07-13-2017 | On May 14, 2017, at 0233 hrs, Indian Point Unit 3 entered Mode 4 as part of coming out of outage 3R19 and preparing for power operations. Operations test group was preparing for performance of 3-PT- CS004, Residual Heat Removal ( RHR) Check Valve Testing. The team gathered for a pre job brief in accordance with the requirements of EN-HU-102, Human Performance Traps &Tools Procedure. At the time the only allowable access point to the Inner Crane Wall was through the double gate combination of Gates D and E, which require one gate to be maintained closed and secured at all times. Workers needed to enter inside of the Crane Wall to perform a portion of the valve lineup required by 3-PT-CS004. After unbolting and opening the gate, the two operators and a contract Radiation Protection ( RP) Technician went through gate C despite a posted sign stating that the gate was not to be utilized in modes 1 through 4.
While the valve manipulations were in progress the NRC Resident Inspector was also conducting a tour of the Vapor Containment ( VC) and identified that gate C was opened. This gate being open in this plant condition resulted in a safety system functional failure, since with the gate unsecured this made the containment sumps inoperable. |
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Category:Letter
MONTHYEARML24011A1982024-01-12012 January 2024 ISFSI, Notice of Organization Change for Site Vice President ML23342A1082024-01-0909 January 2024 Independent Spent Fuel Storage Installation Security Inspection Plan ML23353A1742023-12-19019 December 2023 ISFSI, Emergency Plan, Revision 23-04 L-23-019, Proof of Financial Protection 10 CFR 140.152023-12-18018 December 2023 Proof of Financial Protection 10 CFR 140.15 ML23339A0442023-12-0505 December 2023 Issuance of Amendment No. 68, 301 and 277 Regarding Changes to Implement the Independent Spent Fuel Storage Installation Physical Security Plan ML23326A1322023-12-0505 December 2023 Issuance of Amendment No. 67, 300 & 276 to Implement the Independent Spent Fuel Storage Installation Only Emergency Plan ML23338A2262023-12-0404 December 2023 Signed Amendment No. 27 to Indemnity Agreement No. B-19 ML23356A0212023-12-0101 December 2023 American Nuclear Insurers, Secondary Financial Protection (SFP) Program ML23242A2772023-11-30030 November 2023 NRC Letter Issuance - IP LAR for Units 2 and 3 Renewed Facility Licenses and PDTS to Reflect Permanent Removal of Spent Fuel from SFPs ML23338A0482023-11-30030 November 2023 ISFSI, Report of Changes to Physical Security, Training and Qualification, Safeguards Contingency Plan, and ISFSI Security Program, Revision 28 ML22339A1572023-11-27027 November 2023 Letter - Indian Point - Ea/Fonsi Request for Exemptions from Certain Emergency Planning Requirements for 10 CFR 50.47 and 10 CFR Part 50, Appendix E IR 05000003/20230032023-11-21021 November 2023 NRC Inspection Report Nos. 05000003/2023003, 05000247/2023003, 05000286/2023003, and 07200051/2023003 ML23050A0032023-11-17017 November 2023 Letter - Issuance Indian Point Unit 2 License Amendment Request to Modify Tech Specs for Staffing Requirements Following Spent Fuel Transfer to Dry Storage ML23100A1172023-11-17017 November 2023 NRC Response - Indian Point Energy Center Generating Units 1, 2, and 3 Letter with Enclosures Regarding Changes to Remove the Cyber Security Plan License Condition ML23100A1252023-11-17017 November 2023 Letter and Enclosure 1 - Issuance Indian Point Energy Center Units 1, 2, and 3 Exemption for Offsite Primary and Secondary Liability Insurance Indemnity Agreement ML23100A1432023-11-16016 November 2023 Letter - Issuance Indian Point Energy Center Generating Units 1, 2, and 3 Exemption Concerning Onsite Property Damage Insurance (Docket Nos. 50-003, 50-247, 50-286) ML23064A0002023-11-13013 November 2023 NRC Issuance for Approval-Indian Point EC Units 1, 2 and 3 Emergency Plan and Emergency Action Level Scheme Amendments L-23-012, Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point2023-11-13013 November 2023 Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point ML23306A0992023-11-0202 November 2023 and Indian Point Energy Center, Notification of Changes in Schedule in Accordance with 10 CFR 50.82(a)(7) ML23063A1432023-11-0101 November 2023 Letter - Issuance Holtec Request for Indian Point Energy Center Generating Units 1, 2, and 3 Exemptions from Certain Emergency Planning Requirements of 10 CFR 50.47 and Part 50 ML23292A0262023-10-19019 October 2023 LTR-23-0211-RI Thomas Congdon, Executive Deputy, Department of Public Service, Chair, Indian Point Decommissioning Oversight Board, Letter Independent Spent Fuel Storage Installation Inspection and Office of the Inspector General Report-RI ML23289A1582023-10-16016 October 2023 Decommissioning International - Registration of Spent Fuel Casks and Notification of Permanent Removal of All Indian Point Unit 3 Spent Fuel Assemblies from the Spent Fuel Pit ML23270A0082023-09-27027 September 2023 Registration of Spent Fuel Casks ML23237A5712023-09-22022 September 2023 09-22-2023 Letter to Dwaine Perry, Chief, Ramapo Munsee Nation, from Chair Hanson, Responds to Letter Regarding Opposition of the Release and Dumping of Radioactive Waste from Indian Point Nuclear Power Plant Into the Hudson River ML23242A2182023-09-12012 September 2023 IPEC NRC Response to the Town of New Windsor, Ny Board Certified Motion Letter Regarding Treated Water Release from IP Site (Dockets 50-003, 50-247, 50-286) ML23250A0812023-09-0707 September 2023 Registration of Spent Fuel Casks ML23255A0142023-08-31031 August 2023 LTR-23-0211 Thomas Congdon, Executive Deputy, Department of Public Service, Chair, Indian Point Decommissioning Oversight Board, Letter Independent Spent Fuel Storage Installation Inspection and Office of the Inspector General Report IR 05000003/20230022023-08-22022 August 2023 NRC Inspection Report 05000003/2023002, 05000247/2023002, 05000286/2023002, and 07200051/2023002 ML23227A1852023-08-15015 August 2023 Request for a Revised Approval Date Regarding the Indian Point Energy Center Permanently Defueled Emergency Plan and Emergency Action Level Scheme ML23222A1442023-08-10010 August 2023 Registration of Spent Fuel Casks ML23208A1642023-07-26026 July 2023 Village of Croton-on-Hudson New York Letter Dated 7-26-23 Re Holtec Wastewater ML23200A0422023-07-19019 July 2023 Registration of Spent Fuel Casks ML23235A0602023-07-17017 July 2023 LTR-23-0194 Dwaine Perry, Chief, Ramapo Munsee Nation, Ltr Opposition of the Release and Dumping of Radioactive Waste from Indian Point Nuclear Power Plant Into the Hudson River ML23192A1002023-07-11011 July 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise the Emergency Plan and Emergency Action Level Scheme ML23194A0442023-07-11011 July 2023 Clarification for Indian Point Energy Center License Amendment Request, Independent Spent Fuel Storage Installation Physical Security Plan ML23171B0432023-06-23023 June 2023 Letter - Indian Point Energy Center - Request for Additional Information for Independent Spent Fuel Storage Installation Facility-Only Emergency Plan License Amendment ML23118A0972023-06-0606 June 2023 06-06-23 Letter to the Honorable Michael V. Lawler, Et Al., from Chair Hanson Regarding Holtec'S Announcement to Expedite Plans to Release Over 500,000 Gallons of Radioactive Wastewater from Indian Point Energy Center Into the Hudson River ML23144A3512023-05-25025 May 2023 Clementina Bartolotta of Pearl River, New York Email Against Treated Water Release from Indian Point Site ML23156A3052023-05-25025 May 2023 Transcript Per April 27, 2023 New York State Decommissioning Oversight Board Public Statement Hearing on Controlled Radiological Effluent Discharges from Indian Point ML23144A3542023-05-25025 May 2023 Terri Thal of New City, New York Email Against Treated Water Release from Indian Point Site ML23144A3532023-05-25025 May 2023 John Shaw of New York Email Against Treated Water Release from Indian Point Site ML23144A3522023-05-25025 May 2023 Loredana Bidmead of New York E-Mail Against Treated Water Release from Indian Point Site ML23144A3492023-05-25025 May 2023 Arthur Fasolino of Rockland County, New York Email Against Treated Water Release from Indian Point Site ML23144A3432023-05-25025 May 2023 Sanford Kane of New City, New York Email Against Treated Water Release from Indian Point Site 2024-01-09
[Table view] Category:Licensee Event Report (LER)
MONTHYEARNL-18-039, LER 2018-001-00 for Indian Point, Unit 2 Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection2018-05-21021 May 2018 LER 2018-001-00 for Indian Point, Unit 2 Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection 05000286/LER-2017-0042017-12-20020 December 2017 Reactor Trip Due to Main Generator Loss of Field, LER 17-004-00 for Indian Point Unit 3, Regarding Reactor Trip Due to Main Generator Loss of Field ML17252A8662017-09-0909 September 2017 Letter Regarding a 04/26/1977 Occurrence Concerning Failure of Number 22 Main Steam Line Isolation Valve to Close to a Manual Signal Initiated by the Control Room Operator - Indian Point Unit No. 2 NL-17-107, LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate fo2017-08-29029 August 2017 LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for 05000247/LER-2015-0012017-08-29029 August 2017 Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size that Results in Exceeding the Allowed Leakage Rate for Containment, LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size that Results in Exceeding the Allowed Leakage Rate for Containment 05000286/LER-2017-0032017-08-29029 August 2017 Condensate Storage Tank Declared Inoperable Per Technical Specification, LER 17-003-00 for Indian Point, Unit 3, Regarding Condensate Storage Tank Declared Inoperable Per Technical Specification 05000247/LER-2017-0032017-08-23023 August 2017 Technical Specification Violation of Section 3.3.1 RPS Instrumentation, LER 17-003-00 for Indian Point Unit 2, Regarding Technical Specification Violation of Section 3.3.1 RPS Instrumentation 05000247/LER-2017-0022017-08-22022 August 2017 Auxiliary Feedwater Flow Indication Inoperable for Longer Than the Allowed Technical Specification Completion Time Due to Failure of Complete Restoration Following Calibration, LER 17-002-00 for Indian Point, Unit 2 Regarding Auxiliary Feedwater Flow Indication Inoperable for Longer Than the Allowed Technical Specification Completion Time Due to Failure of Complete Restoration Following Calibration 05000247/LER-2017-0012017-08-22022 August 2017 Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused By Main Boiler Feedwater Pump Turbine Low Pressure Governor Valves Failed Closed, LER 17-001-00 for Indian Point, Unit 2 Regarding Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused By Main Boiler Feedwater Pump Turbine Low Pressure Governor Valves Failed Closed 05000286/LER-2017-0022017-08-0909 August 2017 Manual Isolation of Chemical and Volume Control System Normal Letdown to Stop a Valve Leak Resulted in an Exceedance of Technical Specification 3.4.9 Condition A Limit for Pressurizer Level, LER 17-002-00 for Indian Point, Unit 3 re Manual Isolation of Chemical and Volume Control System Normal letdown to Stop a Valve Leak Resulted in an Exceedance of Technical Specification 3.4.9 Condition A Limit for Pressurizer Level 05000286/LER-2017-0012017-07-13013 July 2017 Single Flow Barrier Access Point Found Unbolted, LER 17-001-00 for Indian Point, Unit 3 Regarding Single Flow Barrier Access Point Found Unbolted 05000247/LER-2016-0022017-02-28028 February 2017 Automatic Actuation of Emergency Diesel Generators (EDGs) Due to 480 VAC Bus Undervoltage Condition and Loss of Residual Heat Removal (RHR) While in Cold Shutdown, LER 16-002-01 for Indian Point, Unit 2 Regarding Automatic Actuation of Emergency Diesel Generators (EDGs) Due to 480 VAC Bus Undervoltage Condition and Loss of Residual Heat Removal (RHR) While in Cold Shutdown 05000247/LER-2016-0102017-02-28028 February 2017 Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24, Fan Cooler Unit, LER 16-010-01 for Indian Point 2 Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24 Fan Cooler Unit NL-16-108, LER 15-001-01 for Indian Point 2 Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Conta2016-09-29029 September 2016 LER 15-001-01 for Indian Point 2 Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Contai 05000286/LER-2015-0052016-09-14014 September 2016 Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by the Trip of 345kV Main Generator Output Breaker 3 due to a Failure of South Ring Bus 345kV Breaker 5, LER 15-005-01 for Indian Point 3 RE: Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by the Trip of 345kV Main Generator output Breaker 3 due to a Failure of South Ring Bus 345kV Breaker 5 05000286/LER-2015-0042016-09-14014 September 2016 Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by a Failure of the 31 Main Transformer, LER 15-004-01 for Indian Point Unit No. 3 Regarding Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by a Failure of the 31 Main Transformer 05000286/LER-2015-0072016-09-0606 September 2016 Manual Reactor Trip Due to Decreasing Steam Generator Water Levels Caused by a Miss- Wired Circuit Board in the Main Feedwater Pump Speed Control System, LER 2015-007-01 for Indian Point, Unit 3 Regarding Manual Reactor Trip Due to Decreasing Steam Generator Water Level Caused by a Miss-Wired Circuit Board in the Main Feedwater Pump Speed Control System 05000286/LER-2015-0062016-08-0808 August 2016 Technical SpecificatiOn Prohibited Condition Due to Two Pressurizer Code Safety Valves Discovered Outside their As-Found Lift Setpoint Test Acceptance Criteria, LER 15-006-01 for Indian Point Unit No. 3 Regarding Technical Specification Prohibited Condition Due to Two Pressurizer Code Safety Valves Discovered Outside Their As-Found Lift Setpoint Test Acceptance Criteria 05000286/LER-2014-0042016-08-0101 August 2016 Automatic Reactor Trip as a Result of Meeting the Trip Logic for Over Temperature Delta Temperature during Reactor Protection System Pressurizer Pressure Calibration, LER 14-004-01 for Indian Point Unit 3, Regarding Automatic Reactor Trip as a Result of Meeting the Trip Logic for Over Temperature Delta Temperature During Reactor Protection System Pressurizer Pressure Calibration 05000247/LER-2016-0042016-05-31031 May 2016 Unanalyzed Condition due to Degraded Reactor Baffle-Former Bolts, LER 16-004-00 for Indian Point 2 re Unanalyzed Condition Due to Degraded Reactor Baffle-Former Bolts 05000247/LER-2016-0052016-05-25025 May 2016 Technical Specification (TS) Prohibited Condition Due to a Surveillance Requirement Never Performed for Testing the Trip of the Main Boiler Feedwater Pumps, LER 16-005-00 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to a Surveillance Requirement Never Performed for Testing the Trip of the Main Boiler Feedwater Pumps 05000247/LER-2016-0012016-05-0202 May 2016 Technical Specification Prohibited Condition Caused by One Main Steam Safety Valve Outside Its As-Found Lift Set Point Test Acceptance Criteria, LER 16-001-00 for Indian Point 2 RE: Technical Specification Prohibited Condition Caused by One Main Steam Safety Valve Outside Its As-Found Lift Set Point Test Acceptance Criteria 05000247/LER-2015-0042016-02-18018 February 2016 Safety System Functional Failure Due to an Inoperable Containment Caused by a Flawed Elbow on the 21 Fan Cooler Unit Service Water Motor Cooling Return Pipe, LER 15-004-00 for Indian Point 2 Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Flawed Elbow on the 21 Fan Cooler Unit Service Water Motor Cooling Return Pipe 05000286/LER-2015-0082016-02-11011 February 2016 Automatic Reactor Trip Due to a Turbine-Generator Trip as a Result of a Fault on 345 kV Feeder W96 Tower Lines Caused by Pre-Existing Degraded Insulator, LER 15-008-00 for Indian Point, Unit 3, Regarding Automatic Reactor Trip Due to a Turbine-Generator Trip as a Result of a Fault on 345 kV Feeder W96 Tower Lines Caused by Bird Streaming 05000247/LER-2015-0032016-02-0303 February 2016 Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure, LER 15-003-00 for Indian Point, Unit 2, Regarding Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure NL-15-124, LER 15-001-00 for Indian Point 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Cont2015-10-0909 October 2015 LER 15-001-00 for Indian Point 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size That Results in Exceeding the Allowed Leakage Rate for Conta NL-13-166, Report on Inoperable Gross Failed Fuel Detector2013-12-20020 December 2013 Report on Inoperable Gross Failed Fuel Detector NL-13-038, Event Report for Discovery of a Condition That Prevented Immediate Protective Actions to Avoid Exposures to Radiation from Byproduct Material2013-02-19019 February 2013 Event Report for Discovery of a Condition That Prevented Immediate Protective Actions to Avoid Exposures to Radiation from Byproduct Material NL-12-060, Submittal of Report on Inoperable Gross Failed Fuel Detector2012-04-26026 April 2012 Submittal of Report on Inoperable Gross Failed Fuel Detector ML1101906402010-11-0909 November 2010 Event Notification Report; Subject: Power Reactor Indian Point Unit 2 NL-09-108, Submittal of Report on Inoperable Core Exit Thermocouples2009-08-10010 August 2009 Submittal of Report on Inoperable Core Exit Thermocouples ML0509600412004-12-17017 December 2004 Final Precursor Analysis - IP-2 Grid Loop ML0509600512004-12-17017 December 2004 Final Precursor Analysis - IP-3 Grid Loop NL-03-136, LER 03-03-00 for Indian Point Unit 3 Regarding Automatic Turbine Trip/Reactor Trip Due to Fault in 345kV Generator Output Breaker 32003-08-21021 August 2003 LER 03-03-00 for Indian Point Unit 3 Regarding Automatic Turbine Trip/Reactor Trip Due to Fault in 345kV Generator Output Breaker 3 ML0209104352002-03-19019 March 2002 LER 98-001-01 for Indian Point Unit 3 Re Potential Failure or Inadvertent Operation of Fire Protection Systems, Caused by Personnel Error in Design ML17252A8951979-05-25025 May 1979 Letter Reporting a 05/18/1973 Occurrence of a Pressure Transient within the Reactor Coolant System Due to the Closure of Certain Air Operated Valves in the Reactor Coolant Letdown System - Indian Point Unit 2 ML17252A8461974-02-19019 February 1974 Letter Regarding Performance of a Surveillance Test PT-M2 Reactor Coolant Temperature Analog Channel Functional Test - Delta T Overtemperature and T Overpower - Indian Point Unit No. 2 ML17252A8481974-02-19019 February 1974 Letter Regarding a February 1, 1974 Occurrence Where Both Door of the 80 Foot Elevation Personnel Air Lock to the Containment Building Were Inadvertently Open at the Same Time for a Period of About Thirty Seconds - Indian Point Unit. 2 ML17252A8471974-02-0808 February 1974 Letter Regarding an Occurrence on 1/25/1974 at the Indian Point Unit No. 2 Reactor Was Brought Critical in Preparation for Placing the Plant Back in Service Following Completion of Repairs Associated with No. 22 Steam Generator Feedwater Li ML17252A8491974-02-0606 February 1974 Letter Regarding an Occurrence Where Both Doors of the 80 Foot Elevation Personnel Air Lock to the Containment Building Were Inadvertently Opened at the Same Time for About Thirty Seconds - Indian Point Unit 2 ML17252A8501974-02-0505 February 1974 Letter Regarding an Occurrence Where a Slight Reactor Coolant System Pressure Transient Was Experienced in the Course of Placing a Reactor Coolant Pump in Service - Indian Point Unit No. 2 ML17252A8511974-02-0101 February 1974 Letter Regarding an Inspection of All Bergen-Paterson Hydraulic Shock and Sway Arrestors (Snubbers) Located in the Vapor Containment Was Performed and Two Did Not Meet the Established Criterion for Operability - Indian Point Unit No. 2 ML17252A8521974-01-31031 January 1974 Letter Regarding an Occurrence Where the Reactor Was Brought Critical Preparatory to Placing the Plant Back in Service Following Completion of Repairs Associated with the 11/13/1973 Feedwater Line Break Incident - Indian Point Unit No. 2 ML17252A8591974-01-28028 January 1974 Letter Regarding an Occurrence 01/23/1974 Where a Slight Reactor Coolant System Pressure Transient Above the Technical Specifications Limit Was Experienced in the Course of Placing a Reactor Coolant Pump in Service - Indian Point Unit No. 2 ML17252A8721974-01-18018 January 1974 Letter Regarding Analysis of Results of Monthly Periodic Surveillance Test PT-M11 (Steam Line Pressure Analog Channel Function Test) Indicated That One of the Low Steam Line Pressure Bistables Associated with High Steam - Indian Point Unit ML17252A8761973-12-28028 December 1973 Letter Regarding 12/17/1973 Analysis of the Results of Periodic Tests and Calibration Checks Relating to Pressurizer Level Indicated a Setpoint Drift - Indian Point Unit 2 ML17252A8771973-12-18018 December 1973 Letter Regarding a 12/17/1973 Analysis of Results of Periodic Tests and Calibration Checks Relating to Pressurizer Level Indicated a Setting for One of the Bistables Was Above the Technical Spec. Limit - Indian Point Unit 2 ML17252A8791973-12-0303 December 1973 Letter Regarding a 11/18/1973 Occurrence Relating to the Discovery of the Erroneous Setting for 1 of the Bistables Associated with Low Pressurizer Safety Injection Required by the Technical Specifications - Indian Point Unit No. 2 ML17252A8781973-11-30030 November 1973 Letter Providing Supplemental Information Concerning the 11/13/1973 Incident at Indian Point Unit No. 2 ML17252A8821973-11-19019 November 1973 Letter Concerning a 11/16/1973 Occurrence Regarding Periodic Tests and Calibration Checks Indicating the Setting for 1 of the Bistable Device Was Below the Technical Specification Requirements - Indian Point Unit 2 2018-05-21
[Table view] |
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 001 2017 - 00 a LER NUMBER Indian Point Unit 3 commenced shutdown for refueling outage 3R19 at 2000 hrs on March 12, 2017. At 0928 hrs on March 13, 2017, Mode 5, Cold Shutdown, was entered. At this point containment integrity was no longer required per Technical Specifications (TS).
On May 14, 2017, at 0233 hrs Unit 3 entered Mode 4, Hot Shutdown, in preparation for returning the unit to power operations. On May 14, 2017, dayshift, Operations Test Group was scheduled to perform 3-PT-CS004, Residual Heat Removal (RHR) Check Valve testing. This test requires operators to go inside the Vapor Containment (VC) Inner Crane Wall to perform valve manipulations. A pre job brief in accordance with EN-HU-102, the Human Performance Traps &Tools Procedure was held that discussed the job task, dose rates, stay times, heat stress and the plant being in Mode 4. A radiological brief in accordance with EN-RP-101 was also held that discussed dose rates and radiological conditions before the two operators and a radiation protection (RP) technician entered the Vapor Containment. A containment entry brief per OAP-007, Containment Entry And Egress, was not held.
Entry inside the 46ft VC Crane Wall requires personnel to go through one of four crane wall entrances.
Three entry points have a single gate (gates A, B and C) and one entry point has a double gate (gates D and E). In Mode 1, 2, 3, or 4, entry inside the Crane Wall SHALL use the double gate entry point via Gates D and E. One gate SHALL remain shut and secured at all times to maintain flow channeling and sump operability. Securing gates requires using slide latch which can be performed from inside or outside of the gate. The organization with the largest job scope requiring VC entry is responsible for coordinating the VC entry, and ensuring a brief is conducted.
OAP-007 contains a NOTE prior to step 4.1.7 stating that the responsible VC Coordinator SHALL perform the pre-entry briefing. Step 4.1.7 directs a pre-entry briefing per EN-HU-102 and directs that as a minimum certain items must be covered in this briefing. One of these items is Crane Wall entry gate requirements with direction to refer to Precautions and Limitations P&L 2.31.1.
Personnel entering the VC for 3-PT-CS004 did not receive this required briefing on VC Crane Wall Entry Gates since OAP-007 was not utilized. Therefore the two operators and the RP technician were not made aware of the requirements for gate use when going inside the crane wall.
When entering the VC, the two operators and the RP technician went directly to Gate C. This is the normal gate that had been used for entry inside the crane wall during the outage. When the gate was reached, one operator proceeded to undo the bolt used to secure Gate C. The gate contains a sign stating it is not to be used with the unit in Modes 1 through 4. Installation of the sign was a corrective action for CR-IP3-2013-00975 (While fencing the Reactor Coolant Drain Tank (RCDT) during the Locked High Radiation Area (LHRA) down-posting on initial entry after shutdown, the RP work party entered through the wrong gate (bolted gate) and left the gate unbolted until it was identified by a third party).
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 001 2017 - 00 The operator who unbolted the gate stated he did not notice the sign when he unbolted and opened the gate. The second operator stated that he did not see the sign since the RP technician was between him and the gate. After unbolting and opening the gate, the operators and the RP technician went inside the crane wall and performed the valve lineup for 3-PT-CS004. The event recollections document that the gate was unbolted and open for approximately three to five minutes.
While the two operators and the RP technician were inside the crane wall, an NRC inspector also entered the VC. The NRC inspector was briefed by Radiation Protection in accordance with EN-HU-102, OAP-007 and EN-IS-108, Working in Hot Environments. This brief was performed by the same person who gave the RP brief to the two operators and the RP technician.
The NRC inspector and another RP technician proceeded to Gate C where they discovered the gate was not secured in place. The RP technician notified the Control Room of the unsecured gate.
The Shift Manager contacted a third operator who was on 68ft VC and directed him to check the status of Gate C and secure it if it was unsecured.
At this time the two operators and the RP technician that were inside the crane wall exited through Gate C and noticed the second rad tech. It was at this point that the first operator stated he noticed the sign on Gate C stating the gate was not to be opened in Modes 1, 2, 3 and 4. The operator closed and bolted the gate.
CR-IP3-2017-02737 was initiated due to Gate C being opened above Cold Shutdown. At 1532 hrs on May 14, 2017, the Control Room made an 8-hour notification (non-emergency event) on the NRC Hotline due to Unit 3 VC 4611 Gate C being unbolted in Mode 4. This is a condition which could have prevented the fulfillment of the safety function of structures or systems needed to mitigate the consequences of an accident.
Stand downs were conducted with the dayshift and nightshift operations crews. Field Senior Reactor Operator (SRO) oversight was implemented for all testing and evolutions.
Cause of Event
Direct Cause: Operators removed the bolt securing Gate C in the closed position and opened the gate to gain entry to the inner crane wall area in performance of 3-PT-CS004.
Root Cause: The root cause of the event is — Lack of Commitment to Program Implementation — The standards and procedural requirements of OAP-007 are not uniformly adhered to or enforced by all levels of the organization. Specifically, the requirements of OAP-007 regarding the requirements to have a VC comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nro.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 001 2017 - 00 coordinator for VC entries as well as the responsibilities of the VC coordinator are either not understood or not adhered to. Additionally, the significance of the Crane Wall gates and the requirement to use the double gate access when above Mode 5 have not been fully integrated into the thought and work processes of all work groups that enter the crane wall. This is evidenced by the recurrence of issues with Crane Wall gates despite multiple Apparent Cause Analyses and subsequent corrective actions performed by several groups. This caused the condition since an OAP-007 VC Entry brief was not performed so the entry team was not aware of the restriction to use only Gates D and E and the added restriction that one of these two gates must be closed and secured at all times. (F003) Contributing Cause #1 — The 1st contributing cause is Inadequate Task Barrier — The bolt on Gate C was not sufficient to overcome a Human Performance (HU) Error and prevent opening Gate C in a Mode where it was required to remain closed and secured. A better method of securing the gate by using a controlled lock instead of a bolt would have made it much more difficult for operators to gain access through this gate. (FP08) Contributing Cause #2 — The 2nd contributing cause is Poor use of HU Tools — Collectively the team exhibited poor use of HU tools. The group was focused on the task at hand and missed additional requirements that were in effect based on plant condition (Tunnel Vision). The operators and RP that entered the VC assumed it was okay to use Gate C because it was the gate they had been using to access the inner Crane Wall throughout the outage (Wrong Assumptions / Habit Intrusion). The operators and RP failed to notice and read the sign on Gate C stating that the gate was not to be opened in Modes 1, 2, 3 and 4 (Unawareness).
Corrective Actions
The following corrective actions have been or will be performed under Entergy's Corrective Action Program to address the Cause:
- Site clock reset issued to inform the entire site of the event and lessons learned from the event.
- Gate C was closed and secured. All other gates in the Unit 3 46ft VC Crane wall were verified to be closed and secured.
- Stand downs were held with the dayshift and nightshift operations crews to discuss the event and the lessons learned. Field SRO oversight was put in place for testing and evolutions.
- Operator human performance shortcomings were addressed in accordance with the MARC process.
- Ops, Maintenance, Production and Projects Managers agree upon and implement standards and expectations for VC entry including who will fill the position of VC entry coordinator during Modes 1, 2, 3 and 4 and what the expectations for this position are. Incorporate the resulting standards and expectations into OAP-007.
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
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3. LER NUMBER
- Lock single access Crane Wall gates. A/B/C gates (and corresponding gates on U2) will have a common key with a label stating that these doors cannot be opened above Mode 5. Keys will be maintained in lock box in Shift Manager Office, to be issued by Shift Manager as needed.
Event Analysis
The event is reportable under 10CFR50.73(a)(2)(v)(D) as a safety system functional failure as the condition could have prevented adequate post-accident core cooling due to Design Basis Accident (DBA) debris blockage of the recirculation and/or the containment sump. An Emergency Core Cooling System (ECCS) train is inoperable if it is not capable of delivering design flow to the Reactor Coolant System (RCS). Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available. TS 3.5.2 (ECCS-Operating) requires three ECCS trains to be operable in Modes 1, 2 and 3, and TS 3.5.3 (ECCS-Shutdown) requires one ECCS RHR subsystem and one ECCS recirculation subsystem to be operable in Mode 4. The licensing and design basis of the ECCS per Updated Final Safety Analysis Report Section 6.2.2 (ECCS System Design and Operation) credits flow channeling barriers installed in containment in response to the resolution of Generic Letter GL-2004-02. The single flow barrier gate that was used for access had no latching mechanism to prevent it from being forced open during a DBA. The unbolted gate was not in accordance with design and not a sufficient robust barrier to prevent debris from entering the Internal Recirculating and Containment Sumps had a DBA occurred while in Mode 4. The condition is also reportable under 10CFR50.73(a)(2)(vii) (common cause inoperability of independent trains or channels) as the condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to (D) mitigate the consequences of an accident.
Past Similar Events
A review was performed of the past four years of Licensee Event Reports (LERs) for events that involved SSFFs and/or common cause inoperability of the Engineered Safety Feature System. The review identified two LERs. Unit 2 LER 2016-007-00 — Personnel were moving scaffolding from inside the crane wall to areas outside the crane wall through the two open barrier gates. Having both sump barrier gates open violated ECCS operability basis which requires the sump barrier system to be operable in Modes 1 through 4. This was reported as a safety system functional failure as the condition could have prevented adequate post-accident core cooling due to DBA debris blockage of the recirculation and/or the containment sump. This event was reportable under 10 CFR 50.73(a)(2)(y)(D) and comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct-or sponsor, and a person is not required to respond to, the information collection.
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3. LER NUMBER
10 CFR 50.73(a)(2)(vii). The apparent cause was a latent organizational weakness associated with the .
used of procedure OAP-007 which had not been communicated well within the organization.
Unit 3 LER-2013-002-00 — Reported a common cause inoperability of the Emergency Core Cooling System while in Hot Shutdown due to violation of the design basis for Containment sump debris barrier, which was reportability under 10 CFR 50.73(a)(2)(v) and 10 CFR 50.73(a)(2)(vii). Radiation Protection (RP) personnel entered the reactor containment building to install plastic RP fencing for the RCDT. The apparent causes were an inadequate pre-job brief and inadequate procedure for Containment Entry and Egress (OAP-007, 0-RP-RWP-405) due to poor change management. The pre job brief did not cover the requirement to use the dual sump barrier gate access point when in Modes 1-4, nor did it address the type of fencing allowed.
Safety Significance
There were no actual safety consequences while Gate C was unbolted because there were no accidents or transients during the time period the gate were unbolted and Gate C was immediately secured following identification. Gate C was open and unsecured for a period of 3 to 5 minutes. Therefore, this event had no effect on the health and safety of the public. The risk if no action is taken is Medium and there is a likely chance of reoccurrence if adequate correction actions are not put in place. This was based on referring to EN-HU-104, Engineering Task Risk & Rigor and utilizing Attachments 9.1 Consequence Risk Factors, 9.2 Human Performance Risk Factors, and 9.3 Process Risk Factors. Using Attachment 9.1, the Risk Level associated with having an operability issue affecting multiple trains of safety related systems is considered Medium. Using these attachments and the justification above regarding the potential consequences of this event, this risk level was identified.
The immediate/interim/mitigating actions to reduce the frequency or consequence are:
- Gate C was immediately bolted closed.
- Management Associated Results Company (MARC) process was implemented for the individuals involved in this event.
- Operating crews were coached on the issue.