Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000390/FIN-2017002-052017Q2Watts BarLicensee-Identified ViolationWatts Bar Nuclear Plant TS 5.7.1.1 states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures in Regulatory Guide (RG) 1.33, Revision. 2, Appendix A, February 1978. Procedures for surveillance tests are applicable procedures under RG 1.33 Appendix A, 8.b. Contrary to this requirement, on April 4, 2017, surveillance procedure 0-SI-82-4, 18 Month Loss of Offsite Power with Safety Injection Test DG 1B-B, Revision 63, was not implemented as written. Specifically, Step 3.1 (3) was not followed when the 1B-B safety injection pump discharge isolation valve was closed but not tagged as directed by the procedure. As a result of not being tagged, there was no programmatic control in place to return the valve to the open position upon completion of 0-SI-82-4. Therefore, the valve was left in the closed position, causing the B train of safety injection to be inoperable from April 11, 2017, until May 10, 2017, when the valve was discovered to be closed during operator rounds. Because the 1B safety injection pump was inoperable for longer than its TS allowed outage time of 72 hours, a regional senior reactor analyst conducted a detailed risk evaluation using SAPHIRE (Version 8.1.5) and the standard model for Watts Bar (SPAR Version 8.50). The resulting change in core damage frequency was less than 1E-6; therefore, the finding was determined to be of very low safety significance (Green). The licensee entered this issue into their corrective action program as CR 1294133.
05000390/FIN-2017002-032017Q2Watts BarFailure to Follow Procedure Results in Reactor Coolant Pump Failure to Transfer and Unit 1 Reactor TripGreen. A self-revealed Green finding was identified for the failure to follow procedure NPG-SPP-22.207, Procedure Use and Adherence Revision 4, which requires that applicable procedures are used for all activities controlled by a written procedure. The licensee entered this into their corrective action program as CR 1291140 The failure to follow procedure NPG-SPP-22.207, Procedure Use and Adherence, Revision 4, was a performance deficiency. The performance deficiency was more than minor because it affected the Initiating Events Cornerstone attribute of Human Performance and adversely affected the cornerstone objective in that it resulted in two reactor trips. The inspectors determined that the finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. The finding was not assigned a cross-cutting aspect since none of the CCAs described in IMC 0310 corresponded to an apparent cause or most significant causal factor of the performance deficiency. (Section 4OA3.6)
05000390/FIN-2017002-022017Q2Watts BarFailure to Implement Clearance on Containment Isolation Valve Results in TS 3.6.3 ViolationGreen. A self-revealed non-cited violation of Technical Specification (TS) 3.6.3, Containment isolation Valves, was identified for a failure to properly implement a clearance for containment isolation valve surveillance testing. Clearance 1-30-1011-WW removed fuses from a different valve than the one specified in the clearance. The licensee entered this issue into their corrective action program as CR 1245529. The failure to comply with NPG-SPP-10.2, Steps 3.1.2.B.5 and 6, was a performance deficiency. The performance deficiency was more than minor because it adversely affected the configuration control attribute of the Barrier Integrity Cornerstone because the incorrectly placed clearance resulted in the inoperability of the containment isolation valve for longer than its TS allowed outage time, reducing ensurance that the containment function assumed in the safety analyses would be maintained. The inspectors determined that this violation was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters. The finding has a cross-cutting aspect in the Avoid Complacency component of the Hum an Performance area as defined in NRC IMC 0310, because multiple personnel failed to recognize and plan for the possibility of 3 mistakes and error reduction tools, such as concurrent verification, were not appropriately implemented (H.12).
05000390/FIN-2017002-062017Q2Watts BarLicensee-Identified ViolationTitle 10 CFR 50.72(b)(3)(v)(C) requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems or components that are needed to control the release of radioactive material. Contrary to the above, on March 9, 2017, the licensee failed to notify the NRC that reactor containment was inoperable, resulting in a condition that could have prevented fulfillment of a safety function. Specifically, an inner containment door equalizing valve was not fully shut when the outer containment door was open for entry into upper containment, thereby resulting in a direct path from containment to the auxiliary building. This failure to report was assessed using Section 2.2.4 of the NRCs Enforcement Policy using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72 or 50.73, and the issue was determined to be a SL IV violation. The licensee entered this issue into their corrective action program as CR 1273873.
05000390/FIN-2017002-042017Q2Watts BarFailure to Report Multiple Examples of a Loss of Safety Function in accordance with 10 CFR 50.72 and 50.73Severity Level IV. The inspectors identified a Severity Level IV non-cited violation of 10 Code of Federal Regulations (CFR) 50.72 and 50.73, with multiple examples due to the licensees failure to make the required eight-hour non-emergency notification and submit a Licensee Event Report (LER) to the NRC within 60 days for conditions that, at the time of discovery, could have prevented fulfillment of a safety function. These issues have been entered into the licensees corrective action program as condition report (CR) 1310096. The inspectors determined that the licensees failure to comply with 10 CFR 50.72(b)(3)(v) and 50.72(a)(2)(v) was a performance deficiency. This performance deficiency was dispositioned under traditional enforcement because the failure to make a non-emergency notification and submit an LER within the time requirements may impact the ability of the NRC to perform its regulatory oversight function. The violation was assessed using Sections 2.2.4 and 6.9.d.9 of the NRCs Enforcement Policy and determined to be a SL IV violation. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000391/FIN-2017002-012017Q2Watts BarInadequate Chemistry Procedure Results in Inoperable Containment Isolation ValvesSL IV. A self-revealed severity level (SL) IV non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when implementing an inadequate procedure resulted in rendering the steam generator chemistry sample containment isolation valves inoperable. The licensee entered this issue into their corrective action program as CR 1160910. The inspectors determined that the use of an inadequate procedure that rendered the containment isolation valves inoperable was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC-2517, Appendix C, because the use of an inadequate procedure rendered the containment isolation valves inoperable. The inspectors determined this finding to be of very low safety significance because it did not represent a breakdown of the licensees quality assurance program. This finding had a cross-cutting aspect in the work management component of the Human Performance cross-cutting area because the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities (H.5).
05000390/FIN-2016004-022016Q4Watts BarInadequate Immediate Determination of Operability for Containment Penetration X-65Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to address all the design criteria for check valve, 1-CHV-31-3407, in the basis of the immediate determination of operability (IDO) for containment penetration X-65 to conclude that a reasonable expectation of operability existed. On September 19, Technical Specification (TS) compliance was restored when Penetration X-65 returned to operable when it was isolated and drained. The violation was entered into the licensees corrective action program as condition report (CR) 1216892. The performance deficiency was more than minor because it adversely affected the design control attribute of the barrier integrity system cornerstone. Specifically, reasonable assurance of operability did not exist for containment penetration X-65 from September 18, 2016, until September 19, 2016. The inspectors performed an initial screening of the finding and determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components; and hydrogen igniters are not applicable. The cause of this finding had a cross-cutting aspect of Evaluation in the area of Problem Identification and Resolution, because the licensee did not consider all functions of check valve 1-CKV-31-3407 when performing the IDO after the valve failed to pass the surveillance instruction. (P.2).
05000390/FIN-2016501-012016Q4Watts BarFailure to Maintain Minimum On-Shift Emergency Response Staffing LevelsGreen: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.47(b)(2) for the licensees failure to maintain the effectiveness of its emergency plan, when on more than one occasion, the number of control room operators fell below minimum staffing, as required by Appendix C of NP-REP Tennessee Valley Authority (TVA) Nuclear Power Radiological Emergency Plan (E-Plan). The licensees corrective actions included entering the issue into their corrective action program as CR 1233650. The performance deficiency was more than minor because it was associated with the emergency response organization readiness attribute of the Emergency Preparedness cornerstone and adversely impacted the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the finding in accordance with Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and using Table 5.2-1 Significance Examples for 50.47(b)(2), determined that this finding represented an example of a staffing process that would permit a shift to go below E-Plan minimum staffing requirements. The inspectors determined that the licensees process, on more than one occasion, failed to ensure that on-shift staffing met E-Plan minimum staffing requirements between March 20 and May 6, 2016. The cause of the finding was determined to be associated with the cross-cutting aspect of thorough evaluation of problems in the corrective action component of the Problem Identification and Resolution area because the organization failed to periodically analyze information from the corrective action program and other assessments in the aggregate to identify programmatic and common cause issues (P.4).
05000390/FIN-2016004-032016Q4Watts BarNotice of Enforcement Discretion 16-2-01 for Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications(Opened) Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications and Notice of Enforcement Discretion 16-2-01 Introduction: The inspectors opened an unresolved item associated with a potential noncompliance with TS 3.8.1 that occurred on October 15, 2016. Notice of Enforcement Discretion 16-2-01 was granted by the NRC staff agreeing not to enforce compliance with the TS completion time for an additional 130 hours. Description: At 6:32 a.m. on October 12, 2016, Watts Bar operations staff declared the 1A-A EDG inoperable when the output breaker to the 1A shutdown board opened unexpectedly due to phase overcurrent during performance of the load test required by procedure 0-SI-82-13, 24 Hour Load Run - DG 1A-A. The 1A-A emergency diesel generator was operating normally prior to the opening of the breaker. The licensees initial assessment determined the likely cause of the breaker trip was operation of the tap changer associated with the offsite power supply transformer. A subsequent 24 hour EDG load test was started at 12:35 a.m. on October 13, 2016. At 6:45 p.m. on October 13, 2016, operations staff noted mega volt amps (reactive) swings. During subsequent troubleshooting activities, it was determined that the mega volt amps (reactive) variance could be consistently reproduced by slight movement of a potentiometer on the 1A-A EDG voltage regulator. The licensee determined that an issue in the voltage regulator circuit was the most likely cause of the output breaker trip, and made preparations to replace and calibrate the voltage regulator on which the potentiometer was located. The licensee determined that it would require more than 72 hours to complete the removal and replacement of the voltage regulator and post-maintenance testing. The licensee requested a notice of enforcement discretion and an additional 144 hours to restore EDG 1A-A. A notice of enforcement discretion for an additional 130 hours was granted by the NRC staff at 9:30 p.m. on October 14, 2016. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific TSs in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (URI 05000390, 391/2016004-03, Notice of Enforcement Discretion 16-2-01 for Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications) This activity constitutes completion of one event follow-up sample, as defined in IP 71153
05000390/FIN-2016004-012016Q4Watts BarInadequate Immediate Determination of Operability for Essential Raw Cooling Water PumpsGreen: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to base an immediate determination of operability (IDO) for essential raw cooling water (ERCW) pumps on information sufficient to conclude that a reasonable expectation of operability existed. The licensee restored compliance on November 30, 2016, when they documented an IDO that met the requirements of OPDP-8. The violation was entered into the licensees CAP as CR 1237178. The performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone. Specifically, reasonable assurance of operability did not exist for the ERCW pumps from November 29, 2016 until November 30, 2016. The inspectors determined the finding was of very low safety significance (Green) because it did not represent an actual loss of function for at least a single train for longer than its technical specification allowed outage time. The cause of this finding had a cross cutting aspect of Teamwork in the Human Performance area, because individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries such that nuclear safety is the overriding priority. (H.4).
05000390/FIN-2016001-092016Q1Watts BarAppropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant ResponseThe inspectors identified an URI associated with the timely and effective corrective action associated with an adverse trend in safety related pump performance, including mechanical seal degradation and failure. This item is unresolved pending review and evaluation of the licensees response to the CRs generated to determine if a performance deficiency exists. During Unit 1, 2015 fall outage, the 1A Safety Injection (SI) pump mechanical seal was replaced. The mechanical seal had degraded to a point at which the leakage was greater than the Technical Specification limit for ECCS leakage outside of containment. The inspectors identified several issues during a review of the Prompt Determination of Operability for CR 1125623 and WO 116050574 to replace the seal. Specifically, inspectors found that non-QA1 parts were being used for seal replacement, the seal was the original equipment manufacturer part from startup, the failure mechanism was not clearly understood, and an extent of condition review was not performed. The inspectors reviewed other safety related pump mechanical seal performance and corrective action program entries. The inspectors are awaiting the completion of the licensees evaluation to determine the licensees compliance with applicable procedures and TS relative to pump operability and ECCS leakage limits outside containment. Additional inspection activities are needed to determine the extent of condition and compliance with the procedures and TS. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000390/2016001-09, Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response.
05000247/FIN-2014005-022014Q4Indian PointIncomplete and Inaccurate Medical Information Provided by the Licensee Which Impacted an Operators License RenewalEntergy identified two AVs of NRC requirements related to Entergy not notifying the NRC within 30 days of a change in a licensed reactor operators (ROs) medical condition and to providing information to the NRC pertaining to renewing a RO license that was not complete and accurate in all material respects. Specifically, Entergy identified an AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.74, Notification of Change in Operator or Senior Operator Status, for Entergys failure to notify the NRC within 30 days after learning, on October 25, 2012, that a Unit 3 RO had a permanent disability or illness (sleep apnea). Entergy also did not request an amended license with a condition to account for the medical issue, resulting in the RO performing licensed duties without a properly restricted license. Additionally, Entergy identified an AV of 10 CFR 50.9, Completeness and Accuracy of Information, pertaining to Entergys failure to provide information to the NRC in the ROs license renewal application in that it did not specify that the RO had a medical condition (sleep apnea) that required a restriction (for use of a continuous positive airway pressure (CPAP)). The NRC, in turn, issued a license renewal that did not contain the necessary restriction. Compliance was restored on July 7, 2014, when Entergy submitted a letter to the NRC with a Form 396 indicating the new restriction for the use of a CPAP machine. On August 14, 2014, the NRC issued a license amendment with the new restriction. These issues were entered into Entergys corrective action program (CAP) as condition report (CR)-IP3-2014-1416 and CR-IP2-2014- 4202. The inspectors determined that Entergys failure to report a change in a licensed operators permanent medical condition to the NRC and subsequently provide complete and accurate information to the NRC was a performance deficiency that was within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue impacted the NRCs ability to perform its regulatory function. The inspectors screened the issue using Section 6.4.c.4(b) of the NRC Enforcement Policy and preliminarily determined that these AVs meet the definition of a Severity Level III violation because Entergy failed to report a condition that would have required the addition of a license restriction within the required timeframe and, again, for the ROs license renewal. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and the final significance will be dispositioned in future correspondence. Because the significance determination of this issue is not complete, it is identified as TBD.
05000247/FIN-2014005-012014Q4Indian PointLicensed Operator Requalification Remedial Exam Standard AdherenceThe inspectors identified a Green finding (FIN) because Entergy did not adhere to their procedural standards for generating remedial written exams. Entergy failed to follow the guidance as stated in their procedure EN-TQ-201-03, Systematic Approach to Training, Section 5.4, regarding remedial exam construction when an operator was retested on April 25, 2013. The inspectors determined that Entergys failure to adhere to their remedial examination standards in EN-TQ-201-03 was a performance deficiency. The inspectors determined that the finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding affected the quality and level of difficulty of the remedial quiz which potentially impacted Entergys ability to appropriately evaluate the licensed operator. The inspectors determined that this issue had a cross-cutting aspect in Human Performance, Procedure Adherence, because Entergy did not follow their procedural standards for generating remedial written exams.
05000247/FIN-2014005-032014Q4Indian PointLicensee-Identified Violation10 CFR 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, requires measures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. Contrary to 10 CFR 50, Appendix B, Criterion XII, Indian Point did not properly implement their measuring and test equipment (M&TE) program resulting in the use of uncalibrated M&TE to perform maintenance, tests, and meet surveillance requirements on safety-related SSCs. Entergy identified deficiencies in their M&TE program during an Entergy Nuclear Oversight Quality Assurance audit of the Entergy Maintenance department. As a result of the Quality Assurance finding, a root cause analysis was conducted and corrective action plan developed. The corrective action plan CR-IP2-2014-03809 was reviewed by NRC inspectors as well as operability assessments conducted by Entergy operations personnel on safetyrelated SSCs worked on with out-of-tolerance M&TE. The issue screened to be of very low safety significance (Green) using IMC 0609, Appendix A, because the affected safety-related SSCs maintained their operability. No additional findings resulted from the NRC inspector review.
05000247/FIN-2014005-042014Q4Indian PointLicensee-Identified ViolationAccording to 10 CFR 55.21 and 33, licensed operators are required to have a physical examination every two years to ensure that their medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety. As a part of licensed operator medical evaluations, olfactory testing is required as specified in ANSI/ANS-3.4-1983, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants. Olfactory testing in the standard states Nose. Ability to detect odor of products of combustion and of tracer and marker gases. License procedure, EN-NS-112, Medical Program, has the same wording. Contrary to this requirement, in CR-IP2-2014-04622, Entergy identified that they had not been testing operators for two tracer/marker gases used on site wintergreen in the carbon dioxide systems and mercaptan used in natural gas. This violation is subject to traditional enforcement because of the potential impact upon the regulatory process because the operators medical conditions are reviewed by the NRC when issuing or renewing operator licenses. This issue meets the criteria for a Severity Level IV violation because, upon subsequent olfactory testing, all operators were found to meet the health requirements for licensing.
05000286/FIN-2014005-072014Q4Indian PointLicensee-Identified ViolationOn March 1, 2013, Entergy personnel tested Unit 3 main steam safety valves and determined main steam safety valve MS-46-3 had a lift setpoint outside of the +/-3 percent lift setting required by TS 3.7.1. Subsequently, MS-46-3 was declared inoperable and further testing found valve MS-48-3 also lifted out of the TS band. TS 3.7.1 requires the main steam safety valves be operable or reduce neutron flux trip setpoint to less than that listed in TS Table 3.7.1-1. Contrary to the above, as of March 1, 2013, main steam safety valves MS-46-3 and MS-48-3 had lift setpoints outside of the TS required band and flux trip setpoints were not reduced to those listed in TS Table 3.7.1-1. The affected valves were adjusted at the time of testing to within the required band, the condition was documented in the CAP as CR-IP3-2013- 0869 and CR-IP3-2013-0892, and an evaluation was initiated. Other valves similarly tested were satisfactory. No performance deficiency was identified because it was not reasonable for Entergy to foresee and prevent the change in main steam safety valve setpoint during plant operation. Corrective actions to prevent recurrence were documented in LER 05000286/2013-001-00. The violation was more than minor because it impacted the Equipment Performance attribute of the Mitigating Systems cornerstone. The issue screened to be of very low safety significance (Green) using IMC 0609, Appendix A because the overall pressure mitigating function was not affected by the degradation of the two valves of the twenty total.
05000247/FIN-2014005-062014Q4Indian PointLicensee-Identified ViolationOn February 24, 2014, Entergy personnel determined that a condition prohibited by Unit 2 TSs existed when a pinhole leak from a drain valve body was identified which resulted in an inoperable 23 SG. TS 3.4.4 requires during Mode 1 and 2 that four RCS loops be operable or be in Mode 3 within 6 hours. Contrary to the above, prior to February 24, Indian Point Unit 2 operated in Modes 1 and 2 with an inoperable SG when a pinhole leak from a valve body on the 23 SG existed in excess of 6 hours without entering Mode 3. Although attempts had been made to identify the source of a small secondary leak in containment during plant operation, the drain valve was not accessible with the reactor in operation and plant shutdown was required to complete the inspection on February 24, 2014. No performance deficiency was identified because it was not reasonable for Entergy to foresee and prevent the pinhole leak. The leak when found was documented in CR-IP2-2014-0975, and the valve was replaced. The violation was more than minor because it impacted the Equipment Performance attribute of the Initiating Events cornerstone. The issue screened to be of very low safety significance (Green) using IMC 0609, Appendix A when loss of coolant analysis assumptions and equipment performance were not affected by the degradation.
05000247/FIN-2014005-052014Q4Indian PointLicensee-Identified ViolationAccording to 10 CFR 55.25, if an operator develops a permanent physical or mental condition that causes the operator to fail to meet the requirements of 10 CFR 55.21, the facility licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c) which states that the regional administrator shall be notified if a licensed operator develops a permanent disability or illness. Contrary to these requirements, during the time frame of July through September 2014, the facility licensee identified four operators (in addition to the one mentioned above) that required medical restrictions and that the NRC needed to be notified. These four cases have been documented in CR-IP2-2014-04202, CR-IP3-2014-1961, and CRIP3- 2014-2156. In all four cases, the individual operators were untimely in notifying the facility licensee of the changes in their medical conditions or the licensee physician failed to recognize the need to report the condition to the NRC. This violation is subject to traditional enforcement because of the potential impact upon the regulatory process for issuing restrictions to operators licenses. This issue meets the criteria for a Severity Level IV violation because all of the operators met the criteria of ANSI/ANS-3.4-1983 but failed to report conditions requiring a license restriction.
05000327/FIN-2009005-022009Q4SequoyahReactor Trip due to Inadequate Transformer Bus Duct Maintenance ProcedureA self-revealing finding was identified for an inadequate maintenance procedure which was used to perform a periodic maintenance activity to clean and inspect the bus duct associated with the D common station service transformer (CSST). This resulted in the bus duct being left in a condition that allowed for water intrusion to occur, which led to a fault that caused a loss of one offsite power supply and an automatic reactor trip of both units with main feedwater unavailability. The licensee entered this issue into the corrective action program (CAP) as PER 166884. The finding was determined to be greater than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the use of an inadequate procedure directly contributed to the loss of one offsite power supply and an automatic reactor trip of both units with main feedwater unavailability. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be applicable to a Phase 2 analysis since the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigating systems will not be available. Using IMC 0609 Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, a Phase 2 analysis was performed using the site specific risk-informed inspection notebook. The finding was assumed to affect the initiating event likelihood (IEL) of a Transient With Loss of Power Conversion System (TPCS), since power availability to the unit boards affects reactor coolant pump function as well as main condenser availability. A regional Senior Reactor Analyst performed a Phase 3 Significance Determination Process evaluation. The evaluation concluded the finding was of very low safety significance (Green) based on an assumed unavailability of the CSST B fast transfer function of 0.11/yr. No cross-cutting aspect was identified since the issue was not reflective of current licensee performance, in that the inadequate maintenance procedure was implemented in December 2006
05000327/FIN-2009005-012009Q4SequoyahFailure to Evaluate Mission Dose for Manual Operator Actions Required by Plant ProceduresThe inspectors identified a non-cited violation (NCV) of Units 1 and 2 Technical Specification 6.8, Procedures & Programs, for the licensees failure to follow procedures involving the review and approval of revisions to a plant abnormal operating procedure (AOP). The incorporation of manual operator actions regarding closure of the containment equipment hatch in the event of a fuel handling accident into a plant AOP without performing a mission dose evaluation resulted in the likelihood that personnel involved with the activity would receive a dose in excess of regulatory limits for occupational exposure. The licensee entered this issue into their corrective action program as PERs 167420 and 167428. The finding was determined to be greater than minor because it was associated with the program and process attribute of the occupational radiation safety cornerstone and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The cornerstone objective was affected since adequate worker protection from exposure to radiation was not ensured through the AOP revision process. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because it did not affect the licensees ability to assess dose, did not involve an overexposure or substantial potential for overexposure, and was not related to ALARA planning. No cross-cutting aspect was identified since the issue was not reflective of current licensee performance, in that the performance deficiency occurred in 2004
05000327/FIN-2009003-012009Q2SequoyahContainment Equipement Hatch Closure Capability during Fuel Handling AccidentThe inspectors identified an unresolved item (URI) involving the licensees implementation of a commitment concerning closure of the containment equipment hatch in the case of a fuel handling accident in the containment building. This issue is unresolved pending further NRC inspection and review of additional information to be provided by the licensee. On April 1, 2009, during a refueling outage of Unit 1, the inspectors performed a routine inspection of the licensees ability to close the containment equipment hatch should residual heat removal (RHR) cooling be lost while the reactor coolant system (RCS) is open to the containment atmosphere. The inspectors noted that the licensee had performed an analysis of containment environmental conditions, following the loss of RHR, to determine how much time was available to close the hatch prior to conditions within the containment becoming so harsh as to potentially prohibit hatch closure. The licensee conducted a drill to ensure that personnel could be mobilized and the hatch could be closed within analyzed time limits. Upon review of the analysis, the inspectors noted that it covered only the condition of loss of RHR. Further inspection revealed that the licensee also intended to leave the equipment hatch open during fuel movement in the containment building, and that the plant Technical Specification (TS) Bases specified that a method to promptly close the containment equipment hatch during movement of irradiated fuel assemblies will be in place. This commitment was introduced into the plants licensing basis as part of a license amendment issued on October 28, 2003, which was TS change 02-08, Partial Scope Implementation of the Alternate Source Term and Revision of Requirements for Closure of the Containment Building Equipment Door During Movement of Irradiated Fuel. This TS change revised LCO 3.9.4 to remove the requirement for the containment equipment hatch to be closed during movement of fuel within the containment, unless the fuel had been irradiated (i.e. part of a critical core) within the previous 100-hour period. The change included a commitment to establish the capability to close the equipment hatch in the event of a fuel handling accident. This commitment was reflected in the revision to the TS Bases, as noted above, and was implemented through a revision to the licensees procedure AOP-M.04, Refueling Malfunctions, revision 6, on October 25, 2004. The inspectors requested that the licensee provide a copy of the analysis which determined that the environmental conditions which would be present within the containment following the design basis fuel handling accident would not prohibit plant personnel from closing the hatch in accordance with the commitment reflected in the TS Bases. The licensee was unable to provide such an analysis. The inspectors noted that licensee design basis document SQN-DC-V-21.0, Sequoyah Nuclear Plant Environmental Design, revision 20, identified that a fuel handling accident is among those design basis accidents that could result in plant personnel approaching GDC-19 dose limits, and requires that a post accident mission dose analysis shall be performed where plant personnel are required to enter vital areas of the plant via a preplanned procedure to maintain the plant design basis following a fuel handling accident. The inspectors also noted that plant procedure EPM-7-1, EOI Administrative Controls, revision 8, required that the mission dose estimate be evaluated, prior to implementing new manual operator actions in EOPs or AOPs, for all activities required to be performed outside the control room in the event of a design basis accident as identified by SQNDC- V-21.0. The inspectors requested the mission dose calculation for hatch closure following a design basis fuel handling accident. The licensee was unable to provide such a calculation. It was identified that this evaluation had not been performed in conjunction with revision 6 to AOP-M.04. These issues were entered into the licensees corrective action program as PERs 167420 and 167428. Pending additional information from the licensees evaluation of their ability to close the equipment hatch following a fuel handling accident, this item is identified as URI 050000327,328/2009003-01, Containment Equipment Hatch Closure Capability During Fuel Handling Accident
05000321/FIN-2008005-012008Q4HatchFailure to Report a Reportable ConditionA NRC-identified violation of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73, Licensee Event Report System, was identified when the licensee did not recognize the loss of all three main control room (MCR) air handling units (AHUs) was a reportable condition. Consequently, the licensee failed to make an eight hour report as required by 10 CFR 50.72 and submit a licensee event report (LER) within 60 days as required by 10 CFR 50.73. This violation does not apply to Unit 1 because it was in a refueling outage and the AHUs were not required to be operating. This violation has been entered into the licensees CAP as CR 2008111957. Failure to recognize the loss of the MCREC system safety function was reportable is a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function of event assessment. The inspectors determined this finding was a SL IV violation because the failure to report this condition did not substantively impact the Agency\'s regulatory responsibilities and the Agency would not have responded in a significantly different manner had the information been properly reported. This finding had the cross-cutting aspect of evaluating for reportability in the area of Problem Identification and Resolution (P.1(c)) because the licensee evaluated reportability only for the entry into TS LCO 3.0.3. (Section 4AO5)
05000321/FIN-2008005-032008Q4HatchLicensee-Identified ViolationTechnical Specification 5.7.1.a requires, in part, that each high radiation area, in which the intensity of radiation is > 100 mrem/hr but < 1000 mrem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on October 14, 2008, the licensee was transferring Unit 1 condensate phase separator resin to the vendors equipment for receiving the resin; however, the licensee did not barricade nor conspicuously post the areas that contained the pipes used for transferring the resin as a high radiation area. Licensee evaluations performed after the event showed that the intensity of radiation was >100 mrem/hr but <1000 mrem/hr measured at 30 cm from the pipe surfaces in those areas. This finding was entered in the licensees corrective action program as Condition Report 2008110421. This finding is of very low safety significance because there was no evidence of unauthorized worker entry into the area and no unexpected /unintended radiation exposures to licensee personnel
05000321/FIN-2008005-022008Q4HatchLicensee-Identified Violation10 CFR 50.73(a)(2)(i)(B) requires in part that the licensee shall report any condition which was prohibited by technical specifications. Contrary to this, on May 19, 2008, the licensee determined that pressure boundary leakage resulting from a weld failure in an instrumentation sensing line was discovered on March 8, 2005, and not reported. This issue was entered in the licensees corrective action program under CR 2008103067. This finding is of very low safety significance because the leak was very small and within the RCS leakage accident analysis
05000259/FIN-2007003-012007Q2Browns FerryReactor Core Isolation Cooling System Loss of Configuration ControlThe inspectors identified an unresolved item (URI) involving a mispositioned and faulted switch on the 1C 250 VDC Reactor Motor-operated Valve (RMOV) Board used for Unit 1 RCIC operation from outside the main control room. Description: On June 15, while conducting a system alignment walkdown, inspectors found two out-of-position RCIC barometric condenser pump emergency handswitches on the 1C 250 VDC RMOV Board with respect to the 1-OI-71, Reactor Core Isolation Cooling System, Attachment 2, Panel Lineup Checklist. Both handswitches were found in the STOP position versus the required START position per the checklist. To address this problem, the licensee initiated PER 126345. The specific handswitches in question were: 1-HS-71-31C, RCIC Vacuum Pump 1-HS-71-29C, RCIC Vacuum Tank Condensate Pump Upon notification of the mispositioned switches, Operations commenced an independent performance of 1-OI-71, Attachment 2, RCIC Panel Lineup Checklist which would reposition the above handswitches in addition to verifying all other RCIC panel components. While performing this checklist, operators discovered that the RCIC Barometric Condenser Vacuum Pump Backup Control Switch, 1-HS-71-31C, on the 1C 250 V RMOV Board, was mechanically bound in the STOP position. The licensee initiated Work Order (WO) 07-719158-000 to repair the switch and PER 126352 to document an unplanned 30-day LCO entry into Technical Specification 3.3.3.2.A.1 for an inoperable backup control system function of the RCIC Barometric Condenser Vacuum Pump. After further review, Operations also discovered a difference between the 1-OI-71, Attachment 2 checklist and the Monthly Emergency Control Switch Verification 0-GOI- 300-1, Operator Round Log, Attachment 15.12, Monthly Emergency Control Switch Verification - Unit 1, which had placed the aforementioned handswitches in the STOP position. The inspectors verified that the correct switch positions were START, as required by 1-OI-71, Attachment 2. The licensee initiated Procedure Change Request (PCR) 07002587 to correct the GOI-300-1 attachment. In evaluating the implications of past operability of the Unit 1 RCIC system given the mispositioned switches (one of which was faulted), the inspectors first reviewed drawings and wiring schematics to verify that the emergency control handswitches in question would not have adversely impacted the RCIC pump automatic and manual control circuit when other emergency control handswitches in the circuit, separate switches from those in question, were in the NORMAL position. Based on this review, the inspectors concluded that the mispositioned switches would not have adversely affected RCIC pump automatic operation, or manual operation from the main control room (MCR). However, with the emergency control handswitches in EMERGENCY, the Start/Stop handswitches in question would be in the control circuits. Therefore, the inspectors examined whether the RCIC system would be capable of performing its safety function during an event necessitating MCR abandonment (requiring th emergency control handswitches in EMERGENCY) with a loss of the RCIC Vacuum Pump due to the faulted switch. In particular, the inspectors needed additional information from the licensee in order to determine whether a sufficiently high temperature environment (turbine gland seals and valve packing exhausting to the RCIC room) could be created that would cause an automatic isolation of the RCIC System steam supply thereby rendering RCIC inoperable. In order to fully assess the enforcement implications and safety significance of this issue, additional information from the licensee will be needed. Consequently, pending the receipt of additional information and further review by the NRC (e.g., determination of the safety significance), this issue will be identified as URI 05000259/2007003-01, Reactor Core Isolation Cooling System Loss of Configuration Control.
05000321/FIN-2007002-012007Q1HatchManual Operator Actions Allowed Due to an Inadequate 10 CFR 50.59 EvaluationAn NRC-identified Severity Level IV non-cited violation (NCV) was identified for an inadequate 10 CFR 50.59 evaluation. The licensee proceduralized manual actions in place of automatic actions to close the door to an adjacent office to maintain the main control room (MCR) pressure boundary operable without prior NRC review and approval. Violations of 10 CFR 50.59 potentially impact the NRCs ability to perform its regulatory function. Therefore, this finding was subject to traditional enforcement. This finding was determined to be of very low safety significance because the door only impacted the radiological response of the MCR, the door was capable of being closed, and procedural guidance was in place to close the door. In accordance with the NRC Enforcement Policy, Supplement I.D.5, this finding was determined to be a Severity Level IV violation. This violation has been entered into the licensees corrective action program as Condition Report (CR) 2006112331.
05000321/FIN-2006004-012006Q3HatchFailure to Report Safety Relief Valve Test Results Outside Technical Specification LimitsAn NRC-identified non-cited violation of 10 CFR 50.73 a2iB was identified for failure to report past conditions prohibited by plant Technical Specifications TS. The inspectors determined that, during the most recent operating cycle for both Units 1 and 2, several main steam safetyrelief valves exceeded the TS lift setting tolerance. These represented reportable events. This finding was evaluated using the traditional enforcement process because the failure to accurately report events has the potential to impact the NRCs ability to perform its regulatory function. This finding was determined to be a Severity Level IV violation based on Supplement I of the NRC Enforcement Policy.
05000321/FIN-2004004-012004Q3HatchFailure to Perform 10 CFR 50.59 EvaluationThe inspectors identified a SL-IV non-cited violation (NCV) when the licensee failed to perform a 10 CFR 50.59 screening or evaluation for failing closed the RHRSW minimum flow valves. This evaluation was required to demonstrate that the change did not create the possibility of a malfunction of equipment important to safety with a different result than any previously evaluated in the updated final safety analysis report (UFSAR). As described in the NRC Enforcement Policy, violations of 10 CFR 50.59 are considered to potentially impede or impact the regulatory process. Therefore, the significance of this finding was assessed using the Enforcement Policy Supplements. The inspectors determined the finding was more than minor because the inspectors could not reasonably determine that the change would not ultimately require NRC approval, based on the lack of licensee documentation related to compensatory measures, short or long term corrective actions. Based on the inspectors review of the licensee's 10 CFR 50.59 evaluation, this violation was determined to be of very low safety significance.
05000346/FIN-2003016-012002Q3Davis BesseFailure to Proerly Implement the Boric Acid Corrosion Control and Corrective Action ProgramsThe inspectors identified an apparen violation involving the failure to follow the corrective action procedure and take timely corrective action for a condition adverse to quality, in that the licensee failed to implement a modification to permit complete inspection and cleaning of the reactor vessel head and CRDM nozzles. This finding is more than minor because the corrosion of the reactor head and the resulting cavity represented a significant loss of the design basis barrier integrity. The significance of this finding will be determined by the Significance Determination Process for the issue, which was begun following the Augmented Inspection Team activities
05000346/FIN-2001015-012001Q4Davis BesseSL Iv Violation of 10 CFR 50.7The NRC concluded that a security officer was discriminated against for engaging in protected activities within the scope of 10 CFR 50.7, "Employee Protection." A security supervisor subjected the officer to a fact-finding meeting on January 12, 2001, and placed a copy of the documentation from the meeting in the security officer's personnel file. The NRC determined that these actions were taken, at least in part, as a result of the security officer engaging in protected activity when he identified and documented in the condition report the potential security department training deficiency. The NRC issued a Notice of Violation by letter dated December 20, 2001, requiring a response by the licensee (VIO 50-346/01-15-01).
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Douglas-20Simpkins-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"