directly behind Section C of the MCB A plant trip was experienced Event did not cause a plant trip. The unit was in cold shutdown.
A reported loss of greater than $5,000 Physical damage limited to the handswitch (< $1,000)
A burning duration or suppression time of Burning duration was less than two 10 minutes or longer minutes Enclosure 3
Non-Challenging Control Room Fires From Fire Events Database Involving Intervention Fire Description Type of Intervention Incident No 374 SDV high level RPS relay burned Control room personnel extinguished the burning relay 425 A relay burned due to its old age Portable CO2 extinguisher used 815 A relay burned up in the primary Portable CO2 extinguisher containment isolation panel used 2224 Defective insulation on windings led to Portable fire extinguisher fault within current protection relay used 2266 Small fire found in control panel Operator blew out flame transformer Enclosure 3
Conclusion: Non-Challenging
- None of the objective or subjective criteria was met
- Event is non-challenging Enclosure 3
SNC Risk Significance Determination Enclosure 3
SDP Factors HRA Control Room Abandonment Dominated Unit 1 MCR Panel (Unit 2)
Suppression Non-Fire Ignition Will Fire before Fire Severity Shutdown propagation to Suppression Probability Propagate? Factor outside MCR cable bundle Probability No Risk Not Significant Unit 2 Enclosure 3
Phase III Summary Fire Propagation Suppress before MCR Shutdown CCDP Ignition Probability propagation to Abandonment outside Probability cable bundle Probability MCR NRC 1.0 0.5 1.0** 1.1E-3 (upper) 9.9E-3 5.5E-6 (white)
5.6E-4 (lower) 2.8E-6 (white)
SNC 1.0 0.5* 1.0** (upper) 6.3E-5 3.1E-3 1.0E-7 (green)
0.01 (realistic) 6.4E-4 1.0E-8 (green)
- SNC Position - fire not challenging
- Considered in MCR abandonment probability Enclosure 3
MCR Abandonment - HVAC
- HVAC allows Main Control Room (MCR) to remain manned for larger fires
- Larger fires are less likely to occur
- Risk of Abandonment reduced by a factor of 20
MCR Abandonment (cont.)
- SNC determined HVAC was operating normally and would continue to operate
- HVAC controls/circuitry physically separated from fire location
- Operators and Fire Brigade would not trip HVAC Enclosure 3
Farley Common MCR Fire Location HVAC Controls Enclosure 3
Shutdown from Outside MCR NRC Value SNC Value Comments 9.9E-3 3.1E-3 *Dominated by loss of aux feedwater due to human error
- Equipment failure of AFW negligible
- Difference is due to modeling of human error Enclosure 3
SDP Summary
- Even assuming a challenging fire and an event probability of 1.0, the CCDP is green
- HVAC was a operating normally making the likelihood of control room evacuation less
- SNC performed a more realistic HRA reducing the calculated risk Enclosure 3
Corrective Actions Enclosure 3
Root Cause of the Event
- No written work package generated for the replacement of the flex conduit. We stepped out of process mixing CM with PM
- Management has not been successful in getting Maintenance to internalize Human Performance tool usage 100% of the time.
- No discussion of human performance tools when problem identified
- Skill-of-the-craft is accepted behavior Enclosure 3
Corrective Actions
- Revised fleet procedures to define allowable work scope for minor maintenance, tool pouch work, and CR initiation.
- Maintenance personnel were trained using dynamic training methods.
- Implemented leadership action plan for Maintenance Superintendents. Accomplished with the assistance of leadership expert.
Enclosure 3
Conclusion and Wrap-Up
- The event is non-challenging as it does not meet the NUREG / CR 6850 criteria for challenging
- Tests revealed that the handswitch consistently fails at the same location, resulting in a self-extinguishing condition
- Target cables will not catch fire because heat rates are low.
- Modeling control room HVAC results in green risk.
- The event is not risk significant
- The event should be characterized as green Enclosure 3