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05000458/FIN-2018301-0230 September 2018 23:59:59River BendExam Security Compromise While Administering Simulator JPMThe team reviewed a self-revealed Green non-cited violation of 10 CFR Part 55.49, Integrity of examinations and tests, for the licensees compromise of a simulator JPM during exam administration. Specifically, on June 8, 2018, during a review of the draft exam, the NRC identified that the draft examiner guide for simulator JPM S7 contained exam material for simulator JPM S8. The licensee removed the erroneous exam material from the examiner guide, but failed to evaluate the extent of condition to ensure that the applicant handout did not also contain exam material from JPM S8. On July 26, 2018, while performing JPM S7, an applicant was reviewing his handout and found that the last page was the cue sheet for JPM S8, which was intended to be administered the following day. As an immediate compensatory measure, JPM S8 was administered to all applicants the day of the compromise, to prevent any opportunity for the compromise to spread or for applicants to specifically prepare for the JPM. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2018-04141 The compromise of initial license exam material to an applicant before the intended date of administration is a performance deficiency. The inspectors determined the performance deficiency was more than minor, and therefore a finding, because it adversely affected the Human Performance attribute of the Mitigating Systems cornerstone, the objective of which is to ensure the availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Additionally, if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern, by causing license decisions to be made based on compromised exams which were not administered in an equitable or consistent manner. The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix I, Licensed Operator Requalification Significance Determination Process. The team determined that the finding was of very low safety significance (Green) because the finding was related to initial license exam security (block 10), but did not cause an actual negative effect on the equitable and consistent administration of the initial license exam (block 11), because the exam team took immediate compensatory action to rearrange the exam schedule and administer the compromised JPM that day, preventing any opportunity for applicants to unduly prepare for the compromised JPM, or for other applicants to learn of it. The finding was related to the cross-cutting aspect of Evaluation in the cross-cutting area of Problem Identification and Resolution because after the NRC identified that draft simulator JPM S7 erroneously contained exam material from simulator JPM S8, the licensee failed to evaluate the extent of condition of this error to ensure that other exam materials were not also affected.
05000461/FIN-2018003-0130 September 2018 23:59:59ClintonFailure to Revise an Operability Evaluation When No Longer Meeting a Compensatory MeasureThe inspectors identified a Green finding for the failure to revise an operability evaluation when no longer meeting a compensatory measure, in accordance with OPAA115, Operability Determinations, Revision 21. Specifically, the licensee failed to revise the operability evaluation documented in EC 387664 when no longer maintaining the Division 1 and Division 2 safety-related buses in a split bus configuration from November 2017 through June 2018.
05000461/FIN-2018003-0230 September 2018 23:59:59ClintonMinor ViolationTitle 10 CFR 50, Appendix B,Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established procedure CPS 8219.01, Personnel Airlock Maintenance, Revision 19, as the implementing procedure for performing maintenance on a safety-related personnel airlock, an activity affecting quality.Procedure CPS 8219.01, Section 2.1.4 states in part, Personnel Airlock Maintenance Checklist, shall be filed with completed work documents.Contrary to the above, on March 30, 2018, the licensee failed to follow Section 2.1.4 of procedure CPS 8219.01. Specifically, the licensee failed to file the personnel airlock maintenance checklist with the completed work documents. After the inspectors questioned the whereabouts of the checklist, it was discovered that it was not used when performing the repair or post maintenance test on the personnel airlock even though the procedure directs personnel to record pertinent data on the checklist during the maintenance activity. Screening: The inspectors determined the performance deficiency was minor because it was determined to be a documentation issue and the values required to be documented in the checklist were satisfactory, therefore, there was no adverse impact. The licensee documented this issue in AR 4126058, NRC ID: Documentation Deficiency Identified.Violation: This failure to comply with 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.Licensee Event Report 05000461/201800100 is closed.
05000440/FIN-2018003-0130 September 2018 23:59:59PerryApplication of ASME Code Case N5133 for the Emergency Service Water Piping DegradationsThe inspectors identified an Unresolved Item concerning the Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, and ASME Code requirements for the ESW piping systems with regards to the licensees application of ASME Code Case N5133, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping, Section XI, Division 1. Updated Safety Analysis Report (USAR) Section 9.2.1 describes that the function of ESW system is to provide a reliable source of water to safety-related components required for normal and emergency reactor operation. USAR Table 3.21, Equipment Classification, delineates that the ESW piping system is safety-related and designed in accordance with the requirements of ASME Section III, Subsection ND (Class 3). The regulation in 10 CFR 50.55a(g) requires, in part, that Class 3 components and their supports meet the requirements of ASME Section XI of the ASME Boiler and Pressure Vessel (BPV) Code or equivalent quality standards. The ASME also publishes Code Cases, which provide alternatives to existing Code requirements. The NRC Regulatory Guide (RG) 1.147 identifies that Code Case N5133 provides acceptable alternatives to applicable parts of Section XI, provided it is used with any identified conditions or limitations. Code Case N5133, Section 2(d) requires that a flaw evaluation shall be performed to determine the conditions for flaw acceptance. Section 3 provides accepted methods for conducting the required analysis. In addition, Section 3 requires, in part, that nonplanar flaws shall be evaluated in accordance with the requirements in 3.2. Additionally, Section 5 requires that an augmented volumetric examination or physical measurement to assess degradation of the affected system shall be performed as follows: (a) From an engineering evaluation, the most susceptible locations shall be identified. A sample size of at least five of the most susceptible and accessible locations, or, if fewer than five, all susceptible and accessible locations shall be examined within 30 days of detecting the flaw. (b) When a flaw is detected, an additional sample of the same size as defined in 5(a) shall be examined. (c) This process shall be repeated within 15 days for each successive sample, until no significant flaw is detected or until 100 percent of susceptible and accessible locations have been examined. On June 13, 2018, a through-wall leakage on the 20 ESW piping was identified in CR 201805504. As a result, the licensee invoked the Code Case to evaluate this flaw and permit the degraded ESW piping system to remain in service for a limited period without repair/replacement. The licensees evaluation involved characterization of this flaw as nonplanar, and subsequently, the methodology as described in Section 3.2 of the Code Case was utilized for this nonplanar flaw. Additionally, the licensee identified the five most susceptible and accessible locations in the ESW system and performed examination in accordance with Section 5(a). From the examination of the five additional locations, another localized wall degradation was identified on the 8 ESW pipe elbow on July 10, 2018. The licensee initiated CR 201806205 to document this condition. The licensee characterized this degradation also as a nonplanar flaw, and this degradation represented approximately 80 percent wall loss from its nominal thickness. During the review of the licensee evaluation of this degraded pipe elbow, the inspectors identified that the methodology as described in Section 3.2 of the Code Case had not been utilized. Instead, the licensee elected to use an alternate methodology to evaluate and disposition for its acceptability. Furthermore, the inspectors identified that the licensee essentially redefined the term flaw in the Code Case to reflect the ASME Section XI, IWA9000 definition of the term defect. The ASME Section XI, IWA9000 defines a flaw as an imperfection or unintentional discontinuity that is detectable by nondestructive examination. It also defines a defect as a flaw (imperfection or unintentional discontinuity) of such size, shape, orientation, location, or properties as to be rejectable. With respect to the Code Case, the licensee essentially restricted the criteria for examination scope expansion only to the flaws that were rejectable; therefore, the licensee had not expanded the scope to perform examination of additional locations in accordance with Section 5(b). In essence, two items are to be further evaluated and addressed: (1) whether the use of methodology not described in the Code Case Section 3.2 was appropriate for evaluation of the nonplanar flaw on the 8 ESW pipe elbow, and (2) whether the stopping of scope expansion for examination as required by the Code Case Section 5(b) was appropriate based on the licensees redefining of the term flaw. In response to the inspectors concern, the licensee initiated CR 201808483, NRC ID: Code Case N5133 Interpretation, September 26, 2018. The licensee also plans to perform examination of five additional locations in November of 2018. This represents an item where the inspectors identified Code interpretation issues that resulted in a disagreement with the licensee. This will require additional review to determine whether a violation exists. Therefore, this issue is considered an unresolved item pending completion of inspector review and evaluation and discussion with the Office of Nuclear Reactor Regulation. Licensee Action: The licensee plans to perform examination of five additional locations in November of 2018. Corrective Action Reference: CR 201808483
05000461/FIN-2018412-0130 September 2018 23:59:59ClintonSecurity
05000416/FIN-2018003-0130 September 2018 23:59:59Grand GulfFailure to Develop Adequate Work InstructionsA self-revealed, Green finding was identified when feedwater heater drain tank oscillations caused a feedwater perturbation which required a manual reactor scram. Specifically, the licensee failed to develop appropriate work instructions for filling and venting the feedwater heater 6A level transmitters.
05000440/FIN-2018010-0130 September 2018 23:59:59PerryFailure to Correctly Establish Maintenance/Replacement Frequencyfor the WeedTemperature Transmitters In Zone FB-7The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations(CFR), Part 50.49(e)(5), Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the licensees failure to correctly establish maintenance/replacement frequency for Weed temperature transmitters installed in harsh environment. Specifically, Calculation EQ-115, Qualified Life Calculation for Weed RTD/RTDT and TC Assemblies,incorrectly established a qualified life for Weed temperature transmitters installed in Zone FB-7. The calculation determined that the qualified life for these transmitters in Zone FB-7 as 18.9 years plus accident. However, the calculation failed to account for the accident time and temperature.
05000416/FIN-2018003-0230 September 2018 23:59:59Grand GulfMinor ViolationMinor Violation: The licensee did not include any unplanned power changes as inputs for the Unplanned Power Changes per 7,000 Critical Hours performance indicator (PI) that was reported to the NRC for the second quarter 2016. Based on a plant event that took place on June 17, 2016, the inspectors noted that the PI data submitted by the licensee may not have been accurate. In response, the licensee submitted frequently asked question (FAQ) 17-01 to the reactor oversight process working group. This FAQ resulted in the determination that three unplanned power changes should have been reported associated with the event in question. Following resolution of the FAQ, the licensee reported the associated PI data. As required by 10 CFR 50.9, Completeness and accuracy of information, information provided to the NRC by a licensee shall be complete and accurate in all material respects. Contrary to the above, from July 2016 through May 2017, information provided to the NRC by the licensee was not complete and accurate in all material respects. Specifically, the data for the Unplanned Power Changes per 7,000 Critical Hours PI did not include any unplanned power changes for the second quarter 2016. Screening: The inspectors determined that this violation was of minor significance in accordance with the NRC Enforcement Policy, Section 6.9.d.11, since the PI data in question did not ultimately result in the PI changing from Green to White. Enforcement: The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2016-06028. The licensee took action to restore compliance by submitting an appropriate correction to the PI data. This failure to comply with 10 CFR 50.9 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes URI 05000416/2017001-02, Grand Gulf Unplanned Power Changes per 7000 Critical Hours Performance Indicator.
05000458/FIN-2018301-0130 September 2018 23:59:59River BendInadequate Procedure for Shutdown Operations Protection PlanThe team reviewed a self-revealed Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide accurate qualitative procedural guidance to determine Shutdown Cooling Safety Function color state in Modes 4 and 5, using OSP-0037, Rev 36, Shutdown Operations Protection Plan, Attachment 1, Shutdown Cooling Function Color States. Specifically, OSP-0037 Rev 36 defines the term Flooded Up as, Flooded Up condition requires greater than 23 ft in the Reactor Cavity and the Cavity Gate open. OSP-0037 Rev 36, Attachment 1 contains a table with 6 columns that list combinations of decay heat loads (high, medium, and low), and reactor cavity water inventory, which are used to determine Shutdown Cooling risk. Only one of the six columns uses the correctly-defined term, Flooded Up, for reactor cavity water inventory. Two of the six columns state Flooded, and three of the columns state Not FL, neither of which are defined terms in OSP-0037. This creates the potential for an operator to misinterpret the meaning of the column, and select a color code for Shutdown Cooling that represents a lower risk than is actually present. This potential was confirmed by erroneous applicant performance on the July 2018 NRC initial license exam. As an interim corrective action, the station issued a night order clarifying the typographical error, and initiated action to revise the procedure. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2018-04414 The failure to provide accurate qualitative procedural guidance to determine Shutdown Cooling Function Color State is a performance deficiency. The inspectors determined the performance deficiency was more than minor because it adversely affected the Procedure Quality attribute of the Mitigating Systems cornerstone, the objective of which is to ensure the availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the procedure errors could cause a crew to underestimate Shutdown Cooling risk, with an adverse effect on conservative implementation of defense in depth in the planning, scheduling, and implementation of outage activities. The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix G, Shutdown Operation Significance Determination Process, dated May 9, 2014. The team determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of safety function of any train or safety system for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program, with cavity either flooded or not flooded; (5) did not degrade a functional auto-isolation of RHR on low reactor vessel level; (6) did not screen as potentially risk significant due to an external event; (7) did not involve Fire Brigade training and qualification requirements, or brigade staffing; (8) did not involve the response time of the Fire Brigade to a fire; and (9) did not involve fire extinguishers, fire hoses, or fire hose stations. No Cross-Cutting Aspect is assigned, because the procedural errors were introduced in Revision 19, issued September 16, 2009, and are therefore not indicative of current licensee performance.
05000416/FIN-2018201-0130 September 2018 23:59:59Grand GulfSecurity
05000458/FIN-2018012-0230 June 2018 23:59:59River BendFailure to Identify and Correct a Broken Feedwater Chemistry ProbeTwo examples of a self-revealed non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified for the licensees failure to identify that a broken chemistry probe in the feedwater system had the potential to cause an adverse impact on plant safety, and promptly implement appropriate measures to address that condition.
05000416/FIN-2018002-0730 June 2018 23:59:59Grand GulfLoss of Shutdown CoolingA self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes
05000458/FIN-2018012-0730 June 2018 23:59:59River BendFailure to Perform 10 CFR 50.59 Evaluation for Main Feedwater System Sparger Nozzle DamageThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59 , Changes, Tests, and Experiments, for the licensees failure to provide a written safety evaluation for the determination that operation with compensatory measures for damaged feedwater sparger nozzles did not require a license amendment pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. Specifically, the licensee failed to recognize that compensatory measures prohibiting operation in single loop conditions required technical specification changes, and as such required prior NRC approval.
05000461/FIN-2018050-0330 June 2018 23:59:59ClintonEquipment Operator Rounds Points Inadequate Acceptance CriteriaOn May 17, 2018,a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self-revealed for the licensees failure to include appropriate quantitative acceptance criteria for the Division 2 EDG parameters to ensure the Division 2 EDG could perform its safety function
05000440/FIN-2018410-0130 June 2018 23:59:59PerrySecurity
05000416/FIN-2018002-0330 June 2018 23:59:59Grand GulfFailure to Adequately Test NUS Temperature SwitchA self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred.
05000458/FIN-2018012-0330 June 2018 23:59:59River BendFailure to Establish Procedural Guidance for Determining Core Flow During Unanticipated Single Loop OperationsThe inspectors reviewed a self-revealed,non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish appropriate instructions in the abnormal operating procedure for thermal hydraulic instabilities. Specifically, the procedural step for determining core flow when in single loop operations at low power did not provide appropriate instructions to operators. As a result, station personnel could not conclusively determine core flow and inserted a manual reactor scram.
05000461/FIN-2018002-0130 June 2018 23:59:59ClintonFailure to Perform an Operability Determination for Suspected Leakage Past Shutdown Service Water Isolation ValvesThe inspectors identified a Green finding for the failure to perform an operability determination in accordance with Procedure OPAA108115, Operability Determinations (CM1). Specifically, the licensee failed to determine and document the operability status of the shutdown service water system and the ultimate heat sink after the discovery of leakage past the 1CC075A and 1CC076A isolation valves.
05000416/FIN-2018002-0830 June 2018 23:59:59Grand GulfPerformance of Surveillance Testing Following Maintenance on Containment AirlockThe inspectors identified a Green non-cited violation of 10CFRPart50,AppendixB, Criterion XI, Test Control, for the licensees failure to perform surveillance testing of containment airlock seals under appropriate conditions. The licensee failed to appropriately control the sequence of maintenance and testing activities to ensure that surveillance testing was not performed subsequent to maintenance which could affect the validity of surveillance test results.
05000458/FIN-2018406-0130 June 2018 23:59:59River BendSecurity
05000461/FIN-2018002-0330 June 2018 23:59:59ClintonMinor ViolationDuring the inspection quarter, the inspectors reviewed a significant number of licensee CAP documents to assess the following performance attributes: complete, accurate, and timely documentation of the identified problem in the CAP; evaluation and timely disposition of operability and reportability issues; consideration of extent of condition and cause, generic implications, common cause, and previous occurrences; classification and prioritization of the problems resolution commensurate with the safety significance; and identification of negative trends associated with human or equipment performance that can potentially impact nuclear safety. Minor Performance Deficiency: The inspectors determined that issues which could impact the operability of TS-related equipment were generally entered into the CAP in a timely manner. However, operability determinations were not always performed within the timeframes established in Section 4.1 of Procedure OPAA108115, Operability Determinations (CM1), because some issue reports were not directly routed to the operating shift crew for review. The CAP software program used by the licensee included a standard set of questions which were normally answered by the individual entering the issue into the CAP. Depending on the answers to the questions, the CAP document routing could automatically bypass the operating shift crew for review. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The inspectors did not identify any instance where the failure to perform a timely operability determination had a significant consequence on licensed activities. However, the inspectors discussed the vulnerability between the CAP and the operability determination process with the licensee. The licensee implemented a standing order to require a shift review by the operating crew of condition reports not directly routed to the shift. In addition, the licensee is trending the number of condition reports which are returned by the Station Ownership Committee to the shift for review to determine whether further actions are warranted. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000458/FIN-2018002-0130 June 2018 23:59:59River BendFailure to Correct Inadequate Technical Specification Pressure Temperature CurvesThe inspectors identified a Severity Level IV non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, after receiving vendor information indicating that existing technical specification pressure temperature (PT) curves were inadequate, the licensee failed to promptly identify and correct the condition through the license amendment process.
05000416/FIN-2018002-0430 June 2018 23:59:59Grand GulfHigh Radiation Area Boundary ViolationA self-revealed, Green non-cited violation of Technical Specification 5.7.1 was identified when an individual received a dose rate alarm when the individual failed to comply with established radiological barriers and protective measures and entered a high radiation area. Specifically, an individual leaned over a high radiation area barricade rope, thereby entering the high radiation area. The individuals radiation work permit (RWP) did not permit entry into a high radiation area.
05000458/FIN-2018012-0430 June 2018 23:59:59River BendFailure to Submit a Licensee Event Report for a Manual ScramThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to submit a required licensee event report (LER). Specifically, on February 1, 2018, after an unexpected trip of the recirculation pump B, the licensee initiated a manual scram of the reactor that was not part of a preplanned sequence and failed to submit an LER within 60 days.
05000461/FIN-2018002-0230 June 2018 23:59:59ClintonFailure to Establish Adequate Leak Rate Test Procedures for Shutdown Service Water Isolation Valve TestingThe inspectors identified a Green finding and a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure testing of the shutdown service water (SX) isolation valves was performed with procedures which: (1) incorporated the requirements and acceptance limits contained in applicable design documents; and (2) included provisions for assuring that all prerequisites for the given test had been met. Specifically, the licensee failed to establish leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions for ensuring the required differential test pressure was met during testing.
05000458/FIN-2018406-0230 June 2018 23:59:59River BendSecurity
05000461/FIN-2018002-0430 June 2018 23:59:59ClintonMinor ViolationThe inspectors reviewed AR 4082490, Reactor SCRAM from Trip of 1AP07EJ. The inspectors selected this sample for review due to the safety significance of the Division 1 and 2 safety-related transformers, which is the subject of the AR. This review focused on actions associated with newly installed Divisions 1 and 2 4160V to 480V transformers. As appropriate, the inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition report and other related condition reports: classification and prioritization of the resolution of the problem commensurate with safety significance; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. As a result of this review the inspectors identified the following minor violation: Minor Violation: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow procedures associated with the CAP. Specifically, on May 10, 2018, the licensee identified discrepant results while testing safety-related transformers 0AP06E2 and 1AP12E2 but failed to enter this issue into the CAP in accordance with PIAA120, Issue Identification and Screening Process, Revision 8, Step 4.3.4, until prompted by the inspectors. Instead, the licensee evaluated the discrepant results within the work order and found them to be acceptable. The licensee generated AR 4137994, Insulation Power Factor Results For 0AP06E & 1AP12E, dated May 15, 2018, after being challenged by the inspectors regarding the need to enter the discrepant test results into the CAP. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The failure to document the discrepant values in the CAP did not adversely impact the safety-related transformers. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion II, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000458/FIN-2018002-0230 June 2018 23:59:59River BendEnforcement Action (EA)-18-053: Enforcement Discretion for Tornado-Generated Missile Protection Noncompliances

Title 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 2, Design Bases for Protection Against Natural Phenomena, states, in part, that systems, structures, and components (SSCs) important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes. Criterion 4, Environmental and Dynamic Effects Design Basis, states, in part, that SSCs important to safety shall be appropriately protected against dynamic effects including missiles that may result from events and conditions outside the nuclear power unit. Section 3.5.2, Structures, Systems, and Components to be Protected from Missiles, of the Updated Safety Analysis Report (USAR) details the structures that are designed to withstand tornado missile impact.On February 7, 2017, the NRC issued Enforcement Guidance Memorandum (EGM) 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance, Revision 1 (ADAMS Accession Number ML16355A286). The EGM referenced a bounding generic risk analysis performed by the NRC staff that concluded that tornado missile vulnerabilities pose a low risk significance to operating nuclear plants. Because of this, the EGM described the conditions under which the NRC staff may exercise enforcement discretion for noncompliance with the current licensing basis for tornado-generated missile protection. Specifically, if the licensee could not meet the technical specification required actions within the required completion time, the EGM allows the staff to exercise enforcement discretion provided the licensee implements initial compensatory measures prior to the expiration of the time allowed by the limiting condition for operation. The compensatory actions should provide additional protection such that the likelihood of tornado missile effects are lessened. The EGM then requires the licensee to implement more comprehensive compensatory measures within approximately 60 days of issue discovery. The compensatory measures must remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Because EGM 15-002 listed River Bend Station as a Group A plant, enforcement discretion expired on June 10, 2018. On May 10, 2018, River Bend Station submitted a request to extend the enforcement discretion period to June 10,

8 2020. On May 31, 2018, River Bend Station submitted asupplement to the May 10 request. On June 6, 2018, the NRC granted an extension to the enforcement discretion until June 10, 2020. The initial conditions of Design Basis Accident (DBA) and transient analyses in the USAR, Chapter 6 and Chapter 15, assume Engineered Safeguards Features (ESF) systems are operable. The AC, DC, and AC vital bus electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, reactor coolant system, and containment design limits are not exceeded.The onsite standby power source for each 4.16 kV ESF bus is a dedicated emergency diesel generator (EDG). An EDG starts automatically on a loss of coolant accident signal (i.e., low reactor water level signal or high drywell pressure signal) or on an ESF bus degraded voltage or under voltage signal. In the event of a loss of preferred power, the ESF electrical loads are automatically connected to the EDGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a DBA such as a loss of coolant accident. Standby service water (SSW) is required by Technical Specification 3.7.1. The ultimate heat sink (UHS) consists of one 200 percent capacity cooling tower and one 100 percent capacity water storage basin. The UHS basin capacity is required by Regulatory Guide 1.27 and USAR 9.2.5 to maintain a minimum of 30 days inventory to mitigate the consequences of a DBA without replenishment. The UHS is designed to perform its safety function assuming a single failure coincident with a loss of offsite power and with respect to the 30 day mission time assuming a single division of SSW is in service.The safety design bases of these SSCs includes ensuring the SSCs are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).On May 4, 2018, the licensee identified vulnerabilities in the EDG building, the control building, and the SSW cooling tower where tornado-born missiles could potential render safety-related equipment contained in these buildings inoperable. Potentially affected equipment included all three EDGs, Division II DC electrical power distribution subsystem, residual heat removal (RHR) pumps B and C, SSW pumps A, B, C, and D, Division I standby cooling tower fans, and multiple Division I SSW motor operated valves. These vulnerabilities were identified as part of the licensees review of Regulatory Information Summary 2015-06, Tornado Missile Protection. These issues were entered into the corrective action program as Condition Reports CR-RBS-2018-02687, 02768, and 02775.Corrective Actions: As a result of these issues, the licensee declared all three EDGs, the Division II DC electrical power distribution subsystem, RHR pumps B and C, SSW pumps A, B, C, and D, Division I standby cooling tower fans, and multiple Division I SSW motor operated valves inoperable, complied with the applicable technical specification action statements, initiated Condition Reports CR-RBS-2018-02687, 02768, and 02775, invoked the EGM discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The licensee instituted compensatory measures intended to reduce the likelihood of tornado missile effects. These included verifying that guidance was in place for severe weather procedures, abnormal and emergency operating procedures, and FLEXsupport guidelines, verifying that training on these
procedures was current, and verifying that a heightened level of awareness of the vulnerability was established.Corrective Action Reference(s) : CR-RBS-2018-02687, CR-RBS-2018-02768, and CR-RBS-2018-02775Enforcement:Violations: Technical Specification 3.8.1 requires, in part, that three diesel generators shall be operable in Modes 1, 2, and 3. Technical Specification 3.8.1.H requires entry into LimitingCondition for Operation 3.0.3 when three or more required AC sources are inoperable. Limiting Condition for Operation 3.0.3 requires that action shall be initiated within one hour to place the unit in Mode 2 within 7 hours, in Mode 3 within 13 hours, and in Mode 4 within 37 hours.Contrary to the above, prior to May 4, 2018, three diesel generators were not operable, and action was not initiated to place the unit in Mode 2 within 7 hours, in Mode 3 within 13 hours,and in Mode 4 within 37 hours. Specifically, the EDG building was not designed to withstand the effects of natural phenomena, such as tornadoes. The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02687. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Technical Specification 3.8.9 requires, in part, that the Division II AC and AC vital bus electrical power distribution subsystems shall be operable in Modes 1, 2, and 3. Technical Specification 3.8.9.D requires the station to take action to place the unit in Mode 3 within 12 hours when one or more AC or AC vital bus electrical power distribution subsystems have been inoperable for more than 8 hours. Contrary to the above, prior to May 4, 2018, the Division II AC and AC vital bus electrical power distribution subsystems were not operable for more than 8 hours, and action was not initiated to place the unit in Mode 3 within 12 hours. Specifically, the control building was not designed to withstand the effects of natural phenomena, such as tornadoes. The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02768. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Technical Specification 3.5.1 requires, in part, that each emergency core cooling system (ECCS) injection subsystem shall be operable in Modes 1, 2, and 3. Technical Specification 3.5.1.D requires the station to take action to place the unit in Mode 3 within 12 hours when two ECCS injection subsystems have been inoperable for more than 72 hours. Contrary to the above, prior to May 4, 2018, two required ECCS injection subsystems that included RHR pumps B and C were inoperable for more than 72 hours, and action was not initiated to place the unit in Mode 3 within 12 hours. Specifically, the control building was not designed to withstand the effects of natural phenomena, such as tornadoes.The licensee
initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02768. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Technical Specification 3.7.1 requires, in part, that two SSW subsystems shall be operable in Modes 1, 2, and 3. Technical Specification 3.7.1. H requires the station to take action to place the unit in Mode 3 within 12 hours when both pumps associated with one SSW subsystem have been inoperable for more than 72 hours. Contrary to the above, prior to May 4, 2018, SSW pumps P2B and P2D, associated with SSWsubsystem B, were inoperable for more than 72 hours, and action was not initiated to place the unit in Mode 3 within 12 hours. Specifically, the SSW cooling tower was not designed to withstand the effects of natural phenomena, such as tornadoes. The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02775. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Severity/Significance: Not ApplicableBasis for Discretion: The NRC exercised enforcement discretion in accordance with EGM 15-00, Revision 1, because the licensee implemented initial compensatory measures in accordance with the EGM.
05000416/FIN-2018002-0530 June 2018 23:59:59Grand GulfFailure to Follow Procedure Requirements Resulting in Unplanned DoseA self-revealed, Green non-cited violation of Technical Specification 5.4.1 was identified when an individual alarmed a personnel contamination monitor upon exit from the radiologically controlled area. Specifically, the licensee failed to follow procedures to establish a decontamination plan or procedure, conduct a specific pre-job brief addressing appropriate contamination risk, and receive approval by radiation protection supervision prior to conducting decontamination activities on thereactor pressure vessel(RPV) O-rings
05000458/FIN-2018012-0530 June 2018 23:59:59River BendFailure to Develop an Adequate Operational Decision-Making Issue for Compensatory Measures Related to a Degraded Condition of the Feedwater System Sparger NozzlesThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to develop an adequate Operational Decision-Making Issue (ODMI) document per Procedure EN-OP-111, Operational Decision-Making Issue Process. Specifically, the licensee failed to develop an ODMI that provided adequate guidance to the operators for safely operating the plant with degraded feedwater sparger nozzles.
05000461/FIN-2018050-0130 June 2018 23:59:59ClintonFailure to Follow Multiple ProcedureOn May 17, 2018, a To-Be-Determined (TBD) finding and an associated Apparent Violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and Technical Specification 3.8.2, Condition B.3,were self-revealed for the licensees failure to follow multiple procedures that affected quality.This resulted in the unavailability and inoperability of the Division 2 Emergency Diesel Generator when it was relied upon for plant safety
05000416/FIN-2018002-0130 June 2018 23:59:59Grand GulfFailure to Institute Effective Corrective Action to Preclude RepetitionAn NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective.
05000461/FIN-2018002-0530 June 2018 23:59:59ClintonMinor ViolationThe inspectors reviewed AR 4116223, Blown Fuses during CPS 9080.23 8.4 for Fast Transfers. The inspectors selected this sample for review due to repetitive fuse failures within the safety-related Division 3 NUS Modules dating back to 2013. As appropriate, the inspectors verified the following attributes during their review: complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery; consideration of the extent of condition, generic implications, common cause, and previous occurrences; evaluation and disposition of operability/functionality/reportability issues; classification and prioritization of the resolution of the problem commensurate with safety significance; identification of corrective actions, which were appropriately focused to correct the problem; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. Description: While reviewing the historical ARs associated with the NUS fuse failures, the inspectors discovered licensee information indicating the NUS fuse failures were likely caused by voltage/current transients within the upstream, safety-related 480V to 120V regulating transformer. The purpose of the transformer was to regulate voltage and current to the downstream components including the NUS modules. However, degradation in the transformers ability to regulate voltage and current levels could create a condition where the voltage and current levels exceeded the NUS fuse rating causing fuse failure. The licensee documented the potential transformer degradation issue on September 20, 2013, in AR 1561455, Division 3, Group 1 Instruments Found De-energized during CPS 9080.23, Specifically, the licensee stated, The most probable cause of the failure of the NUS modules was the transient voltage overshoot of the regulating transformer causing the transient protection varistors on the five NUS modules to actuate, drawing a near fault current until the individual and line feed fuses blew. Station procedure PI-AA-125, Corrective Action Program, defined equipment failure as, damage to or degradation of a system, structure or component that may cause or contribute to the event. Based on the information documented in AR 1561455, the licensee identified transient voltage overshoots in the 480V to 120V regulating transformer, which was a degraded condition causing the NUS modules to fail. Per the licensee definition this would constitute an equipment failure. No further action was taken to identify and correct the regulating transformer degradation until the transformer failed on March 18, 2018, impacting multiple pieces of safety-related Division 3 equipment. Minor Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to this requirement, on September 20, 2013, the licensee identified a failure of the 480V to 120V regulating transformer, which manifested itself as a voltage overshoot causing the failure of the NUS modules, but failed to take actions to correct the condition. On March 18, 2018, the regulating transformer subsequently degraded further causing it to fail in a manner that tripped the upstream breaker and impacted additional pieces of safety-related Division 3 equipment. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. Specifically, the inspectors determined that although the transformer failure affected Division 3 equipment, the failure would not have impacted the Division 3 equipments ability to respond to a DBE or the capability to shut down the reactor and maintain it in a safe shutdown condition. Enforcement: The failure to comply with 10 CFR 50, Appendix B, Criterion XVI, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000458/FIN-2018012-0130 June 2018 23:59:59River BendFailure to Conduct Adequate Transient Snap Shot Assessment Following Recirculation Pump TripThe inspectors identified a finding for the licensees failure to adequately validate simulator response during a transient snap shot assessment following an unexpected trip of reactor recirculation pump A on December 19, 2012.
05000416/FIN-2018002-0630 June 2018 23:59:59Grand GulfImproper Evaluation and Resolution of Intermediate Range MonitorNoise Leads to Manual Reactor ShutdownA self-revealed, Green non-cited violation of 10CFRPart50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement appropriate corrective actions related to intermediate range monitor (IRM) nuclear instrument (NI) electronic noise spiking. The failure to implement adequate corrective actions over the course of at least 5 years resulted in a plant shutdown due to declaration of multiple IRM channels inoperable while in Mode 2.
05000458/FIN-2018012-0630 June 2018 23:59:59River BendFailure to Provide Adequate Procedures for Post-Scram RecoveryThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to establish, implement and maintain a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, Procedure OSP-0053, Emergency and Transient Response Support Procedure, Revision 22, which is required by Regulatory Guide 1.33, inappropriately directed operations personnel to establish feedwater flow to the reactor pressure vessel using the main feedwater regulating valve as part of the post-scram actions. This resulted in the main feedwater regulating valves being operated outside their design limits. This resulted in catastrophic failure of the main feedwater regulating valve variseals and subsequent damage to multiple fuel assemblies.
05000461/FIN-2018050-0230 June 2018 23:59:59ClintonFailure to Promptly Identifya Condition Adverse to QualityOn May 17, 2018,a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were self-revealed for the licensees failure to promptly identify that the safety-related Division 2 EDG had its starting air receivers isolated, which was a condition adverse to quality that rendered the EDG inoperable and unavailable
05000440/FIN-2018002-0130 June 2018 23:59:59PerryFailure to Control Transient Combustible Materials in a Designated Combustible Control ZoneThe inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Perry Operating License Condition 2.C(6), Fire Protection, for the licensees failure to control transient combustible materials in a designated combustible control zone within fire area 1AB1g on Auxiliary Building elevation 574 10. Specifically, on May 16, 2018, the inspectors identified transient combustible materials left unattended in the designated combustible control zone in the corridor outside the emergency core cooling system (ECCS) pump rooms, which exceeded the ten pound limit established in the Fire Protection Program document, PAP1910, for ordinary combustibles (loose) in designated combustible control zones without a transient combustible permit.
05000416/FIN-2018002-0230 June 2018 23:59:59Grand GulfFailure to Follow ASME Requirements for Maintaining Inservice Inspection (ISI) Cycles and Perform ASME Required Inservice Inspections within the Scheduled ISI CycleThe inspector identified 15 examples of a Green non-cited violation (NCV)of 10 CFR 50.55(a)(g)(4)(ii), which requires that inservice examination of components classified as American Society of Mechanical Engineers (ASME), Section XI, Code Class 1, Class 2, and Class 3 be conducted during successive 120-month inspection intervals, and requires compliance with the requirements of the latest edition and addenda of the ASME Code (and all its paragraphs) applicable to the specific interval, including maintaining the 120-month inspection interval in accordance with the ASME Code, Section XI, Paragraph IWA-2430. Specifically, the licensee inappropriately adjusted its second inservice inspection 120-month cycle, and failed to perform VT-3 and MT examinations of 15 class 1, class 2, and class 3 components, including the high pressure core spray pump attachment weld and reinforcing band before the third inservice inspection cycle expired on November 30, 2017, as required by 10CFR50.55(a)(g)(4)(ii).
05000416/FIN-2017011-0331 March 2018 23:59:59Grand GulfFailure to Conduct Common Cause Failure Evaluation in Response to Inoperable Emergency Diesel GeneratorThe inspectors identified three instances of a non-cited violation of Technical Specification 3.8.1, AC Sources Operating, for the licensees failure to take required actions for an inoperable emergency diesel generator. Specifically, after classifying the Division I or Division II emergency diesel generator as inoperable on the basis of nonconforming conditions, and after failing to either verify that the opposite train emergency diesel generator was not inoperable due to common cause failure within 24 hours or conduct a surveillance run on the opposite train emergency diesel generator within 24 hours, the licensee failed to enter Mode 3 within 12 hours as required by Technical Specification 3.8.1, Actions B.3.1, B.3.2, and G.1, respectively. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11393. The licensee initiated corrective actions to conduct an adverse condition analysis. The failure to take required actions for an inoperable emergency diesel generator was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment reliability attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Actions B.3.1 and B.3.2 of Technical Specification 3.8.1 exist to ensure the availability, reliability, and capability of at least one emergency diesel generator in scenarios where there is a potential for a common cause failure of both emergency diesel generators, and the licensee took neither action. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of either the Division I or Division II emergency diesel generator for greater than its technical specifications allowed outage time. The finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee failed to use a consistent, systematic approach to make decisions. Specifically, the licensee failed to review the required actions of the applicable technical specification to ensure that all of those actions would be properly carried out (H.13).
05000461/FIN-2018001-0231 March 2018 23:59:59ClintonFailure to Identify a Single Point Vulnerability Results in Manual Reactor ScramA self-revealed Green finding was identified for the licensees failure to identify a single point vulnerability in accordance with procedure ERAA2004, Revision 1. Specifically, during a site single point vulnerability review of the feedwater system, the licensee failed to identify a single point vulnerability that subsequently resulted in a loss of a feedwater heating string. The loss of the heater string caused a drop in temperature in the reactor of 100 degrees which prompted a manual scrambe initiated by the operators
05000416/FIN-2018001-0231 March 2018 23:59:59Grand GulfFailure to Follow Procedure when Returning Containment Airlock to Operable StatusThe inspectors reviewed a self-revealed non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow written procedures for returning a technical specification component to service. Specifically, the licensee failed to follow Procedure 01-S-06-12, Surveillance Program Procedure, Revision 112, when performing a completion review on the 208 foot elevation inner door personnel airlock seal test, which is a Technical Specification required surveillance.
05000458/FIN-2018001-0231 March 2018 23:59:59River BendInstallation of an Incorrectly Specified Relay Causes Plant Transient and Reactor ScramThe inspectors reviewed two examples of a self-revealed finding for the licensees installation of an incorrectly specified relay in 1) the control circuitry for the feedwater level control systemand 2) the turbine generator voltage regulator circuitry. In each instance, the incorrectly specifiedrelay failed in service, causing a plant transient and automatic reactor scram
05000416/FIN-2017011-0431 March 2018 23:59:59Grand GulfFailure to Install the Residual Heat Removal Pump A Mechanical Seal in Accordance with ProceduresThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures. Specifically, on September 22, 2016, maintenance did not install seal assembly drive pins in accordance with Step 7.8.2 of Procedure 07-S-14-279, Revision 007. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2017-08269 and CR-GGN-2017-11009. The licensee implemented immediate corrective actions by declaring the pump inoperable and replacing the mechanical seal. The failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, 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. Specifically, on September 22, 2016, mechanical maintenance installed the residual heat removal pump A seal drive pins backwards. As a result, the drive pins damaged the seal and on August 23, 2017, caused an unisolable leak from the seal. This resulted in unplanned inoperability and unavailability of the residual heat removal pump A from August 23, 2017, through August 25, 2017, when the mechanical seal was replaced. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, avoid complacency, because the licensee failed to recognize and plan for the possibility of mistakes, and individuals failed to implement appropriate error reduction tools. Specifically, the licensee failed to use appropriate error reductions tools such as self-check or peer checking which resulted in incorrect performance of procedural steps (H.12)
05000416/FIN-2018001-0331 March 2018 23:59:59Grand GulfInadequate Procedural Guidance Which Resulted in Control Room Air Conditioning InoperabilityThe inspectors reviewed a self-revealed non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to have adequate procedural guidance while performing a standby service water surveillance procedure. Specifically, the licensees procedural guidance was not adequate to prevent the control room air conditioning subsystem B compressor from starting while condenser cooling water was isolated, which caused damage and rendered the subsystem inoperable and unavailable.
05000416/FIN-2017011-0531 March 2018 23:59:59Grand GulfFailure to Correct Control Room Boundary Door Resulted in Loss of Safety FunctionThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to appropriately correct a condition adverse to quality. Specifically, the control room envelope door had been documented in several condition reports for not consistently working properly. Subsequent to these condition reports, on July 9, 2017, the door was opened and did not close automatically, and therefore the door was left in an unsecured position. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-06705. The licensee restored compliance by securing the door and replacing the hinge bushings to ensure the door would close properly. The failure to correct a condition adverse to quality for a control room envelope boundary door was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the structures, systems, and components and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (functionality of the control room) protect the public from radionuclide releases caused by accidents or events. Specifically, on July 9, 2017, since the licensee had not corrected the adverse conditions identified on the control room envelope door, the door was left in an unsecured position and resulted in the station declaring both trains of standby fresh air inoperable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system, and did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The period of the barrier in the open position was of short duration, approximately 1 minute, and therefore did not result in significant risk input. This finding had a cross-cutting aspect in the area of problem identification and resolution, resolution, because the licensee did not take corrective actions in a timely manner commensurate with their safety significance. Specifically, the licensee did not ensure proper priority of corrective actions on the degraded control room envelope boundary door (P.3).
05000416/FIN-2018001-0431 March 2018 23:59:59Grand GulfInadequate Procedural Guidance which Resulted in Undemanded Control Valve Movements and Manual ScramThe inspectors reviewed a self-revealed non-cited violation of Technical Specification 5.4.1.a, Procedures, associated with the licensees failure to provide appropriate procedural guidance while performing calibration of a steam line compensator. Specifically, Work Order 4449267, Task 14 did not contain adequate instructions to calibrate a steam line compensator circuit card potentiometer, which led to undemanded control valve opening and closing and a subsequent manual reactor scram.
05000416/FIN-2017011-0131 March 2018 23:59:59Grand GulfFailure to Categorize Condition Reports for Significant Conditions Adverse to Quality as Required by ProceduresThe inspectors identified five examples of a finding for the licensees failure to categorize and evaluate conditions in accordance with procedural requirements. Specifically, the licensee did not categorize adverse conditions that represented the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 28. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to categorize conditions that represent the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, root cause evaluations, corrective actions to prevent recurrence, and effectiveness reviews are used per licensee Procedure EN-LI-102 to ensure availability and reliability of structures, systems, and components are maintained. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently evaluate the conditions during initial screening led to the incorrect categorization of the condition reports (H.13)
05000416/FIN-2017011-0631 March 2018 23:59:59Grand GulfFailure to Perform Functionality Assessments as Required by ProceduresThe inspectors identified a finding for the licensees failure to follow Procedure EN-OP-104, Operability Determination Process, Revisions 10 through 12. Specifically, the licensee did not perform functionality assessments for adverse conditions on the offgas system as required by the procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11265. The licensee initiated corrective actions to perform functionality assessments for the conditions identified and to evaluate any potential programmatic issues. The failure to perform functionality assessments required by Procedure EN-OP-104 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to perform functionality assessments could affect the availability and reliability of the offgas system to maintain the doses associated with releases to the environment as low as reasonably achievable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it involved the Effluent Release Program, it did not impair the ability to assess dose, and did not exceed the 10 CFR Part 50, Appendix I, or 10 CFR 20.1301(d) limits. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition adverse conditions associated with the offgas system resulted in the station not performing required functionality assessments (H.13)
05000440/FIN-2018001-0131 March 2018 23:59:59PerryFailure to Notify the NRC within 60 Days of a Condition Prohibited by Technical SpecificationsThe inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to report a condition that was prohibited by the plants Technical Specifications to the U.S. Nuclear Regulatory Commission (NRC) within 60 days. Specifically, the licensee did not report a condition that, as determined by the NRC, rendered the Division 2 Diesel Generator (DG) inoperable for a period longer than the Technical Specification allowed completion times of its associated required actions.