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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000251/FIN-2018003-02 | 2018Q3 | Turkey Point | Inoperable Auxiliary Feedwater Steam Supply Flow Path | A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, was identified when FPL failed to ensure that the torque arm of the 4A steam generator (SG) auxiliary feedwater (AFW) steam supply valve, MOV-4-1403, remained engaged with its valve stem key. A disengaged torque arm subsequently caused the geared limit switch settings for the 4-1403 motor operator to become out of sync with the valve travel and rendered the AFW 4A SG supply flow path inoperable. |
| 05000424/FIN-2018003-01 | 2018Q3 | Vogtle | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 CFR Part 50.54(q)(2), required, in part, the licensee shall follow and maintain the effectiveness of its emergency plan that meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4), required, in part, a standard emergency classification and action level scheme, the bases of which include facility and system effluent parameters, is in use by the nuclear facility licensee. Contrary to the above, from January 30, 2018 to July 20, 2018, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, Units 1 and 2 procedure 19200, F-O Critical Safety Function Status Tree, version 1.0, specified over-conservative reactor coolant system (RCS) temperature values for determining a critical safety function RED Path on RCS Integrity used to evaluate emergency classification FA1 (Alert), potential loss of RCS barrier, in response to a rapid RCS cooldown event. |
| 05000250/FIN-2018003-01 | 2018Q3 | Turkey Point | Vital Inverter Alternate AC Supply Cables Were Not Included in the Nuclear Safety Capability Assessment | On June 25, 2018, the inspectors inquired about an open corrective action item documented in AR 2156812. AR 2156812 was originated by FPL on September 20, 2016, and documented that the NFPA 805 Nuclear Safety Capability Assessment (NSCA) circuit analysis failed to include and analyze cables associated with the alternate power supply to all vital inverters on either Turkey Point Unit. The vital inverters power vital plant instruments and controls and are normally powered by the vital DC batteries. The NSCA analysis incorrectly considered that the alternate AC power supply would be always available to power the vital inverters if the DC power supply was damaged by fire. However, the alternate power supply cables may be impacted by fire damage. Not correctly including the fire damage potential for the inverter alternate power supply cables resulted in a non-conservative analysis when the NSCA was performed. The inspectors inquired why compensatory measures in the form of fire watches were not established for the non-conservative NSCA analysis. In response to the inspectors questions, FPL determined that the non-conservative condition still existed and that it was potentially more than a minimal risk impact. FPL considered that if the fire Probabilistic Risk Assessment (PRA) evaluation determines the issue to not result in a risk increase of more than 1E-7/year for core damage frequency and no more than 1E-8/year for large early release frequency, that the change to the fire protection program to correctly analyze the vital inverter power supplies is no more than minimal risk impact. FPL initiated interim compensatory measures in the form of roving fire watches in all the affected Unit 3 and Unit 4 fire areas. FPL initiated AR 2270522 to document the associated interim compensatory measures. AR 2270522 also tracks completion of the necessary NSCA change and an associated fire PRA evaluation to correctly model the vital inverter power supply cables. FPL expects to complete the fire PRA evaluation in December 2018. Units 3 and 4 Operating License Condition 3.D., Transition License Conditions 1. requires, in part, that risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in Operation License Condition 3.D., Other Changes that May be Made Without Prior NRC Approval, 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact. The results of FPLs fire PRA evaluation expected to complete in December 2018 are necessary to determine if this issue is a violation of Units 3 and 4 Operating License Condition 3.D., Transition License Conditions 1. This issue remains unresolved pending review of FPLs fire PRA evaluation. |
| 05000424/FIN-2018002-02 | 2018Q2 | Vogtle | High Vibrations on Unit 2 NSCW Pump No. 3 Result in Pump Inoperability | An NRC-identified Green NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, was identified for the licensees failure to ensure that design control measures for the Unit 2 train A (2A) nuclear service cooling water (NSCW) pump no. 3 motor replacement, conducted in May 2015, adequately evaluated and addressed structural resonance of the pump, commensurate with the original pumps. As a result, the pump operated at higher than desired vibrations, since installation, causing accelerated bearing wear and premature failure of the motor in February 2018. The licensees failure to ensure that design control measures for the 2A NSCW pump no. 3 motor replacement adequately evaluated and addressed structural resonance of the pump, commensurate with the original pumps was a performance deficiency. |
| 05000424/FIN-2018002-01 | 2018Q2 | Vogtle | Failure to Adequately Load Emergency Deisel Generator (EDG) During 24-Hour Endurance Test | An NRC-identified Green NCV of Vogtle Nuclear Station TS, Section 5.4.1.a, Procedures, was identified for the licensees failure to implement the EDG 24-hour endurance surveillance procedure 14668A-1, Train A Diesel Generator Operability Test, revision 7.2, to operate the EDG as close as practicable to 3390 kVAR. Specifically, the licensee failed to carry out procedure steps and provisions that would assist in loading the EDG closer to the TS value of 3390 kVAR. The failure to follow procedure 14668A-1 and get as close as practicable to 3390 kVAR was a performance deficiency. |
| 05000425/FIN-2018001-03 | 2018Q1 | Vogtle | Inadequate Refurbishment of Emergency Diesel Generator Pneumatic Control System Logic Boards | A Green self-revealing NCV of TS Section 5.4.1.a, Procedures, was identified for the licensees failure to properly preplan and perform maintenance work on the Unit 2 B train (2B)emergency diesel generator (EDG) pneumatic control shutdown logic board. The inadequate shutdown logic board refurbishment resulted in a pneumatic control system air leak that generated an EDG shutdown signal during testing and de-energized the safety-related emergency power bus. |
| 05000321/FIN-2018001-02 | 2018Q1 | Hatch | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Hatch Nuclear Plant Technical Specification (TS) 5.7.2 states in part, areas with radiation levels greater than 1000 mRem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry.Contrary to the above, February 6, 2018, the licensee identified dose rates of 72 Rem/hr on contact, and 3.9 Rem/hr at 30 cm on the U-1 bottom head drain valve located in the 127 foot elevation of the Subpile room, in the Unit 1 Drywell. For approximately 4 hours, the entrance to the room was not locked or continuously guarded to prevent unauthorized entry as required by TS 5.7.2. Significance/Severity: The finding was of very low safety significance (Green) because it was not an as low as reasonably achievable (ALARA) planning issue, there was no overexposure nor potential for an overexposure, and the licensees ability to assess dose was not compromised.Corrective Action Reference(s):The licensee identified and documented the failure to control access to the Lock High Radiation Area (LHRA) in Condition Report 10458608. |
| 05000424/FIN-2018001-02 | 2018Q1 | Vogtle | Inadequate Acceptance Criteria for Testing of NSCW Pump Discharge Valves | An NRC identified Green NCV of 10 CFR 50.55a(f), "Inservice testing(IST)requirements," subsection (4), American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) code Subsection ISTC-5122, Stroke Acceptance Criteria, was identified for the licensees failure to incorporate adequate acceptance criteria for exercise testing of NSCW pump discharge valves into procedures. Specifically, the licensee failed to incorporate acceptance criteria for stroke close exercise testing into in-service test procedures and used inadequate reference values when determining the HIGH/LOW code allowable limits for the stroke open exercise testing. |
| 05000424/FIN-2018001-01 | 2018Q1 | Vogtle | Failure to Provide Work Instructions for Sealing Around NSCW System Pump Shaft Well Access Openings | An NRC identified Green NCV of Vogtle Electric Generating Plant Technical Specification(TS), Section 5.4.1.a, Procedures, was identified for the licensees failure to provide work instructions for the sealing of gaps around cover plates for the nuclear service cooling water (NSCW) system pumps shaft well access openings and for the failure to follow work instructions for NSCW tower clean/inspect. Specifically, the licensee failed to provide instructions for sealing around the well plate covers following well plate cover removal/reinstallation in work orders SNC737852 (Unit 1 NSCW pump #3) and SNC737853 (Unit 1 NSCW pump #5). Also, during the performance of a NSCW tower clean/inspect work order, the licensee failed to generate condition reports, as required by the work instructions, upon the discovery of cracks or gaps in the Foreign Material Exclusion (FME) barrier. As a result, gaps were left around the NSCW pumps which could allow foreign material to enter the NSCW system and adversely affect cooling water flow to essential component coolers. |
| 05000321/FIN-2018001-01 | 2018Q1 | Hatch | Failure to comply with Type B shipping container Certificate of Compliance (CoC) requirements. | An NRC Identified Green NCV of 10 Code of Federal Regulations (CFR)71.17, General license: NRC-approved package, was identified for the licensees failure to comply with the Type B shipping container Certificate of Compliance (CoC) requirements. 10 CFR 71.17(c)(2)states, in part, that a holder of a General license to utilize an NRC-approved package shall comply with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part. Specifically, on several occasions the licensee placed in transit Type B containers which did not pass the CoC leak test requirement(s). |
| 05000424/FIN-2017004-03 | 2017Q4 | Vogtle | Licensee-Identified Violation | The following violations of very low safety significance (Green) or Severity Level IV were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation.Title 10 CFR 50.55a(f), Inservice testing requirements, subsection (4) required, in part, that pumps and valves which are classified as ASME Class 1, Class 2, and Class 3 must meet the inservice test requirements set forth in the ASME OM Code. The ASME Code of record for Vogtle for Operation and Maintenance of Nuclear Power Plants (OM) is the 2004 edition through 2006 addendum. Subsection ISTC-1300, Valve Categories, required in part, that valves within this subsection shall be placed in one or more of the following categories. Category A is for valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function(s), as specified in ISTA-1100. Contrary to the above, since 1991, the licensee did not categorize valves in ECCS recirculation flow paths to the RWST as Category A valves to ensure the ASME OM test requirements were met by leak testing the valves to demonstrate that their seat leakage would limit the consequences of an accident to control room operators and to the public at the site boundary per Title 10 CFR Part 100 limits. The inspectors determined this finding was of very low safety significance (Green) because the issue would only have the potential to represent a degradation of the radiological barrier function provided for the control room. This issue was documented in the licensees CAP as CR 829367 and TE 886122. |
| 05000425/FIN-2017004-01 | 2017Q4 | Vogtle | Failure to Implement and Establish Appropriate Work Instructions for PMT of Namco Limit Switch on 2HV-8920 | A Green, self-revealing, non-cited violation (NCV) of TS 5.4.1.a, Procedures, was identified for the licensees failure to implement maintenance work instructions and establish appropriate procedures concerning the post-maintenance testing (PMT) of the Namco limit switch on Unit 2 for 2HV-8920 following removal and reinstallation of the limit switch. As a result, during ECCS interlock testing, 2HV-8804B (RHR Pump B to SI Pump B Isolation Valve) failed to open due to 2HV-8920 Namco limit switch being installed improperly. The licensees failure to perform a PMT on the Namco limit switch for 2HV-8920 following removal and reinstallation, as required by NMP-MA-014-001 (Post Maintenance Testing Guidance), was a performance deficiency (PD). The licensee reinstalled the limit switch correctly and performed the interlock testing satisfactory following the corrective maintenance. The issue was entered into the corrective action program (CAP) as condition report (CR) 10410863.The PD was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the PD affected the reliability of the ECCS valve interlock system. The finding was of very low safety significance (e.g. Green) because while logic path II (2HV-8920 and 2HV-8814) for the opening of 2HV-8804B was inoperable, the system maintained its functionality due to the availability of logic path I (2HV-8813). The inspectors determined there was no cross-cutting aspect since the finding is not indicative of current performance. |
| 05000424/FIN-2017004-02 | 2017Q4 | Vogtle | Failure to Maintain NEMA Type 4 Qualification for the Nuclear Service Cooling Water Pumps | A Green, self-revealing, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to properly implement and establish procedures to maintain watertight requirements of the nuclear service water system (NSCW) pumps motor main power cables termination box. As a result, the Unit 2 B train NSCW pump no. 4 failed due to aphase-to-ground fault caused by water and moisture intrusion into the power cable splice connections. Failure to adequately implement and establish procedures to maintain watertight requirements of the NSCW pumps motor main power cables termination box during maintenance, as required by maintenance procedures and specifications, was a performance deficiency. The licensee replaced the motor and faulted cable; and sealed all potential water and moisture intrusion enclosure locations until watertight enclosure standards are fully restored. This issue was entered into the licensees CAP as CRs10399125, 10404327, and corrective action report 270905.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (e.g. core damage). Specifically, the Unit 2 NSCW pump no. 4 was rendered inoperable, adversely affecting the NSCW system reliability. The finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of safety system function, and it did not represent a loss of function of one or more than one train for more than its Technical Specification (TS) allowed outage time or greater than 24 hrs. The finding was assigned a cross-cutting aspect of Resources, because procedures and/or work instructions were not available to maintenance personnel for properly verifying motor termination boxes were installed in compliance with NEMA 4 specifications. (H.1) |
| 05000424/FIN-2017003-04 | 2017Q3 | Vogtle | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion XI, Test Control stated, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. UFSAR Section 8.1.4.3.C.2 stated that the onsite electrical system was designed in accordance with IEEE 308 -1974, Criteria for Class 1E Power System at Nuclear Generating Stations. IEEE 308 -1974 Section 6.3 recommended periodic tests be performed at scheduled intervals to detect deterioration of equipment to demonstrate operability of the components that are not exercised during normal operation. Contrary to the above, the licensee did not establish adequate test control measures to assure that the protective function of all 1E lockout relays were periodically verified. Specifically, there was no preventative maintenance to test the 1E lockout relays for non- MSPI loads. This condition has existed since plant initial operation and was identified during a licensee Nuclear Oversight audit on July 13, 2017. The inspectors determined this finding was of very low safety significance (Green) because the inspectors found no documented history of in- service failures of 1E lockout relays rendering safety -related equipment inoperative. This issue was documented in the licensees corrective action program as CR 10381797. |
| 05000424/FIN-2017003-05 | 2017Q3 | Vogtle | Licensee-Identified Violation | 10 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Contrary to the above, on June 28, 2017, the licensee failed to evaluate radiological conditions in room 1- AB -C-94, Back flushable Filter Crud Tank Pump Room, following the tank being placed in recirculation by Operations. On July 2, 2017, during a routine survey of room 1- AB- C-94, general area dose rates in the area were found to be as high as 600 mrem/hr. On the previous survey, conducted on June 19, 2017, maximum dose rates were found to be as high as 60 mrem/hr. This finding was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety SDP, and was determined to be of very low safety significance (Green) because the finding is not related to ALARA dose planning, did not result in an overexposure or the substantial potential for overexposure, and the ability to assess dose was not compromised due to the use of appropriate personnel dosimetry. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This issue was entered into the licensees corrective action program as CR 10383067. |
| 05000424/FIN-2017003-01 | 2017Q3 | Vogtle | Failure to Implement and Establish Appropriate Work Instructions Affecting Safety-Related Chiller | A Self -Revealing, Green, non- cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow measurement and test equipment (M&TE) in support of essential safety features (ESF) chilled water pumps in- service testing (IST). As a result, the Unit 1 A train safety -related chiller was inadvertently rendered inoperable when technicians isolated a flow transmitter associated with the chillers auto -start control logic when installing and removing M&TE in support of the IST. The licensee entered this issue into their corrective action program (CAP) under condition report (CR) 10390340 and corrective action report 270610 and planned to revise the procedure. Failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow M&TE in support of ESF chilled water pumps IST, which can affect ESF chiller performance, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD). The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because while the unit 1 A train ESF chiller was rendered inoperable, it did not represent a loss of function of the train for greater than its TS Allowed Outage Time. The finding was assigned a cross cutting aspect of Challenge the Unknown because questions and risks regarding the use of flow M&TE for the test were not properly evaluated and managed before proceeding. (H.11) |
| 05000425/FIN-2017003-03 | 2017Q3 | Vogtle | Failure to Maintain Cleanliness of Motor Operated Valve Limit Switch Compartment | A Self -Revealing , Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to perform an adequate cleanliness inspection of the Unit 2 nuclear service cooling water (NSCW) system pump no. 6 discharge motor -operated -valve (MOV) limit switch compartment, as required by the maintenance procedure. As result , the valve failed to operate when demanded and rendered the NSCW pump inoperable. The failure to perform an adequate cleanliness inspection of NSCW pump no. 6 discharge MOV limit switch compartment following preventive maintenance, as required by maintenance procedure NMP -ES- 017- 008, was a performance deficiency (PD). The licensee cleaned affected MOV sub -components, verified proper operation, and restored operability of the pump. This issue was entered into the licensees CAP as CR10399054 . The performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The finding was determined to be of very low safety significance (Green) because although the performance deficiency affected the qualification and operability of the NSCW pump, it did not represent a loss of function of an NSCW train for greater than its TS Allowed Outage Time . The finding was assigned a cross cutting aspect of Avoid Complacency, because maintenance technicians did not recognize the possibility of making mistakes when performing routine tasks of inspecting and manipulating grease containing components inside the limit switch compartment. (H.12) |
| 05000425/FIN-2017003-02 | 2017Q3 | Vogtle | Failure to Maintain ECCS Flow Balance and Check Valve Inservice Test Procedure | An NRC- Identified, Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to maintain a Unit 2 surveillance procedure that demonstrated satisfactory performance of the forward flow safety function of emergency core cooling system ( ECCS ) check valves. The licensee revised and performed the test to verify satisfactory valve performance. This issue was entered into the licensees CAP as CR10410794. The failure to maintain procedure 14721D -2 to ensure test conditions that adequately demonstrated satisfactory performance of ECCS check valves 2- 1205- U6 -001/00 2, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD ). The performance deficiency was more than minor because if left uncorrected, it could result in degradation of ECCS check valves to go undetected. The finding was associated with the mitigating system cornerstone. The finding was determined to be of very low safety significance (Green) because the performance deficiency did not result in a loss of operability or functionality of ECCS check valves. The finding was assigned a cross cutting aspect of Resources, because the licensee did not ensure that an ECCS surveillance procedure was adequate to support nuclear safety . (H.1) |
| 05000424/FIN-2017002-01 | 2017Q2 | Vogtle | Failure to Correct a Condition Adverse to Quality involving an MSIV Manufacturing Deficiency | (Green). A self -revealing, Green, non -cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to identify and correct a condition adverse to quality (i.e., manufacturing deficiency), which led to a repetitive failure of main steam isolation valve ( MSIV ) 1HV -3006B. The fail ure to determine the cause of a significant condition adverse to quality and take corrective action to preclude repetition was a performance deficiency. Specifically, the licensee failed to identify the root cause of an MSIV actuator failure on April 12, 2014, that resulted in a reactor trip. As a result, appropriate corrective actions were not taken and a repeat failure of the valve actuator caused another reactor trip on February 3, 2017 . The licensee has entered this issue into the corrective action pr ogram as condition report 10326456. This performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective of limiting the likeli hood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance (Green) because the finding did not result in a loss of mitigation equipment use d to transition the reactor to a stable shutdown condition. The finding was not assigned a cross cutting aspect since it was not indicative of current licensee performance due to the root cause evaluation in question being performed greater than three years ago |
| 05000424/FIN-2017002-02 | 2017Q2 | Vogtle | Failure to Follow Work Instructions for Implementation of Open Phase Protection System | (Green). A self -revealing, Green, non -cited violation of Technical Specifications 5.4.1.a, Procedures, was identified for the licensees failure to redline new wiring installation associated with an open phase protection system modification, as required by work instructions . As result, control circuit wires were not installed per wiring diagrams and caused a loss of the offsite power feed to the B train 4160- volt emergency power bus. The licensee's failure to redline new wiring installation associated with an open phase protection system modification installation, as required by work instruction SNC804606 and 3 maintenance procedure NMP -MA -017 was a performance deficiency. The licensee entered this issue into their corrective action program under condition reports 10343972 and 10344136 and restored offsite power to the emergency bus by correcting the wiring configuration . The performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the in- service train of shutdown cooling (i.e. , 'A' train of the residual heat removal system ) was not affected. The finding was assigned a cross -cutting aspect of Procedure Adherence, in the Human Performance area becaus e individuals did not follow work instructions and redline procedures when installing new wiring for the open phase protection system (H.8) |
| 05000250/FIN-2016001-01 | 2016Q1 | Turkey Point | Failure to Fully Implement Procedure QI3-PTN-1, Design Control | A self-revealing finding was identified for the licensees failure to provide complete instructions in Maintenance Support Package (MSP) 06-053 for the Isophase Bus Enclosure Collar replacement modification in the Turkey Point switchyard. Specifically, the control power circuitry termination points in the 8W43 switchyard breaker were not identified and documented in the associated MSP for removal as required by procedure QI 3-PTN-1, Design Control. As a result, a direct current (DC) ground was introduced to the back-up protection relay by a b contact when the 8W43 breaker was opened during a planned bus switching sequence. The DC ground on the back-up protection circuitry actuated the protection relay and caused both the supply breakers for the Unit 3 startup transformer (SUT) to open resulting in a loss of off-site power (LOOP) for Unit 3. The licensee entered this performance deficiency in their corrective action program (CAP) as action request (AR) 02092653 The performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to apply procedure QI 3-PTN-1 in its entirety allowed for a DC ground to be introduced to the DC back-up protection relay circuit resulting in a LOOP. Because this finding caused a LOOP and a resultant loss of residual heat removal (RHR), a detailed risk evaluation was required per IMC-0609, Appendix G, Shutdown Operations Significance Determination Process. A Senior Reactor Analyst assessed the risk significance and concluded it was of very low safety significance (Green). The risk of the event was mitigated by the multiple means that the licensee had available to them to either: 1) restore electrical power to the safety related buses, or; 2) establish alternate means of heat removal either via the steam generators or via primary feed and bleed. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not indicative of current performance since the modification package was implemented greater than three years ago. |
| 05000400/FIN-2015008-02 | 2015Q4 | Harris | Failure to Follow EPM-410 Procedure | An NRC-identified Green NCV of 10 CFR 50.54(q)(2) was identified, for the licensees failure to follow and maintain, in effect, the Emergency Plan when performing monthly testing of the Technical Support Center (TSC). Specifically, the licensee failed to follow procedural steps when recorded values did not meet acceptance criteria as specified in EPM-410, Communication and Facility Performance Tests. The issue was placed in the licensees corrective action program as CRs 01942073, 01940053. The finding was more than minor because it was associated with the Emergency Response Organization (ERO) Performance attribute and it adversely affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the failure to follow procedural steps when recorded values did not meet acceptance criteria resulted in a failure to comply with emergency plan. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, Appendix B Emergency Preparedness Significance Determination Process. Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure to comply; Loss of Risk Significant Planning Standard Function (RSPS), NO; RSPS Degraded Function, NO; Loss of Planning Standard Function, No; results in a Green finding. The inspectors identified a cross-cutting aspect in the Problem Identification and Resolution area because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3). |
| 05000250/FIN-2015004-02 | 2015Q4 | Turkey Point | Licensee-Identified Violation | 10 CFR 50.48 states that each operating nuclear power plant must have a fire protection plan that satisfies Criterion 3 of Appendix A of this part. Turkey Point Renewed Operating License condition D, for Units 3 and 4, states that the licensee shall implement and maintain in effect all provisions of the approved FPP as described in the UFSAR Appendix 9.6A. The approved FPP is implemented, in part, by 0-ADM-016, Fire Protection Program, as referenced in Section 7.2 of UFSAR Appendix 9.6A. Section 5.6 of 0-ADM-016 requires that, for non-functional post-fire safe shutdown components, engineering evaluations should identify appropriate compensatory actions, including hourly fire roves. Contrary to the above, between May 1st, 2014, and April 23rd, 2015, hourly fire watch patrols were not conducted on numerous occasions in fire zones that required regular hourly tours due to fire protection equipment impairment. The failure to perform the fire watch tours did not cause the inoperability of any equipment but resulted in the loss of a defense-in-depth feature for fire detection in fire zones affected by an impaired or non-functional fire safety component or feature. This violation was associated with the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of the systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding to be of very low safety significance (Green) after performing a detailed risk evaluation in accordance with Manual Chapter 0609, Appendix A, because the missed fire watch tours reflected a low degradation of the Fire Prevention and Administrative Controls FPP element in that other area fire protection defense-in-depth features such as automatic fire detection (smoke detectors), automatic fire suppression capability (sprinklers), manual suppression capability (fire brigade), and safe shutdown capability from the main control room were still available. The licensee entered this violation into their CAP as AR 02056905. |
| 05000251/FIN-2015004-01 | 2015Q4 | Turkey Point | Failure to correctly follow procedure 3-PMI-072.6, Steam Dump to Atmosphere Control Loop Calibration | A self-revealing NCV of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified when the licensee failed to properly implement procedure 3-PMI-072.6, Steam Dump to Atmosphere Control Loop Calibration. Specifically, the licensee incorrectly installed a temporary electrical jumper in reactor operator console 3C02 instead of 3C04, in contrast to Step 6.3.2 of 3-PMI-072.6. This action resulted in actuation of a 3B 4160 volt (V) vital bus lockout circuit causing loss of power to the B train of Unit 3 (U3) spent fuel pool (SFP) cooling. Immediate corrective actions were taken to remove the jumper and restore the B train of SFP cooling. The licensee entered the condition in its corrective action program (CAP) as action request (AR) 02088911 and 02088914. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases. In addition, the performance deficiency, if left uncorrected, had the potential to lead to a more significant safety concern. The finding was screened using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, dated July 1, 2012, and Appendix G Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 4 for Barrier Integrity, dated May 9, 2014. The inspectors determined the finding was of very low safety significance (Green) because it was not associated with low temperature over pressurization, freeze seals, steam generator nozzle dams, criticality, drain down or leakage paths, or the containment barrier. Furthermore, one train of SFP cooling remained in operation, the rate of SFP temperature rise was low (~ 2 F/hour), and additional methods remained available to limit SFP temperature rise. This finding was assigned a cross cutting aspect associated with the procedure adherence element of the human performance area because the licensee failed to correctly execute step 6.3.2 of procedure 3-PMI-072.6. |
| 05000400/FIN-2015008-01 | 2015Q4 | Harris | Untimely 10 CFR 50.73 Notification of an Inoperable CIV | An NRC-identified Severity Level IV violation of 10 CFR 50.73 was identified for the licensees failure to provide a written report to the NRC within 60 days after discovery of a condition prohibited by Technical Specification (TS) Limited Condition for Operation (LCO) 3.6.3, "Containment Isolation Valves."The issue was placed in the licensees corrective action program as CR 01958628.The inspectors determined that the failure to provide a written report to the NRC within the time limits specified in regulations was a violation 10 CFR 50.73. The violation was evaluated using Section 6.9 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9.d.9of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV, non-cited violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue involving untimely reports to the NRC was strictly associated with a traditional enforcement violation. |
| 05000251/FIN-2015003-01 | 2015Q3 | Turkey Point | Inadequate Work Instructions for Replacing Main Generator Current Transformers | A self-revealing finding was identified for the licensees failure to provide adequate instructions for performing work on the Unit 4 main generator protection control circuitry. As a result, the lugged connections on an installed current transformer lacked the appropriate tightness causing increased electrical resistance and ultimately catastrophic failure of a lug connection. The lug failure produced an open circuit condition on the current transformer causing the generator protection circuit to actuate. This resulted in a turbine trip and reactor trip. Corrective actions included replacing the damaged lug and torqueing all the current transformer lug connections to the vendor recommended value. A root cause evaluation was performed and a revision made to maintenance procedure 0-PME-090.03, Maintenance of Isophase Neutral Bus and Grounding Transformer Connection Assemblies, to include additional instructions on torqueing the lug assemblies. The licensee entered this performance deficiency in their corrective action program (CAP) as action request 02047137. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the work package associated with engineering modification package EC 246904 and work order 40063905 directed the technician to connect the current transformer (CT) lugs hand tight and did not require torqueing per the vendor specified torque value. The inspectors screened the significance of the finding using Manual Chapter 0609, Appendix A, Exhibit 1, Transient Initiators. The inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a reactor trip and a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. The finding was associated with a cross-cutting aspect in the resources component of the human performance area because the licensee failed to ensure an adequate work instruction document was available to support nuclear safety (H.1) (Section 4OA3). |
| 05000250/FIN-2015003-02 | 2015Q3 | Turkey Point | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, required, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, after the licensee determined that the site had an inadequate relay PM program (AR 02053778), they failed to identify that the FMR for the 3B EDG needed a 10 year replacement PM. As a result, during the monthly surveillance run of the 3B EDG, the 3B EDG was rendered inoperable when the 3B EDG output breaker failed to close due to the failure of the 3B EDG FMR. The FMR for the 3B EDG had been installed since 1991. A detailed risk evaluation was performed on this licensee identified violation and was determined to be of very low risk significance, i.e., < 1E-6 (Green). The dominant risk result was a grid-related Loss of Offsite Power where multiple EDGs fail and neither offsite power nor the EDGs are recovered. This violation was associated with the Mitigating Systems Cornerstone and determined to be of very low safety significance (Green) after performing a detailed risk evaluation in accordance with Manual Chapter 0609, Appendix A. The licensee entered this violation into their CAP as AR 02024373. |
| 05000251/FIN-2015002-01 | 2015Q2 | Turkey Point | Inadequate General Operating Procedure to Prevent Inadvertent AFAS While Performing a Reactor Plant Planned Shutdown | A self-revealing non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures, was identified for the licensees failure to maintain adequate guidance in procedure 4-GOP-103, Power Operation to Hot Standby. Specifically, 4-GOP-103 did not contain adequate instructions to control reactor power prior to opening the reactor trip breakers in order to minimize steam generator inventory loss to prevent an auxiliary feed water (AFW) system actuation. As a result, the AFW actuation system (AFAS) actuated unexpectedly during a planned unit shutdown resulting in an excessive reactor coolant system cool down and the operators closing the main steam isolation valves. Corrective actions included entering this issue into their corrective action program (CAP) and revising the procedure to reduce reactor power to at least 20 percent to prevent steam generator inventory loss due to shrinkage following a manual reactor trip during a planned reactor plant shutdown from power operations to hot standby. The performance deficiency was more than minor because it is associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to have specific guidance in procedure 4-GOP-103 to ensure reactor power is lowered to at least 20 percent prior to initiating a manual reactor trip during a planned shutdown resulted in an inadvertent AFAS actuation, reactor coolant system cool down, closing of the main steam isolation valves, and a reduced safe shutdown margin. The inspectors screened the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding was associated with a cross-cutting aspect in the resources component of the human performance area because the licensee failed to ensure an adequate general operating procedure was available to support nuclear safety. |
| 05000250/FIN-2015001-02 | 2015Q1 | Turkey Point | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion III, Design Control, required, in part, that measures shall be established to assure that applicable regulatory requirements and the plants design basis are correctly translated into drawings, procedures, and instructions and that these measures shall provide for verifying or checking the adequacy of the design. Contrary to the above, the licensee did not translate the design basis requirement, the 4A ECC fan auto-start upon failure to run of the 4C ECC fan, into the 4C ECC breaker replacement instructions and procedures. As a result, during the replacement of the 4C ECC fan breaker, the 4A ECC fan auto-start control power leads were disconnected and not re-terminated upon completion of the work. Consequently, the 4A and 4C ECC fans were unable to automatically start on a Train B safety injection (SI) signal for a period of approximately 25 hours. This violation was associated with the Barrier Integrity Cornerstone and determined to be of very low safety significance (Green) because the loss of this equipment or function by itself, did not represent an actual open pathway in the physical integrity of the reactor containment, containment isolation, or heat removal components, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The licensee entered this violation into their CAP as AR 1990010. |
| 05000251/FIN-2015007-03 | 2015Q1 | Turkey Point | Required Appendix R Instrumentation Not Functional on Unit 4 Alternate Shutdown Panel | The inspectors identified an URI regarding the processes and procedures used to evaluate the impacts of Appendix R steam generator (S/G) pressure indicators (PIs), when two of three PIs used for the Unit 4 alternate shutdown panel (ASP) 4C264 were designated as non-functional for approximately ten months. Specifically, the lack of the two S/G PIs during a fire event that requires main control evacuation may have adverse impacts on the ability to safely shutdown the plant and the effects of this condition may not have been evaluated. The USFAR, Revision C26, Appendix 9.6A, Fire Protection Program Report, Section 5.0, Alternate Shutdown Capability, stated, in part, that instrumentation and controls to achieve and maintain hot standby are provided on the ASP and supplemented by manual actions at local stations for achieving cold shutdown. Table 9.6A-2 in Section 5.0, lists components, instruments, and controls required for alternate shutdown. This table included PI-3(4)-1606/-1607/-1608, S/G pressure for A/B/C respectively. On January 29, 2015, the inspectors identified there were three ARs/works requests (WRs) on 4-PI-1606 and 4-PI-1607. These were two of the three required Appendix R S/G PIs on the Unit 4 ASP. The ARs were initiated on October 16, 2013; April 24, 2014; and July 13, 2014. At the time of discovery, the licensee did not have compensatory actions in place for this condition. The licensee captured the inspectors concerns in their corrective action program as AR 02027171, and initiated an apparent cause evaluation. As a result, the licensee performed a calibration check on 4-PI-1607 on February 21, 2015, under work order (WO) 40316782-01 and identified that the surveillance was satisfactory. In addition, they performed corrective maintenance on 4-PI-1606 on February 25, 2015, under WO 40262270-02 and returned the PI to functional status. Based upon the two non-functional S/G PIs on Unit 4 ASP for approximately ten months, the inspectors requested additional information, including the completed apparent cause evaluation, to determine if the licensee followed their processes and procedures required for Appendix R equipment. This issue is unresolved pending further licensee analysis to resolve the issue and to determine if a performance deficiency exists. This issue is identified as URI 5000251/2015007-02, Required Appendix R Instrumentation Not Functional on Unit 4 Alternate Shutdown Panel. |
| 05000251/FIN-2015001-01 | 2015Q1 | Turkey Point | Untimely 10 CFR 50.72 Notification of a ECC System Functional Failure | The NRC identified a SL IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the Unit 4 emergency containment cooler (ECC) system was not reported to the NRC within eight hours of the time of discovery, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. This issue was subsequently reported to the NRC in accordance with10 CFR 50.72(b)(3)(v), and entered into the corrective action program (CAP) as condition report AR 01990555. Because the issue impacted the regulatory process; in that, a safety system functional failure was not reported to the NRC within the required timeframe thereby delaying the NRCs opportunity to review the matter, the inspectors evaluated the issue in accordance with the traditional enforcement process. Using example 6.9.d.9 from the NRC Enforcement Policy, dated February 4, 2015, the inspectors determined that the violation was a SL IV (more than a minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because the licensee failed to make a report required by 10 CFR 50.72. In accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, dated January 24, 2013, traditional enforcement issues are not assigned cross-cutting aspects. |
| 05000250/FIN-2014005-04 | 2014Q4 | Turkey Point | Failure to Fully Implement Emergency Operating Procedure 3-EOP-ES-0.1, Reactor Trip Response | A self-revealing non-cited violation (NCV) of TS 6.8.1, Procedures, was identified when the licensee failed to fully implement procedure 3-EOP-ES-0.1, Reactor Trip Response. Specifically, the licensee failed to take effective action to implement Step 25 of 3-EOP-ES-0.1 and maintain pressurizer pressure and level within their required bands in order to stabilize plant conditions following a loss of instrument air and a reactor plant trip. Corrective actions included training licensed operators on the implementation of EOP-ES- 0.1. The licensee entered this performance deficiency in their corrective action program as action request 1983618. The performance deficiency was more than minor because it was associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors screened the issue under the initiating events cornerstone using Attachment 4 (June 19, 2012) and Exhibit 1 (June 19, 2012) of Appendix A to IMC 0609, Significance Determination Process (June 2, 2011). The inspectors concluded that a detailed risk evaluation would be required by a senior reactor analyst (SRA) because the finding was associated with a transient initiator and operator actions to utilize equipment to mitigate the associated plant transient. The NRC model for Turkey Point was adjusted by setting the failure probability of the power-operated relief valve (PORV) to remain closed during an event equal to 1.0. This represented the impact of failing to follow the emergency operating procedures resulting in lifting the PORVs during the event. The change in core damage frequency results were below the 1E-6 threshold and the issue was thus determined to be of very low risk significance (Green). This finding was associated with a cross-cutting aspect in the training component of the human performance area because the licensee failed to ensure licensed operator training provided knowledge that the reactor coolant pump seals could operate for a short period of time without seal flow (H.9). |
| 05000250/FIN-2014005-05 | 2014Q4 | Turkey Point | Licensee-Identified Violation | Title 10 CFR 50.55a g (5) (ii), states, in part, that if a revised ISI program for a facility conflicts with the TS for the facility, the licensee shall apply to the Commission for amendment of the TSs to conform the TS to the revised program. Contrary to the above, from February 22, 2004, to October 03, 2014, the licensee failed to apply to the Commission for an amendment of the TSs when the revised ISI program for Turkey Point, Units 3 and 4, conflicted with the TS for the facility. Specifically, TS 3/4.7.6, Snubbers, conflicted with the revised ISI program for dynamic restraints (snubbers) because the TS did not reflect the latest American Society of Mechanical Engineers (ASME) Code of record for the current (Fourth) ISI interval (ASME Section XI, 1998 Edition with 2000 Addenda). This violation was dispositioned through the traditional enforcement process because the failure to submit a TS amendment impacted a regulatory process in that, the NRC was not able to perform its regulatory function in determining the adequacy of a licensed activity. This violation was determined to be of Severity Level IV because it is consistent with Example 6.1.d.3 in the NRC Enforcement Policy (revised July 9, 2013), Reactor Operations. As a result of a self-assessment, this violation was entered in the licensees CAP as AR 01984462. |
| 05000250/FIN-2014005-01 | 2014Q4 | Turkey Point | Non-Compliance with HRA Entry Requirements | A self-revealing NCV of Technical Specification (TS) 6.12.1, High Radiation Area, was identified when a worker did not comply with a radiological barrier and entered a high radiation area (HRA) without proper authorization. Specifically, on March 24, 2014, a worker entered a HRA without a survey meter, without being made aware of radiological conditions in the area, and without a health physics technician (HPT) escort and subsequently received a dose rate alarm. Upon identification, the licensee immediately restricted the workers access to the Radiologically Controlled Area (RCA) and put out a site wide information notice to increase worker awareness of HRA entry requirements. This condition has been placed into the licensees corrective action program as action request (AR) 01951254. The finding was determined to be more than minor because it was related to the Occupational Radiation Safety cornerstone attribute of Human Performance (radiation worker proficiency) and adversely affected the cornerstone attribute to ensure the adequate protection of worker health and safety. Specifically, because the worker failed to comply with TS requirements for entry into a HRA he was not knowledgeable of area radiological conditions. The finding was evaluated in accordance with IMC 0609, Appendix C, where it was determined to be Green because it did not involve ALARA planning or work controls, was not an overexposure, did not contain a substantial potential for an overexposure, and the ability to assess dose was not compromised. This finding involved the cross-cutting aspect of Human Performance, Avoid Complacency (H.12) because the worker failed to apply the human performance tools of self and peer checks prior to entering into an HRA. |
| 05000251/FIN-2014005-02 | 2014Q4 | Turkey Point | Inadequate Procedure to Realign Steam Supply to the Gland Sealing Steam System | A self-revealing non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures, was identified for the licensees failure to maintain an adequate procedure for gland sealing steam supply realignment. Specifically, the licensee failed to have initial conditions in place in the procedure that provided specific direction that steam supply to the gland sealing system cannot be transferred from the main steam system to the auxiliary steam system with a unit in Mode 1 or 2. The licensee took corrective action to add initial conditions to procedure 4-GOP-103 that would not allow gland seal system steam supply realignment while in Modes 1 or 2. The licensee entered this issue into their corrective action program (CAP) as action request (AR) 1967899. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to have specific guidance in procedure 4-GOP-103 that prevented realigning the gland sealing steam supply while in Mode 1 or 2 resulted in lowering condenser vacuum and a subsequent reactor trip on low condenser vacuum when the gland sealing steam supply was being realigned with Unit 4 in Mode 1. The inspectors screened the finding using Attachment 4 to NRC Inspection Manual Chapter (IMC) 0609 and determined that the finding was a transient initiator contributor which required evaluation using Exhibit 1, Initiating Events Screening Questions, of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power (July 19, 2012). The inspectors determined that the finding was of very low safety significance (Green) because the performance deficiency did not result in a reactor trip and loss of mitigating equipment relied upon to transition the plant to a safe shutdown condition. The finding was associated with a cross-cutting aspect in the resources component of the human performance area because the licensee failed to ensure an adequate general operating procedure was available to support nuclear safety (H.1). |
| 05000250/FIN-2014005-03 | 2014Q4 | Turkey Point | Failure to Perform an Adequate Design Verification | A self-revealing finding was identified for the licensees failure to ensure an adequate design change was implemented during Unit 3 and Unit 4 instrument air compressor system upgrade modifications completed in 2013. Specifically, plant modifications EC 246991 and EC 246990 were accepted and placed in service by the licensee without verifying the control logic configuration would function properly and load under all conditions. As a result, the diesel-driven compressors would not load and pressurize the instrument air header in the event of a loss of instrument air pressure while in the standby mode of operation. Corrective actions included an immediate modification to the standby compressor loading control circuit to ensure the machine loaded automatically and revising general procedural guidance for compressor operation. The licensee entered this performance deficiency in their corrective action program as AR 01983607. The performance deficiency was more than minor because it was associated with the design control attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to have an adequate design for controlling the operation of the standby instrument air compressor resulted in a reactor trip due to the loss of instrument air pressure. The inspectors screened the issue under the initiating events cornerstone using Attachment 4 (June 19, 2012) and Exhibit 1 (June 19, 2012) of Appendix A to Inspection Manual Chapter (IMC) 0609, Significance Determination Process (June 2, 2011). The inspectors concluded that a detailed risk evaluation would be required because the finding was associated with the loss of a support system that resulted in a reactor trip and affected equipment that could be used by plant operators to mitigate the resulting plant transient. A senior reactor analyst (SRA) performed a detailed risk evaluation of this issue. The NRC model for Turkey Point was adjusted by: 1) increasing the initiating event frequency for a loss of instrument air (LOIA) event by one order-of-magnitude, and 2) the failure-to-run probability of the backup air compressors was set equal to 1.0. The change in core damage frequency results were below the 1E-6 threshold and the issue was determined to be of very low risk significance (Green). The finding was associated with a cross-cutting aspect in the resources component of the human performance area because the licensee failed to ensure instrument air system equipment was available and adequate to support nuclear safety (H.1). |
| 05000250/FIN-2014004-02 | 2014Q3 | Turkey Point | Notice of Enforcement Discretion (NOED) due to Exceeding Ultimate Heat Sink Temperature | The inspectors identified an unresolved item (URI) regarding Turkey Point NOED 14-2-001 granted on July 20, 2014. The inspectors reviewed Turkey Point NOED 14-2-001 and related documents to determine the accuracy and consistency with the licensees assertions and implementation of the licensees compensatory measures and commitments during the period of enforcement discretion, those of which included, in part, keeping a third CCW heat exchanger in service, increased frequency of CCW heat exchanger performance tests and cleaning, increased heat sink and system temperature monitoring and management oversight, just-in-time operator training, and minimizing the performance of risk-significant maintenance activities. Additional NRC inspection is required to conduct a review of the LER, root cause, and planned corrective actions. This URI is identified as URI 05000250, 251/2014004-02, Turkey Point Notice of Enforcement Discretion (NOED) 14-2-001 due to Exceeding Ultimate Heat Sink Temperature. |
| 05000250/FIN-2014004-01 | 2014Q3 | Turkey Point | Failure to Identify and Correct Unsealed Condulet to Prevent Water Intrusion | A self-revealing, non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to implement corrective actions to prevent water intrusion into electrical conduits that affected safety related equipment. Specifically, the licensee failed to establish corrective actions to prevent water intrusion into the power supply for the Unit 3 B train (3B) pressurizer back-up heaters. After discovery of the condition, the licensee completed immediate corrective actions to apply waterproofing sealant to an unsealed condulet elbow that was the source of the pressurizer back-up heater water intrusion. The licensee entered this issue into their corrective action program as ARs 1985831 and 1986395. This finding was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to implement corrective actions to prevent water intrusion events which resulted in the inoperability of 3B pressurizer back-up heaters. The inspectors evaluated the significance of the finding under the mitigating systems cornerstone using Table 2 of Attachment 4 (dated June 19, 2012) and Exhibits 2 and 4 of Appendix A (dated June 19, 2012) to Inspection Manual Chapter 0609, Significance Determination Process, (dated June 2, 2011). The inspectors determined the finding was of very low safety significance (i.e., Green) because the exhibit criteria did not screen to a detailed risk assessment. A cross-cutting aspect was not identified because this performance deficiency occurred in 2007 and there have been no recent opportunities for the licensee to apply current processes and procedures for this issue. Therefore, the inspectors concluded that the performance deficiency was not indicative of current licensee performance. |
| 05000250/FIN-2014004-03 | 2014Q3 | Turkey Point | Licensee-Identified Violation | Turkey Point Nuclear Generating Unit 3 and Unit 4 Technical Specification (TS) 3.3.1 required, in part, that the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be operable. Contrary to the above, for approximately 50 days from April 2013 until June 2013, the steam versus feed water flow mismatch reactor trip functions associated with Unit 4 feed water flow instruments F-4-487, F-4-496, and F-4-497 were inoperable because they exceeded their TS allowed actuation set points specified by TS Table 2.2-1, and the affected channels were not placed in trip within six hours or the unit placed in cold shutdown as required by TS. Additionally, for approximately 162 days from August 2012 until February 2013, the steam-feed water flow mismatch reactor trip function associated with Unit 3 feed water flow instrument F-3-476 was inoperable because it exceeded its TS allowed actuation set point and the affected channel was not placed in trip within six hours and the unit placed in cold shutdown as required by TS. The inspectors assessed the significance of the violation using Inspection Manual Chapter 0609 Attachment 4, Appendix A and Exhibit 2 (June 19, 2012). The inspectors noted that the diverse low-low steam generator level reactor trip safety function was not affected by the inoperable feed water flow instruments and the violation did not represent a complete loss of the anticipatory steam versus feed water flow mismatch reactor trip function. Therefore, the inspectors concluded that violation was of very low safety significance (i.e., Green) because the violation was not associated with a significant functional degradation of the reactor protection system. The licensee completed immediate corrective actions following discovery of the condition to adjust the affected instruments to within TS allowed values and entered the issue into the corrective action program as action request (AR) 1961512. |
| 05000251/FIN-2014003-01 | 2014Q2 | Turkey Point | Failure to Implement a Surveillance Procedure to Perform a RCS Unidentified Leak Rate Statistical Calculation | The NRC identified a non-cited violation (NCV) of Technical Specification 6.8.1, Procedures, for the licensees failure to implement procedure 4-OSP-041.1, Reactor Coolant System (RCS) Leak Rate Calculation. Specifically, the licensee did not perform a Unit 4 reactor coolant system leak rate statistical calculation to determine the change in the average unidentified RCS leak rate which resulted in not performing a Level 3 RCS leak rate investigation. Corrective actions included performing the calculation, performing a detailed leak investigation, and entering the performance deficiency in their corrective action program as action request 01962745. The performance deficiency was determined to be more than minor because it was associated with the initiating events cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inspectors determined that the licensees failure to fully implement the procedure directly resulted in not performing an RCS Level 3 leak rate investigation. The finding was screened using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, dated July 1, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1 for Initiating Events , dated July 1, 2012. The inspectors determined the finding was of very low safety significance because after a reasonable assessment of the degradation, the inspectors determined the finding would not have likely affected other systems used to mitigate a Loss of Coolant Accident (LOCA) resulting in total loss of their function. This finding was associated with a cross-cutting aspect in the procedure adherence component in the human performance area because the licensee failed to fully implement the RCS leak rate calculation procedure (H.8). |
| 05000250/FIN-2014003-02 | 2014Q2 | Turkey Point | Reactor Coolant System Pressure Boundary Leakage at Pressurizer Heater Sleeve Attachment Weld | Technical Specification (TS) 3.4.6.2 limiting condition for operation (LCO) required that primary coolant operational leakage shall be limited to No Pressure Boundary Leakage when in Modes 1 through 4. The action statement of TS 3.4.6.2 required that the plant be placed in hot standby (Mode 3) within 6 hours and in cold shutdown (Mode 5) within the following 30 hours. Although the beginning time of the pressure boundary leakage from the number 11 pressurizer heater sleeve could not be precisely determined, the inspectors concluded that the leakage had reasonably existed during the previous Unit 3 operating cycle for greater than the six hour TS action statement time limit to place Unit 3 in Mode 3. Therefore, contrary to the above, during the previous operating cycle which ended on March 17, 2014, Unit 3 operated in Mode 1 and was not placed in Mode 3 within six hours with primary coolant pressure boundary leakage from the number 11 pressurizer heater sleeve. The inspectors concluded that the violation would normally be characterized as Severity Level IV based on its very low safety significance. There had not been any perceptible changes in containment parameters (i.e., radiation levels, humidity or floor drain sump levels) to indicate that a leak existed. The inspectors reviewed the root cause analysis of the event and concluded that the pressure boundary leakage could not have been avoided or otherwise detected by the licensees quality assurance program or other related control measures prior to the licensees discovery of the condition on March 17, 2014. As discussed in Section 2.2.4.d of the Enforcement Policy, a violation involving no performance deficiency is considered an exception to using only the operating reactor assessment program. Therefore, in consultation with the Office of Enforcement, the NRC has concluded that the exercise of enforcement discretion is warranted in accordance with Section 3.5 of the Enforcement Policy, because the violation resulted from matters not within the licensees control. Accordingly, this violation will not be documented or considered in the NRCs assessment process, but has been assigned an Enforcement Action number, EA-14- 124, to document the granting of enforcement discretion. This issue is documented in the licensees CAP as AR 1949021. Corrective action involved a half-nozzle repair of heater sleeve 11 which relocated the reactor coolant system pressure boundary to the outside of the pressurizer at the heater sleeve penetration, a visual inspection of the remaining 77 heaters in the Unit 3 pressurizer to ensure no other heater sleeve penetration leaks, and a visual inspection of the half-nozzle repair with Unit 3 at normal operating pressure and normal operating temperature. The LER is closed. |
| 05000250/FIN-2014002-04 | 2014Q1 | Turkey Point | Loose Breaker Control Power Fuse Holder Caused 3B ICW Pump to be Inoperable Longer than Allowed Outage Time | On September 28, 2013, while Unit 3 was in Mode 1, operators discovered that power to the 3B intake cooling water (ICW) pump breaker closing circuit and charging springs was lost. The under current (UC) fuse holder was noted to be slightly backed out and not firmly in place which resulted in the loss of breaker control power. Control power was restored after the fuse holder was pressed back in place on September 29, 2013. The licensee performed an investigation that determined that the 3B ICW pump had been inoperable for approximately four days (the time the pump was last started until the fuse holder was fully re-inserted), which was longer than the allowed TS 3.7.3, Intake Cooling Water System, outage time of 72 hours. Although operators performed a daily verification that the breaker control power available white indicating light was lit on the breaker cubicle, the licensee determined that reasonable assurance could not be established that the fuses had enough contact with the base to ensure power to the closing spring even though the white control power light was lit. Therefore, the licensee concluded that the 3B ICW pump was inoperable for four days prior to discovery of the fuse holder condition. The inspectors determined the ICW system would have been able to perform its function even with the 3B ICW pump inoperable. The 3A and 3C ICW pumps were available and only one ICW pump is needed to remove design basis heat loads. Based on the availability of the ICW system to perform its heat removal function; and the relatively short duration of the condition prior to its discovery on September 28, the inspectors concluded that the event was of very low safety significance. The inspectors determined that a violation of TS limiting condition for operation (LCO) 3.7.3, Intake Cooling Water System, occurred since Unit 3 was in Mode 1 and the 3B ICW pump was not returned to an operable status within 72 hours or the unit shut down and placed in hot standby within the next six hours. Although a violation of the TS LCO occurred, the violation was not attributable to an equipment failure that was avoidable by reasonable licensee quality assurance measures or management controls. The inspectors concluded that the violation would normally be characterized as Severity Level IV based on its very low safety significance. The NRC exercised enforcement discretion (Enforcement Action (EA)-14-058) in accordance with Section 2.2.4.d of the Enforcement Policy because the violation was not associated with a licensee performance deficiency; and therefore, it will not be considered in the assessment process or the NRCs Action Matrix. This issue is documented in the licensees CAP as AR 1929130. Corrective actions included an investigation of the material condition of the fuse holder and base assembly, and a revision to the breaker operation procedure to include additional guidance on validating proper installation of the fuse holder when racking in a four kilovolt breaker. The LER is closed. |
| 05000250/FIN-2014002-03 | 2014Q1 | Turkey Point | Loose Breaker Control Power Fuse Holder Caused 3B ICW Pump to be Inoperable Longer than Allowed Outage Time | On September 28, 2013, while Unit 3 was in Mode 1, operators discovered that power to the 3B intake cooling water (ICW) pump breaker closing circuit and charging springs was lost. The under current (UC) fuse holder was noted to be slightly backed out and not firmly in place which resulted in the loss of breaker control power. Control power was restored after the fuse holder was pressed back in place on September 29, 2013. The licensee performed an investigation that determined that the 3B ICW pump had been inoperable for approximately four days (the time the pump was last started until the fuse holder was fully re-inserted), which was longer than the allowed TS 3.7.3, Intake Cooling Water System, outage time of 72 hours. Although operators performed a daily verification that the breaker control power available white indicating light was lit on the breaker cubicle, the licensee determined that reasonable assurance could not be established that the fuses had enough contact with the base to ensure power to the closing spring even though the white control power light was lit. Therefore, the licensee concluded that the 3B ICW pump was inoperable for four days prior to discovery of the fuse holder condition. The inspectors determined the ICW system would have been able to perform its function even with the 3B ICW pump inoperable. The 3A and 3C ICW pumps were available and only one ICW pump is needed to remove design basis heat loads. Based on the availability of the ICW system to perform its heat removal function; and the relatively short duration of the condition prior to its discovery on September 28, the inspectors concluded that the event was of very low safety significance. The inspectors determined that a violation of TS limiting condition for operation (LCO) 3.7.3, Intake Cooling Water System, occurred since Unit 3 was in Mode 1 and the 3B ICW pump was not returned to an operable status within 72 hours or the unit shut down and placed in hot standby within the next six hours. Although a violation of the TS LCO occurred, the violation was not attributable to an equipment failure that was avoidable by reasonable licensee quality assurance measures or management controls. The inspectors concluded that the violation would normally be characterized as Severity Level IV based on its very low safety significance. The NRC exercised enforcement discretion (Enforcement Action (EA)-14-058) in accordance with Section 2.2.4.d of the Enforcement Policy because the violation was not associated with a licensee performance deficiency; and therefore, it will not be considered in the assessment process or the NRCs Action Matrix. This issue is documented in the licensees CAP as AR 1929130. Corrective actions included an investigation of the material condition of the fuse holder and base assembly, and a revision to the breaker operation procedure to include additional guidance on validating proper installation of the fuse holder when racking in a four kilovolt breaker. The LER is closed. |
| 05000250/FIN-2014008-01 | 2014Q1 | Turkey Point | Failure to Properly Program the Turbine Generator Digital Control System Load Drop Anticipatory Circuit Results in a Manual Reactor Trip | A self-revealing finding was identified for the failure to establish new digital software set points for the load drop anticipatory (LDA) logic circuit associated with an extended power uprate (EPU) digital turbine electro-hydraulic control (EHC) system design modification. Specifically, the software for the LDA logic circuit was programmed to reset at a value that would not be reached during a normal reactor plant shutdown before the turbine control system sensed a loss of load condition and closed the turbine control valves. As a result, during a planned Unit 3 reactor plant shutdown, the LDA control logic unexpectedly closed the turbine control valves at 25 percent reactor power. The operators then manually tripped the unit based on the indication of loss of turbine load in the control room. Licensee Unit 3 software engineering change (EC) package 246849 change request notice (CRN) 253 Attachment 5, Turbine Control Initial Values, instructed the programmer to set the LDA disarm value to 50 percent turbine load. Contrary to this instruction, the programmer set the disarm value to 50 pounds per square inch gauge (psig) steam pressure. The failure of the programmer to establish the proper set point value in the LDA reset logic was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment reliability attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The finding was determined to be of very low safety significance (Green) based on Exhibit 1, Initiating Events Screening Questions, found in Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process for Findings At-Power (dated June 19, 2012). This was due to the fact that the finding did not result in a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined the cause of this finding was associated with a cross cutting aspect of procedure adherence. Specifically, the licensee set the turbine control valve LDA reset point to 50 psig instead of 50 percent turbine load as prescribed in EC 246849. |
| 05000250/FIN-2014002-02 | 2014Q1 | Turkey Point | TS Channel Calibration of ESF Steam Line Protection Channel III Not Performed | A self-revealing non-cited violation (NCV) of TS Section 3.3.2, Engineered Safety Features Actuation Instrumentation, (ESF) was identified when the licensee failed to perform the channel calibration of Unit 3 ESF steam pressure protection channel III within the required 18-month frequency which resulted in operation with steam generator pressure transmitter PT-3-495 inoperable for approximately 10 months. This issue was placed in the licensees CAP as AR 1938191. Corrective actions included replacing PT-3-495, performing an extent of condition on all other work orders completed during the extended power uprate (EPU) outage to ensure TS compliance, and revising the surveillance tracking program procedure to require verification that the required surveillance testing is completed prior to crediting non-dedicated work orders. The performance deficiency was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform the channel calibration surveillance test procedure for transmitter PT-3-495 within the 18- month required frequency resulted in 10 months of channel inoperability. The finding was screened using Exhibit 1, Mitigating Systems Screening Questions, found in Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process for Findings At-Power (June 19, 2012). The inspectors determined the finding was of very low safety significance (Green) because the finding did not affect design or qualification, did not represent a loss of system function, and did not represent an actual loss of function of a technical specification train of equipment. The finding was associated with a cross-cutting aspect in the work management component of the human performance area because the licensee failed to implement their process for planning, controlling, and executing required surveillance tests (H.5). |
| 05000250/FIN-2014002-01 | 2014Q1 | Turkey Point | Failure to Take Adequate Corrective Actions to Correct Flow Induced Vibration Leads to CCW Piping Weld Failures | A self-revealing non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified when the licensee failed to implement corrective actions that addressed the low stress high cycle fatigue of component cooling water (CCW) relief valve (RV) 4-747B piping caused by flow induced vibration. As a result, CCW system flow induced vibration resulted in weld cracks and system pressure boundary leakage in January 2014. This issue was placed in the licensees corrective action program (CAP) as action request (AR) 1931761. Corrective actions included performing a root cause evaluation, implementing special instructions to minimize the time that split header operation is performed, and developing a plan to replace the existing relief valve with an orifice or alternate relief valve. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to implement adequate corrective actions to address CCW system flow induced vibration resulted in weld cracks and CCW system pressure boundary leakage in January 2014. The finding was screened using Exhibit 1, Mitigating Systems Screening Questions, found in Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process (SDP) for Findings At-Power (June 19, 2012). The inspectors determined the finding was of very low safety significance (Green) because the finding did not affect design or qualification, did not represent a loss of system function, and did not represent an actual loss of function of a TS train of equipment. The finding was associated with a cross-cutting aspect in the evaluation component of the problem identification and resolution area because the licensee did not thoroughly evaluate issues and corrective actions from previous weld failures on CCW system RV-4-747B piping caused by flow induced vibration (P.2). |
| 05000251/FIN-2013005-01 | 2013Q4 | Turkey Point | Inadequate Test Precautions, Limitations, and Instructions for Performing Harmonic Testing on the Unit 4 Turbine Generator | A self-revealing finding was identified for the licensees failure to provide adequate test precautions, limitations, and instructions for performing harmonic testing on the Unit 4 turbine generator control circuitry while in Mode 1 operation. As a result, 480 volt load center voltage was lowered enough to initiate a degraded voltage signal to the 4B safety related 4 kV bus sequencer which tripped reactor coolant pumps causing a reactor trip due to low reactor coolant system flow. This issue was placed in the licensees corrective action program as action request (AR) 1867690. Corrective actions included performing a root cause evaluation and a revision to procedure WM-AA-100-1000, Work Activity Risk Management, to include additional guidance involving online maintenance and risk insights when planning maintenance on the main generator. The licensees failure to provide adequate test precautions, limitations, and instructions for performing harmonic testing on the Unit 4 turbine generator control circuit was a performance deficiency. Specifically, TI-246904-01, 3rd Harmonic Relay Test, did not provide adequate instructions to prevent creating a degraded voltage condition and the test was classified in error as low risk rather than high risk per licensee procedure WM-AA-100-1000, Work Activity Risk Management. The inspectors determined the performance deficiency was more than minor using IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because the performance deficiency was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to have an adequate procedure for controlling the turbine generator harmonic testing resulted in a reactor trip due to the loss of reactor coolant pumps from 4B sequencer 4 kV bus stripping. The inspectors evaluated the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Transient Initiators. The inspectors determined the finding was of very low safety significance (green) because the finding did not result in a reactor trip and a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. The finding was associated with a cross-cutting aspect in the work control component of the human performance area because the licensee failed to include the proper risk insights for work activities related to nuclear safety and prevent a subsequent reactor trip (H.3(a)). |
| 05000250/FIN-2013004-01 | 2013Q3 | Turkey Point | Inadequate Procedure to Vent 3B SGFP Results in AFW Actuation | A self-revealing non-cited violation of Technical Specification 6.8.1, Procedures, was identified for the licensees failure to maintain an adequate procedure for venting the 3B steam generator feed pump (SGFP). Specifically, the licensee had failed to remove temporary instructions in Section 5.4 of procedure 3-NOP-074, Steam Generator Feedwater System, to jumper the contacts on the 3B SGFP breaker such that the breaker appeared open to the auxiliary feedwater (AFW) actuation logic, and as a result, AFW was inadvertently actuated and had to be secured by operators during a start of the 3B SGFP from the control room. The licensee entered the issue into the corrective action program as action request 1855704 and took corrective actions to revise 3-NOP-074 by removing the jumper installation steps from the procedure. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to remove the procedural instructions for installing a jumper in the 3B SGFP control circuit resulted in an inadvertent AFW actuation and required operators to take action to temporarily secure the ability of AFW to feed the steam generators. The inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a reactor trip and a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. The finding was associated with a cross-cutting aspect in the resources component of the human performance area because the licensee failed to ensure an accurate and up-to-date procedure was maintained for operation of the feedwater system (H.2(c)). |
| 05000250/FIN-2013004-02 | 2013Q3 | Turkey Point | Failure to Follow Procedure to Switch Running SGFPs Results in AFW Actuation | A self-revealing non-cited violation of Technical Specification 6.8.1, Procedures, was identified for the licensees failure to implement Section 2.0 of procedure 3-NOP-074, Steam Generator Feedwater System, for starting the 3A steam generator feedwater pump (SGFP). Specifically, the licensee failed to implement 3-NOP-074 and ensure that a second condensate pump (CP) was running before starting a second SGFP which resulted in a loss of normal feedwater to the steam generators and an actuation of auxiliary feedwater (AFW). Operators took action to secure AFW flow to the steam generators to limit plant cool down and opened the reactor trip breakers to obtain additional reactivity shut down margin. Operators also took action to start the A standby steam generator feed pump (SBSGFP) to maintain level in the SGs and both trains of AFW were returned to operable standby status. The licensee entered the issue into their corrective program as action request 1856476. The inspectors determined that the performance deficiency was more than minor because it was associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to ensure that a second CP was running prior to starting 3A SGFP resulted in the trip of the running SGFP 3B and AFW actuation in response to the loss of normal feedwater supply. The inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a reactor trip and a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. The finding was associated with a cross-cutting aspect in the work practices component of the human performance area because the licensee failed to ensure proper supervisory oversight of work activities related to nuclear safety and prevent the loss of running SGFPs (H.4(c)). |
| 05000250/FIN-2013004-03 | 2013Q3 | Turkey Point | Failure to Provide Adequate Instructions during Maintenance on the Gland Seal Steam System | A self-revealing finding was identified due to the licensees failure to provide adequate work instructions for throttling the Unit 3 gland seal steam bypass valve. As a result of the licensees inadequate work instructions, an operator opened the spill bypass valve on the gland seal steam system until system steam pressure dropped and allowed air in-leakage through the turbine gland seals. This resulted in a reactor trip and the main condenser was unavailable for reactor decay heat removal until vacuum could be restored. The licensee entered this issue into their corrective action program as action request 1847369 and revised the system operating procedure to address operation of the bypass line around the spillover control valve. The inspectors determined the performance deficiency was more than minor because it was associated with the configuration control attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to provide adequate work instructions for the operation of the gland seal steam spillover bypass valve resulted in a reactor trip with the main condenser unavailable for reactor decay heat removal until vacuum could be restored. The inspectors screened the finding and determined that the finding was a transient initiator contributor which required a detailed risk analysis because the finding resulted in a reactor trip with a loss of condenser vacuum. A bounding analysis was performed by a regional Senior Reactor Analyst who concluded that the finding resulted in an increase in core damage frequency of less than 1E-6/year and, therefore, was a Green finding of very low safety significance. The finding was associated with a cross-cutting aspect in the work control component of the human performance area because the licensee did not adequately incorporate the need for planned contingencies, compensatory actions or abort criteria to ensure that throttling the gland seal steam spillover bypass valve would not result in a reactor trip and loss of the main condenser (H.3(a)). |