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05000416/FIN-2017003-012017Q3Grand GulfIsolation of Reactor Core Isolation Cooling System during Surveillance TestingThe 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 accomplish quality related activities in accordance with Surveillance Procedure 06-IC-1E31-A-1004, RCIC Equipment Room High Temperature Calibration Channel A, Revision 106. Specifically, on August 21, 2017, the licensee did not follow Step 5.15.4, which states, Identify and disconnect field lead located at Terminal EE-50 in 1H13-P632. This step was not performed correctly; therefore, the reactor core isolation cooling (RCIC) system isolation feature was not bypassed. When performing the next step, an inadvertent isolation of the RCIC system occurred. On August 21, 2017, the licensee restored compliance by performing actions to restore the leads to the correct location and performing the surveillance test satisfactorily. This issue has been entered into the licensees corrective action program as Condition Report CR-GGN-2017-08246.The failure to follow Surveillance Procedure 06-IC-1E31-A-1004 was a performance deficiency. This 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, the licensees failure to follow Surveillance Procedure 06-IC-1E31-A-1004 resulted in unplanned inoperability and unavailability of the reactor core isolation cooling system. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; did not represent a loss of safety function; did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time, and did not represent 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 for greater than 24 hours. The inspectors determined that the finding had a field presence cross-cutting aspect within the human performance area because licensee management failed to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the performer in the field was a supplemental worker that was observed by a licensee instrumentation and controls technician. The technician telephoned the supervisor to ensure that they were performing the steps correctly, and the supervisor did not go into the field to verify the step was performed correctly (H.2).
05000416/FIN-2017003-022017Q3Grand GulfLicensee-Identified ViolationTitle 10 CFR 55.49 requires, in part, that licensees shall not engage in any activity that compromises the integrity of any test or examination required by 10 CFR 55.49. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. Contrary to the above, on August 8, 2017, Grand Gulf Nuclear Station engaged in an activity that compromised the integrity of an examination required by 10 CFR 55.49. Specifically, the licensee left written exam material from a previous weeks exam unattended. The previous weeks exam contained half of the current weeks written exam material. The exam material was marked appropriately and located within an instructors office. This finding was determined to be of very low safety significance (Green) because the finding did not have an actual effect on the equitable and consistent administration of the biennial requalification exam cycle. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2017-07723.
05000458/FIN-2017007-022017Q2River BendFailure to Perform an Adequate Operability Determination for a Condition Identified During an NRC WalkdowGreen. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, between June 15, 2017, and June 28, 2017, the licensee failed to address the operability of a terminal block installed within an unsealed junction box. In response to this issue the licensee performed an operability determination to ensure that the terminal block would perform its design function in this condition. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2017-05084. The team determined that the failure to perform an adequate operability determination was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure operability of valve E51-AOVF054 and its associated circuits would impact the operability of the reactor core isolation cooling system. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to perform an adequate operability determination for an identified condition (P.3)
05000458/FIN-2017007-012017Q2River BendFailure to Evaluate the Extent of Condition for a Degraded 4.16KV Magne Blast Safety-Related Circuit BreakerGreen. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, on October 28, 2014, the licensee failed to perform extent of condition on other safety-related 4.16KV Magne Blast circuit breakers due to the failure of 4.16KV Magne Blast circuit breaker ACB03 on bus E22-S003, with damaged brush and misaligned brush holder of the circuit breaker charging motor, in accordance with Procedure EN-OP-104, Operability Determination Process. Failure to perform this evaluation could adversely impact safety-related circuit breakers. In response to this issue the licensee reviewed their breaker performance records to assure that no additional failures had occurred and revised the procedure to assure that extent of condition is addressed. This finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2017-05078. The team determined the failure to evaluate the impact of a damaged brush and misaligned brush holder of the charging motor of a safety-related 4.16KV Magne Blast breaker was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it related to the equipment performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to perform an extent of condition on other safety-related 4.16KV Magne Blast circuit breakers due to the failure of E22-S003 safety-related circuit breaker ACB03, 4.16KV Magne Blast breaker with damaged brush and misaligned brush holder could adversely affect the ability of these breakers to perform their safety functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance 3 associated with conservative bias because the licensee failed to ensure that individuals used decision-making practices that emphasized prudent choices (H.1)
05000275/FIN-2015003-012015Q3Diablo CanyonFailure to Document an Adequate Evaluation for a Change in Seismic Load Combination MethodologyThe inspectors identified a non-cited violation of 10 CFR 50.59(d)(1) which requires, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to paragraph (c) of this section. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2). Specifically, the licensee changed the method for combining earthquake loads and loss of coolant accident loads from the absolute summation method to square root sum of the squares (SRSS) method without sufficient justification to demonstrate the change did not require prior NRC approval. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate changes to determine if prior NRC approval is required was a performance deficiency. The licensee entered the issue into the corrective action program as Notification 50811191. In accordance with the licensees corrective action program, this issue will be addressed by the licensee through a re-evaluation of the methodology change and the required actions that need to be taken by the licensee will be implemented. Additionally, the licensee performed an operability determination for the affected structures, systems, and components that established a reasonable expectation for operability pending final resolution of the issue. This performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to determine that use of SRSS in the Watts Bar safety evaluation report cited in the PG&E evaluation represented a change in a method of evaluation, in that the Watts Bar safety evaluation report was very narrow in scope and not appropriate for the intended application at Diablo Canyon. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not result in the inoperability of the system. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the inspectors also evaluated the performance deficiency using traditional enforcement. Since the violation is associated with a Green finding having very low safety significance, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. This finding had a cross cutting aspect in the area of human performance associated with design margins because individuals failed to ensure margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2015003-022015Q3Diablo CanyonFailure to Secure a Locked High Radiation AreaThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1(a), Procedures, for failure to secure a locked high radiation area. Specifically, the padlock on the Letdown Filter 1-1 locking bar was found unlocked. Upon discovery, the licensee guarded the area until properly secured. This issue was entered into the licensees corrective action program as Notification 50710852. The failure to secure a locked high radiation area was a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, failure to adequately secure the locked high radiation area could result in unintended exposure to high levels of radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation was of very low safety significance (Green) because: (1) it was not an as low as reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding had an avoid complacency cross-cutting aspect, in the area of human performance, because individuals failed to recognize and plan for the possibility of mistakes, even while expecting positive outcomes. Specifically, licensee personnel failed to ensure that the padlock was secured after completing the task (H.12).
05000397/FIN-2015003-012015Q3ColumbiaFailure to Maintain Seismic Instrumentation Functional to Alert Plant Operators of Ground Motions Exceeding the Operating Basis EarthquakeThe inspectors identified a finding associated with the licensees failure to maintain seismic instrumentation functional as required by Licensee Controlled Specification 1.3.7.2, Seismic Monitoring Instrumentation. Specifically, because of inadequate calibration procedures, several as-left setpoints for the seismic response spectrum recorders indicating lights were non-conservative relative to their function to alert operators of ground motion exceeding the operating basis earthquake (OBE). Following discovery of this issue, the licensee recalibrated the seismic response spectrum recorders using OBE ground motions as the upper tolerance. The licensee entered this issue into their corrective action program as Action Request 333996. The performance deficiency was more than minor because it affected the configuration control attribute of the Mitigating Systems Cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in seismic instruments calibrations that were nonconservative relative to their function to alert plant operators that a shutdown is required. NRC regulations require a plant shutdown since systems necessary for continued operation without undue risk to the health and safety of the public are not designed to remain functional, in all cases, following an OBE. The inspector performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent 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 for greater than 24 hours. Additionally, the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The finding does not have a cross-cutting aspect since the configuration control error is associated with an instrument setpoint change request from 1990 and therefore not reflective of current licensee performance.
05000397/FIN-2015003-022015Q3ColumbiaNon-Conservative Shutdown Criteria in Earthquake Abnormal ProcedureThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, Procedures, for the failure to maintain an adequate abnormal procedure for earthquakes. Specifically, the licensee failed to establish appropriate shutdown criteria for earthquakes that exhibit ground motion exceeding the operating basis earthquake (OBE). The licensees shutdown criteria would allow for continued operations if ground motion at a single frequency exceeded the design response spectrum. In response to this issue, the licensee initiated corrective actions to change the stations earthquake abnormal procedure to provide shutdown criteria consistent with the original licensing basis of the facility. The licensee entered this issue into their corrective action program as Action Request 336875. The performance deficiency was more than minor because it affected the procedura adequacy attribute of the Mitigating Systems Cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in shutdown criteria that would allow for continued operations following events where ground motion at a single frequency exceeded the desig response spectra. NRC regulations require a plant shutdown since systems necessary for continued operation without undue risk to the health and safety of the public are not designed to remain functional, in all cases, following an OBE. The inspector performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technica specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. Additionally, the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The finding does not have a cross-cutting aspect since the procedure error is associated with a 1996 change to the licensing basis and therefore not reflective of current licensee performance.
05000397/FIN-2015003-032015Q3ColumbiaFailure to Provide Design Control Measures for Control Room Emergency ChillersThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the design of the control room HVAC system. Specifically, the licensee failed to demonstrate the ability of control room HVAC design to maintain the temperatures in the main control room below habitability and environmental qualification limits, for the duration of all accident scenarios. The licensee initiated Action Request 332565 to document the concern, issued night order 1662 to communicate the issue, aligned both control room air handling units to their respective chillers, created a quick card procedure to perform the chiller reset actions, and validated the quick card actions could be accomplished within 10 minutes. Additionally, the licensee determined that operators could restore the chillers during accident conditions within 90 minutes to prevent temperatures from exceeding equipment operability limits. The performance deficiency was more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent 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 for greater than 24 hours. This finding had a cross-cutting aspec in the area of problem identification and resolution, evaluation, in that the licensee did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee did not thoroughly evaluate the extent of condition from NRC-identified NCV 05000397/2013002-04, Failure to Obtain NRC Approval for Changes to Control Room HVAC Requirements, fo the effect of this change on other station calculations (P.2).
05000397/FIN-2015003-042015Q3ColumbiaFailure to Implement Procedures to Ensure Availability of Safe Shutdown PersonnelThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to ensure operators could perform time-critical steps for fire events. Specifically, on July 4, 2015, the licensee failed to implement written procedures to ensure that an equipment operator can complete certain post-fire safeshutdown actions within 10 minutes. In response to this conclusion, the licensee initiated Action Request 332747 to document the inability to meet the post-fire safe-shutdown actions in accordance with procedure PPM 1.3.1, Operating Policy, Programs, and Practices, Revision 119. Additionally, the licensee issued Night Order 1655, reminding all operating crews of the requirements of procedure PPM 1.3.1 for leaving the protected area. This performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating System Cornerstone and affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst performed a detailed significance determination process review using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination dated September 20, 2013 and NRC Inspection Manual 0308, Attachment 3, Appendix F, Technical Basis Fire Protection Significance Determination Process (Supplemental Guidance for Implementing IMC 0609, Appendix F) At Power Operations, dated February 28, 2005. The senior reactor analyst determined that the failure of the equipment operator to perform the certain post-fire safe-shutdown actions within 10 minutes would not adversely affect a quantitative risk assessment, and therefore this finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Teamwork, because the licensee failed to communicate and to coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the equipment operator spoke with the shift technical advisor about the need to exit the protected area at the morning turnover meeting but neither individual spoke with the control room supervisor. Communication was ineffective in that the Equipment Operator believed permission was granted and proceeded to exit the protected area (H.4).
05000397/FIN-2015003-052015Q3ColumbiaFailure to Reduce the Free Water in a Class A Unstable Resin Disposal Package to Less than 0.5 Percent of Waste VolumeThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow their Process Control Program as implemented by their solid radioactive waste system procedures. Specifically, the licensee failed to reduce the free standing liquid in a condensate filter demineralizer resin disposal package (Liner 14-033-L) to less than the required 0.5 percent of the tota waste volume. Corrective actions included retrieving the packages from waste shipment 14-32, testing each liner for free standing liquid content, and removing additional water as necessary. The licensee documented this issue in their corrective action program as Action Requests 00316555 and 00316676. The failure to follow the Process Control Program, resulting in the inadequate dewatering of radioactive waste liner contents, was a performance deficiency. The inspectors determined that the performance deficiency was more than minor, because it adversely affected the Public Radiation Safety cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released in the public domain Specifically, the failure to ensure that the free standing liquid in the radioactive waste liner shipped to US Ecology did not exceed 0.5 percent of the total waste volume subjected the disposal facility to the possibility of improper handling of the waste. Using Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined the violation was of very low safety significance (Green) because: (1) radiation limits were not exceeded, (2) there was no breach of the package during transit, (3) there were no Certificate of Compliance issues, and (4) the low level burial ground nonconformance did not involve a 10 CFR 61.55 waste under-classification. The inspectors determined that the finding has a design margin crosscutting aspect in the area of human performance, because the licensee failed to operate and maintain the radioactive waste dewatering system within the vendor design margins when changes were made to the operating procedures (H.6).
05000275/FIN-2015003-032015Q3Diablo CanyonLicensee-Identified ViolationPG&E Part 72 license SNM-2511, Condition #11 requires, in part, that The licensee shall operate the installation in accordance with the Technical Specifications in the Appendix. Appendix Technical Specification 2.1.2 requires in part that Preferential fuel loading shall be used during uniform loading. Contrary to the above, from July 18, 2009 through June 6, 2015, PG&E failed to load 19 casks in accordance with Appendix Technical Specification 2.1.2 for preferential fuel loading. Specifically, the licensee failed to load fuel assemblies with longest cooling times in the periphery of the basket. This violation was identified by PG&E and placed in their corrective action program. The licensee submitted Event Notification 51134 to the NRC on June 6, 2015 and later updated the Event Notification on June 9, 2015. Following the event notification, PG&E submitted a 30-day report to the NRC on July 6, 2015 (ML15187A239). This violation did not have any safety impact, in that all fuel assemblies met the requirements for burn-up, decay heat, and cooling time. All fuel and casks remain in a safe and analyzed condition. However, in order to re-establish compliance with PG&Es Part 72 license, the licensee must submit a license amendment request to the NRC. In accordance with the NRC Enforcement Policy Section 2.2 and IMC 0612 Section 03.23, Part 72, ISFSI inspection findings follow the traditional enforcement process and are not dispositioned through the Reactor Oversight Process or the Significance Determination Process. The violation screened as having very low safety significance, Severity Level IV, and is being treated as an NCV, consistent with Section 2.3.2 a. of the Enforcement Policy. The violation was determined to be more than minor since the violation requires DCPP to request a License Amendment from the NRC for their Part 72 license in order to restore compliance for the 19 affected casks. The violation was entered into the licensees corrective action program as Notifications 50706314 and 50706501. Following identification of the issue, the licensee performed an assessment that showed the casks would continue to perform their design function. Corrective actions for this issue included issuing the revised procedure, performing an extent of condition review, providing just-in-time training to Reactor Engineering staff involved, and added an independent third party review requirement for fuel contents loaded into the canister.
05000397/FIN-2015003-062015Q3ColumbiaImplementation of Enforcement Guidance Memorandum (EGM) 11-003, Revision 2During Refueling Outage 22 in May June 2015, Columbia Generating Station implemented the guidance of Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel, dated December 13, 2013. Consistent with EGM 11-003, Revision 2, secondary containment operability was not maintained during operations with a potential for draining the reactor vessel activities, and required action C.2 of Technical Specification 3.6.4.1 was not completed. The inspectors reviewed this licensee event report for potential performance deficiencies and violations of regulatory requirements. The inspectors reviewed the stations implementation of the EGM 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations logged all potential for draining the reactor vessel activities in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the licensee maintained reactor vessel water level at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1, Secondary Containment requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from May 13 - June 13, 2015, Columbia Generating Station performed a total of five operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Action Request 329328. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request (LAR) to resolve the issue for its plant which the NRC staff LAR acceptance review finds acceptable in accordance with LIC-109, Acceptance Review Procedures. The generic solution will be a generic change to the Standard Technical Specifications, and the NRC will publish a notice of availability (NOA) for the TS solution in the Federal Register. Each licensee that receives discretion must submit its amendment request within 12 months of the NRC staffs issuance of the NOA. Licensees may submit LARs to adopt the NRC-approved approach or to propose an alternative approach for their plants. This licensee event report is closed.
05000416/FIN-2015001-012015Q1Grand GulfFailure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Pump House Ventilation System Due to Degraded RelaysThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with the division 1 and 2 standby service water pump house ventilation systems. Specifically, in June 2011, the licensee identified that relays associated with the standby service water system pump house ventilation system failed due to age/environmental degradation, which resulted in an unplanned inoperability of the standby service water system. However, the licensee did not implement timely corrective actions for replacing these relays, which resulted in the inoperability of the division 1 standby service water system in December 2014, and again in January 2015. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2015-00739. The short-term corrective actions included replacing all of the division 1 and 2 standby service water ventilation pump house relays in February and early March 2015. The inspectors determined that the failure to take timely corrective actions to replace degraded relays in the standby service water pump house ventilation system was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating System 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. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, the inspectors determined the issue to be of very low safety significance (Green) because all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. The inspectors determined that this performance deficiency was not indicative of current plant performance, and therefore no cross-cutting aspect was considered.
05000416/FIN-2015001-022015Q1Grand GulfFailure to Follow a Procedure Resulting in the Unplanned Inoperability of the Reactor Core Isolation Cooling SystemThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for failure to follow a procedure which resulted in the unplanned inoperability of the reactor core isolation cooling system. This occurred when licensee technicians tested for continuity between incorrect points, while performing surveillance activities related to the residual heat removal system. This resulted in an invalid group 4 isolation signal and an isolation of the reactor core isolation cooling steam supply. The licensee entered this issue into the corrective action program as Condition Report CR-GGN- 2015-01532, and took immediate corrective actions to stop the residual heat removal system surveillance activity and restore the reactor core isolation cooling system to service. The failure to properly follow the surveillance procedure, which resulted in the unplanned inoperability of the reactor core isolation cooling system, was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Mitigating Systems Cornerstone. Specifically, the licensees failure to properly follow the surveillance procedure resulted in the unplanned inoperability of the reactor core isolation cooling system, which adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) in that the issue did not affect the design or qualification of the reactor core isolation cooling system; did not represent a loss of the reactor core isolation cooling system function (in that the isolation could have been promptly reset by procedures, had the system operation been required); and did not represent loss of function for greater than the Technical Specification allowed outage time. The inspectors determined this finding had cross-cutting aspect in the area of human performance associated with avoiding complacency, in that the I&C technicians did not implement appropriate error reduction tools to ensure the meter was connected to the correct points, which resulted in the invalid group 4 isolation signal, and inoperability of the reactor core isolation cooling system (H.12).
05000416/FIN-2015001-062015Q1Grand GulfFailure to Adequately Establish Commercial-Grade Items as Basic ComponentsThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the suitability of replacement parts that were procured from commercial suppliers. Specifically, the inspectors noted that none of the tests specified by the licensee were sufficient to ensure that the seismic qualification of an auxiliary relay had been maintained. The finding was entered into the licensees corrective action system as Condition Report CR-GGN-2014-05049. The performance deficiency is more than minor, and therefore a finding, because it was associated with the design control 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. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because the licensee performed an operability determination, which evaluated the safety impacts of postulated relay chatter during a seismic event, for the applications in which these relays were installed. The licensees subsequent operability evaluation determined that potential relay chatter would not impact the safety-related functions of the relays in the applications in which they were installed. Thus, all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. A cross-cutting aspect is not being assigned to this finding.
05000416/FIN-2015001-032015Q1Grand GulfEmergency Action Level Scheme for Nonfunctional Seismic MonitorThe inspectors identified a non-cited violation of 10 CFR 50.54(q)(2) for the licensees failure to follow and maintain the effectiveness of an emergency plan that meets the requirements of the planning standard 50.47(b)(4), which requires that a standard emergency classification and action level scheme, is in use by the licensee. Specifically, the licensee had identified, on October 15, 2013, that the seismic monitoring instrumentation was non-functional, but had not further evaluated the plant configuration, and the effect on emergency action level declaration capabilities for seismic events. The licensee documented this issue in Condition Report CR-GGN-2015-00713. The corrective actions, based on CR-GGN-2013-06514, were implemented, and a new seismic monitor was installed, tested, and brought into service on January 30, 2015. The licensees inability to promptly declare Emergency Action Level (EAL) HA6, as required in the approved emergency classification and action level scheme per 10 CFR Part 50.47(b)(4), was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Emergency Preparedness Cornerstone and adversely affects the cornerstone objective to ensure 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. Specifically, it negatively impacts the cornerstone attribute of procedure quality in that the plant configuration prohibited the timely declaration of the facility EALs, as written. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the issue affected the Emergency Preparedness Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 23, 2014, the inspectors determined that the issue is of very low safety significance (Green) because an Emergency Action Level was rendered ineffective such that HA6 would not be declared, consistent with Table 5.4-1 and Figure 5.4-1. The inspectors determined the finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation, in that the organization did not thoroughly evaluate issues to ensure that resolutions address causes, and extent of conditions, commensurate with their safety significance; in that while following Technical Requirements Manual requirements for a non-functional piece of equipment (seismic monitor), the complete effect was not evaluated to ensure the EALs were still capable of being implemented (P.2).
05000416/FIN-2015001-042015Q1Grand GulfFailure to Properly Calibrate Main Steam Line Radiation Monitors and Containment/Drywall High Range Radiation MonitorsThe inspectors identified a non-cited violation of 10 CFR 20.1501(c) for the licensees failure to properly calibrate the main steam line radiation monitors and the containment/drywell high range radiation monitors. The violation was of very low safety significance and was entered into the licensees corrective action program as Condition Report CR-GGNS-2015-01832. The failure to properly calibrate radiation monitors was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely affects the cornerstone objective to ensure adequate protection of employee health and safety and is associated with the cornerstone attribute of plant instrumentation. Specifically, the failure to properly calibrate radiation monitors impacts their ability to be used to assess dose rates. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding to be of very low safety significance because it was not an as low as reasonably achievable (ALARA) issue, there was no overexposure or substantial potential for overexposure, and the licensees ability to assess dose was not compromised. This finding has a cross-cutting aspect in the resources component of the human performance area because the licensee did not ensure that calibration procedures were adequate, nor was proper calibration equipment designed, characterized, and made available (H.1).
05000416/FIN-2015001-052015Q1Grand GulfFailure to Establish, Implement, and Maintain Appropriate Changes to the Offsite Dose Calculation Manual For REMP Airborne SamplingThe inspectors identified a non-cited violation of Technical Specification 5.5.1, Offsite Dose Calculation Manual (ODCM). Specifically, when changes were made to the Offsite Dose Calculation Manual in 1997, the licensee failed to establish an airborne sampling location for a community with the highest deposition factor (D/Q) for the site. As immediate corrective actions, the licensee evaluated their Offsite Dose Calculation Manual, evaluated the dose differential for the monitoring locations, and developed a plan to meet the environmental sampling requirements. The issue was documented in Condition Report CR-GGNS-2015-01835. The failure to establish an air sampling location in the vicinity of a community having the highest D/Q was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely affects the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the environment and public domain. Specifically, the failure to maintain the Offsite Dose Calculation Manual with appropriate airborne radionuclide sampling requirements adversely impacts the licensee's ability to validate offsite radiation dose assessments for members of the public under certain effluent release conditions. Using Inspection Manual Chapter 0609, Appendix D, dated February 12, 2008, Public Radiation Safety Significance Determination Process, the inspectors determined that the violation had very low safety significance because it involved the environmental monitoring program. This finding has a cross-cutting aspect in the procedure adherence component of the human performance area because licensee personnel failed to follow procedures when they determined the airborne sampling locations for the updated Radiological Environmental Monitoring Program (H.8).
05000416/FIN-2014005-012014Q4Grand GulfFailure to Assure Quality Installation on RCIC Steam LineThe inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to assure quality installation of the steam line tubing of the reactor core isolation cooling (RCIC) system. Specifically, the licensee failed to assure that rated performance limits of the ferrule connection, installed at the tee between the steam line and the pressure transmitter tube line, were met during initial installation. This failure resulted in an unplanned inoperability of the RCIC system. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2014-06792. As an immediate corrective action, the licensee replaced the tubing, the failed transmitter, and recalibrated the instruments. Furthermore, the licensee revised their system operation procedure for the RCIC system. This revision requires all steam isolation valves to be closed during this test, and that system recovery starts by opening Valve 1E51F076 (warming bypass valve around the 1E51F063) to allow adequate warming of the steam lines after isolation. The inspectors determined that the failure to assure quality installation of the ferrule connection on the steam line flow Transmitter 1E31N083B was a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the design control attribute of the Mitigating Systems Cornerstone. Specifically, failure to assure steam lines in the RCIC system meet rated performance limits, may result in the unavailability and unreliability of a system that is relied upon to respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that the issue required a detailed risk evaluation by the regional senior reactor analyst. This was because the finding represented an actual loss of a safety function due to the RCIC system being a single train system that was out of service for approximately 40 hours for repairs. The senior reactor analyst determined the change to the core damage frequency was 8.7E-8/year, and since the change to core damage frequency was less than E-7, no evaluation of external events or the large early release frequency was required. The finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect, as the performance deficiency is not reflective of current plant performance.
05000416/FIN-2014005-022014Q4Grand GulfLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations, Appendix E to Part 50, Section V, Implementing Procedures states, in part, that licensees who are authorized to operate a nuclear power facility shall submit any changes to the emergency plan or procedures to the Commission, as specified in 10 CFR 50.4, within 30 days of such changes. Title 10 of the Code of Federal Regulations, Section 50.54(q)(5) states, in part, that licensees shall submit a report of changes made after February 21, 2012, that includes a summary of its analysis, within 30 days after the change is put into effect. Contrary to the above, Grand Gulf Nuclear Station did not submit changes to emergency plan implementing procedures within 30 days of such changes, and did not submit a summary of its analysis of the changes within 30 days after the changes were put into effect. Specifically, the license did not submit changes to the following procedures; EN-EP-305, Emergency Planning 10CFR50.54(Q) Review Program, Revision 3, EN-EP-306, Drills and Exercises, Revisions 4 and 5, EN-EP-307, Hostile Action Based Drills and Exercises, Revision 2, EN-EP-308, Emergency Planning Critiques, Revision 2, EN-EP-310, Emergency Response Organization Notification System, Revisions 1 through 3, EN-EP-311, Emergency Response Data System (ERDS) Activation Via the Virtual Private Network (VPN), Revision 2, EN-EP-313, Offsite Dose Assessment Using the Unified RASCAL Interface, Revision 0, EN-EP-801, Emergency Response Organization, Revision 8, EN-TQ-110, Emergency Response Organization Training, Revision 7, and EN-TQ-110-01, Fleet E-Plan Training Course Summary, Revision 10. The licensee did not have a process to ensure that fleet procedures necessary to implement the site emergency plan were submitted to the NRC in accordance with the requirements of Appendix E to 10 CFR 50. This violation was evaluated using the NRC Enforcement Policy because the licensees failure to submit required procedures affected the NRCs ability to perform adequate regulatory oversight. The significance of the violation was evaluated at Severity Level IV (Section 6.6.d of the Enforcement Policy) because it did not affect the licensees ability to perform notification or assessment during an emergency. This issue has been entered into the licensees corrective action program as Condition Reports CR-HQN-2014-00380, CR-HQN-2014-00597, and CR-GGN-2014-05539.
05000416/FIN-2014003-012014Q2Grand GulfFailure to Promptly Reinstate an Essential-Critical Preventative Maintenance Task for a High-Critical ComponentThe inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix 8, Criterion XVI, Corrective Action, for the failure to promptly reinstate an essential-critical preventative maintenance task after they identified that it had been improperly retired. Specifically, the licensee did not reinstate and complete Preventive Maintenance Task PMRQ 50024451-04 prior to the failure of diode CR6 on May 21, 2013, which resulted in the division 2 diesel generator failing its monthly functional test and the licensee declaring it inoperable. The operators secured the diesel generator and wrote Condition Report CR-GGN-2013-03423 documenting the issue. The licensee performed a Failure Modes Analysis evaluation to determine the possible cause for the observed conditions. During troubleshooting efforts, the licensee addressed the potential transformer (PT1 ), the potential transformer's fuses, inline fuses, and the voltage regulator circuit bridge diodes. The Failure Modes Analysis evaluation showed that all of the listed components were in satisfactory condition, except that one of the six diodes used in the voltage regulator circuit diode bridge, Diode CR6, had shorted. The licensee replaced the shorted diode and returned the diesel generator to operational status on May 24, 2013. The licensee's failure to implement PMRQ 50024451-04 after discovering it had been improperly retired was a performance deficiency, in that it represented a failure to promptly correct a condition adverse to quality. The performance deficiency is more than minor and therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone's objective of ensuring the availability, reliability, and capability of systems that respond to prevent undesirable consequences. Specifically, Diode CR6 remained in the voltage regulator circuit bridge until it failed, thereby triggering a failure of the division 2 diesel generator, which caused the diesel generator to be inoperable. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, the inspectors determined that the issue required a detailed risk evaluation because the finding represents an actual loss of function of a singl,e train for greater than its Technical Specification allowed outage time. The total exposure period was 15 days. The allowed outage time was 14 days. The senior reactor analyst performed a detailed risk analysis and determined the de!ta-CDF was less than 1.0 x 10-6 and the delta-LERF was less than 1.0 x 1 o-7 , therefore this finding was of very low safety significance (Green). The apparent cause of this finding was that the licensee did not recognize the risk of not performing the preventive maintenance task, which ied to the decision to exclude the task from the division 2 allowed outage time schedule. Therefore, the finding has a cross-cutting aspect in the human performance area associated with conservative bias because the licensee did not use decision-making practices that emphasize prudent choices over those that are simply allowable.
05000416/FIN-2014003-022014Q2Grand GulfLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, states, in part, activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance wit these procedures. Contrary to the above, the licensee failed to assure that activitie affecting quality were prescribed by documented instructions of a type appropriate to th circumstances. Specifically, the licensee failed to meet the requirements of Electrica Standard ES03, Electrical Standard for Installation of Cables, Revision 1, in tha Chico A potting compound was not used to seal the drywell electrical penetration. A immediate corrective actions, the licensee removed the instrument cables and seale the penetration. The licensee entered this issue in the corrective action program unde Condition Report CR-GGN-2014-02141. Furthermore, the licensee evaluated th potential impact the open 4-inch penetration would have on the suppression pool' suppression capability and determined that having an open 4-inch diameter penetratio in the drywell did not cause the drywell bypass leakage criteria to be exceeded. Usin Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, Table 2, Cornerstones Affected by Degraded Condition or Programmati Weakness, the inspectors determined this issue affected the Barrier Integrit Cornerstone. Using Manual Chapter 0609, Appendix A, Significance Determinatio Process (SOP) for Findings at Power, June 19, 2012, Exhibit 3, Barrier Integrit Screening Questions, the inspectors determined that this finding represented an actua open pathway in the physical integrity of reactor containment (valves, airlocks, etc.) containment isolation system (logic and instrumentation), and heat removal components Using Manual Chapter 0609, Appendix H, Containment Integrity Significanc Determination Process, dated May 6, 2004, the inspectors determined that this findin would not influence the likelihood of accidents leading to core damage. However, sinc this finding involved components significant to suppression pool integrity/scrubbing tha are important to LERF, the inspectors determined a detail risk analysis needed to b performed by a senior risk analyst. The senior risk analyst performed a detailed ris evaluation and determined that since the function of the systems/components did no fail, even with the failed penetration, and there was no failure of the safety function Therefore, this finding is of very low safety significance.
05000416/FIN-2014002-022014Q1Grand GulfFailure to Control a Locked High Radiation Area Due to Unsecured Highly Radioactive Materials Stored in the PoolThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.7.3, resulting from the licensees failure to control a high radiation area with radiation levels greater than 1000 millirem per hour. As immediate corrective actions, the licensee stopped the work activity, placed a senior radiation protection technician in control of the area, surveyed all affected areas, and properly posted and controlled the area. The licensee also checked qualifications of the involved individuals and conducted a root cause evaluation for the event. This event was documented in the licensees corrective action program as Condition Reports CR-GGN-2014-02219, CR-GGN-2014-02221, and CR-GGN-2014-02224. The failure to control a high radiation area with radiation levels greater than 1000 millirem per hour was a performance deficiency and a violation of Technical Specification 5.7.3. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because it removed a barrier intended to prevent the worker from receiving unexpected dose. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with procedure adherence, because the licensee failed to follow process, procedures, and work instructions when they did not inventory and ensure control of the dry tube plunger end as it was stored in the horizontal fuel transfer system pool within containment.
05000416/FIN-2014002-012014Q1Grand GulfFailure to Ensure Scaffold Activity Would not Interfere with Fire Brigade ResponseThe inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for the failure to adhere to procedural requirements to ensure that scaffold installed in the plant would not prevent or restrict the fire brigade from accessing a certain route used for response to a fire in the area. On February 4, 2014, the licensee installed a scaffold in the containment building for an inspection. The licensees procedure required a walkdown of proposed scaffold to determine if the scaffold would prevent or restrict fire brigade access. The initial reviewer identified that the ladder to access the scaffold would restrict fire brigade access, thus the ladder was not installed until it was required. On March 1, 2014, the ladder was installed for the four hour inspection. Once completed, the licensee failed to remove the scaffold ladder to restore normal access to the area. On March 4, 2014, the inspectors identified that the scaffold ladder was still installed. The inspectors brought their concern to the licensee, who determined that the scaffold would adversely affect the response of fire brigade members to that area of containment. As an immediate corrective action, the licensee removed the scaffold ladder to allow adequate access for the fire brigade members. The licensee documented this issue in Condition Report CR-GGN-2014-02363. The failure to ensure fire brigade members had adequate access passed a scaffold installed in the containment building was a performance deficiency. The performance deficiency was more than minor and therefore a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone in that the fire brigades inability to gain access to certain areas in containment could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was shutdown for refueling outage RF19. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated February 28, 2005. The inspectors determined that Appendix G did not address fire brigade issues and solicited input from the senior reactor analyst. The senior reactor analyst performed a detailed risk evaluation and determined that Inspection Manual 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, June 19, 2012, Exhibit 2, Mitigating System Screening Questions, adequately bounded the performance deficiency. The inspectors determined that the finding involved the response time of the fire brigade to a fire, and the finding was of very low safety consequence (Green) because the fire brigades response time was mitigated by other defense-in-depth elements such as area combustible limits were not exceeded, installed fire detection systems were functional, and alternate means of safe shutdown were not impacted. Specifically, there were no combustibles in the area beyond limits, all fire detectors for the area were functional, and the plant was in a shutdown condition with the cavity flooded at the time. The apparent cause of this finding was the work groups involved did not communicate the significance of the impact the scaffold ladder had on fire brigade access to the area and the importance of having the ladder removed upon completion of the work. Therefore, the finding has a cross-cutting aspect in the human performance area associated with team work, in that the individuals and workgroups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained.
05000416/FIN-2014002-032014Q1Grand GulfLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that design control measures be established and implemented to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee failed to implement applicable design bases for the Standby Service Water System Pump 4160 VAC cables being submerged. Specifically, on January 31, 2014, the licensee did not prevent water from submerging the cables in Manhole MH-01 due to a failed sump pump. The inspectors verified that the latest megger tests for the standby service water pump cables were acceptable for demonstrating operability. This finding has been entered into the licensees corrective action program as Condition Reports CR-GGN-2014-00616 and CR-GGN-2014-00768. Using Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that this finding had very low safety significance (Green) because it did not result in the standby service water system becoming inoperable.
05000416/FIN-2014002-042014Q1Grand GulfLicensee-Identified ViolationTechnical Specification (TS) 3.3.6.1, Primary Containment and Drywell Instrumentation, requires the primary containment and drywell isolation instrumentation be operable while in Modes 1, 2, and 3. Contrary to the above, on August 3, 2013, the licensee failed to ensure the primary containment and drywell isolation instrumentation was operable prior to changing from Mode 4 (Cold Shutdown) to Mode 2 (Startup). On August 6, 2013, during a supervisory review of procedures in progress, the licensee determined that they were not incompliance with TS 3.3.6.1 due to jumpers that were installed to disable the function of the instrumentation. The licensee immediately entered the TS 3.3.6.1 Limiting Condition for Operation and associated actions. The licensee restored compliance with the TS by removing the jumpers and restoring the primary containment and drywell instrumentation to operable status and documented this issue in the corrective action program under Condition Report CR-GGN-2013-5101. Using Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that this finding had very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment or drywell and did not involve the hydrogen igniters in the reactor containment.
05000416/FIN-2013005-012013Q4Grand GulfFailure to comply with Technical Specification 3.4.11 (Section 1R20)The inspectors identified a non-cited violation of Technical Specification 3.4.11 for the failure to comply with the Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) during plant cold startups. Specifically, the PTLR had a lower limit of zero psig, and the licensee operated the reactor pressure vessel (RPV) below zero psig during the plant start-up that commenced on November 2, 2013. A review of plant data showed that the RPV pressure was maintained below zero psig for approximately 2 hours. The licensee performed an engineering evaluation and determined that the maximum compressive stress experienced by the RPV did not exceed the maximum yield strength of RPV. Immediate corrective action included revising Procedure 03-1-01-1, Cold Shutdown to Generator Carrying Minimum Load, to ensure the RPV is pressurized prior to opening the main steam isolation valves (MSIVs) and providing training on the procedural changes to all the operating crews. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-07021. The failure to comply with the RCS Pressure and Temperature Limits Report specified in Technical Specification 3.4.11 was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that a physical design barrier (reactor coolant system) protects the public from radionuclide release caused by accidents or events. Specifically, without NRC review and approval of revised pressure and temperature limits that include operating the RPV below zero psig, the inspectors did not have reasonable assurance the RPV would not be adversely affected. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined that the issue affected the Barrier Integrity Cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012, Exhibit 3, the inspectors determined that since this finding involved the reactor coolant system boundary, a detailed risk evaluation was required. The Senior Reactor Analyst reviewed the finding and determined that a detailed risk evaluation was not required. The licensee performed an engineering evaluation and concluded that there was no impact to the reactor vessel. As a result, the Senior Reactor Analyst concluded that there was no change in risk due to the performance deficiency. The inspectors determined that since the procedural steps to perform Attachments VIII and X concurrently had been in place since 1994, this was a latent issue; therefore no cross-cutting aspect was assigned.
05000416/FIN-2013005-022013Q4Grand GulfFailure to Provide Adequate Procedures Results in Loss of Safety FunctionThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to provide an adequate procedure for a safety related activity. On December 17, 2013, while performing Surveillance Procedure 06-IC-1E31-Q-1016-02, RCIC Steam Supply Pressure Low Functional Test, Revision 111, the reactor core isolation cooling (RCIC) system became inoperable due to the procedure being incorrectly revised. Furthermore, the procedure error resulted in the containment isolation capability for RCIC being lost for approximately 1 hour. As an immediate corrective action, the licensee restored the breakers regaining isolation capability, and reopened the RCIC inboard isolation valve, thus restoring RCIC to operable status. The licensee entered this issue into the corrective action process under Condition Reports CR-GGN-2013-07720, CR-GGN-2013-07733, and CR-GGN-2013-07374. The failure to have an adequate procedure for the reactor core isolation cooling steam supply pressure low functional test is a performance deficiency. The performance deficiency was more than minor and therefore a finding because it was associated with the procedure quality 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. This performance deficiency was also associated with the procedural quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined the issue affected the Barrier Integrity Cornerstone. The inspectors used Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, May 6, 2004, and determined the finding was a type B finding at full power. Using Table 6.1, Phase 1 Screening-Type B Findings at Power, the inspectors concluded that since this issue involved containment isolation valves in a BWR Mark III containment, a Phase 2 analysis was necessary. Using Table 6.2, Phase 2 Risk Significance Type B Findings at Full Power, the inspectors concluded that the risk significance was very low (Green) because the exposure time was less than 3 days. Furthermore, the inspectors determined that this issue affected the Mitigating System Cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power , June 19, 2012, Exhibit 2, the inspectors determined that since the finding represented a loss of system and/or function, a detailed risk evaluation was required. The inspectors utilized the Grand Gulf Standardized Plant Analysis Risk model to determine the change in core damage frequency (CDF) due to the loss of safety function. The inspectors assigned the RCIC system a failure probability of 1.00 for a conservative duration of 1 hour. The resulting change in CDF was 1.9E-9/year, thus the finding was of very low safety significance (Green). The Senior Risk Analyst reviewed the inspectors evaluation and verified the conclusions to be correct. The apparent cause of this finding was that the licensee failed to effectively utilize human error prevention techniques. Therefore, the finding had a crosscutting aspect in the area of human performance, work practices because the licensee did not perform adequate self and peer checking while performing an activity affecting quality.
05000416/FIN-2013005-032013Q4Grand GulfEntry Into A High Radiation Area Without A Required Radiation Monitoring DeviceThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.7.1, resulting from an individual entering a high radiation area without the required radiation monitoring device. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2012-04112. As a corrective action, the radiation protection manager coached the individual on the need for proper dosimetry devices in high radiation areas. The entry into a high radiation area without all required radiation monitoring devices was a performance deficiency and was a violation of Technical Specification 5.7.1. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because it removed a barrier intended to prevent the worker from receiving unexpected dose. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation had a cross-cutting aspect in the human performance area, associated with the work practices component, because the worker and crew members did not use human error prevention techniques, such as self and peer checking.
05000416/FIN-2013005-042013Q4Grand GulfFailure To Survey Resulting in Personnel Entry To A High Radiation AreaThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR 20.1501(a) for failure to survey, which resulted in a worker entering an unposted high radiation area. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-08436 and CR-GGN-2012-09225. As corrective actions, the licensee coached radiation protection personnel on exhibiting a questioning attitude, walked down all affected areas; verified correct postings were used, and surveyed for any other unanticipated dose rate alarms. The failure to survey and determine radiation levels was a performance deficiency. The significance of the performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because the failure exposed a pipefitter to higher than anticipated radiation dose rates. The inspectors used Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, to determine the significance of the violation. The violation had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation had a cross-cutting aspect in the human performance area, associated with the work control component, because licensee personnel failed to appropriately plan a work activity by not incorporating risk insights, job site conditions, including environmental conditions, which may impact human system interface and radiological safety, and the need for planned contingencies or compensatory actions, such as surveying and up-posting affected areas after a power ascension.
05000416/FIN-2013004-012013Q3Grand GulfFailure to Follow Procedure Results in Inadequate Operability DeterminationThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, the inspectors identified that the licensee failed to establish an adequate basis for operability when a degraded or nonconforming condition had been identified. On August 30, 2013, Condition Report CR-GGN-2013-05604 was initiated to document a step change in the standby service water (SSW) siphon line K factor, which is a measure of flow through the siphon line. The K factor could have increased due to air entrapment in the siphon line that resulted from using air to mix the basin water following chemical treatments. The inspectors challenged the validity of the evaluation because the second step change in K factor, from 48 to 64, represented new information that had not been evaluated in the previous condition report. As an immediate corrective action, the licensee re-performed the operability determination and provided an adequate basis of operability by evaluating the system with the additional K factor data. Furthermore, the licensee verified the siphon line did not have any obstructions by observing the SSW basin levels equalize as water flowed through the siphon line. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-05687. The failure to perform an operability determination in accordance with procedure was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the reliability, availability and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. The inspectors determined that the apparent cause of this finding was that the licensee had identified and used previously completed operability evaluations without verifying that the previously completed evaluations were fully applicable to the identified conditions. Therefore, the finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality.
05000416/FIN-2013004-022013Q3Grand GulfFailure to Obtain NRC Approval for a Change in Method of Evaluation for Determining Reactor Vessel FluenceThe team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, involving the licensees failure to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a new method of evaluation for determining reactor vessel neutron fluence. On November 4, 2003, the NRC issued Amendment Number 160 to the Facility Operating License of the Grand Gulf Nuclear Station. The amendment revised the Updated Final Safety Analysis Report (UFSAR) to change the Reactor Vessel Material Surveillance Program to reflect participation in the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Additionally, the amendment revised the UFSAR to state that neutron fluence calculations performed after 2002 will be in accordance a methodology that has been approved by the NRC staff and is consistent with the attributes identified in NRC Regulatory Guide 1.190, Calculation and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The licensee developed a new neutron fluence calculation method which was based on a neutron fluence calculation method that had been previously approved by the NRC for another facility, which was documented in Nine Mile Point Nuclear Station, Unit No. 1 Issuance of Amendment RE: Pressure-Temperature Limit Curves and Tables, dated October 27, 2003. The NRC identified that the calculation, which was developed for GGNS, used the CASMO-4/SIMULATE code package to calculate the neutron source, whereas the prior calculation performed for Nine Mile Point Nuclear Station (NMP) used the ORIGEN code to calculate the neutron source. The inspectors determined that, although these codes are intended for the same purpose, they are distinct codes and the NRC approved only the use of one neutron source code (i.e., ORIGEN) in the neutron fluence calculation method of evaluation at Nine Mile Point. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-04743. The licensees failure to determine that a change to their method of evaluation for calculating reactor vessel neutron fluence was a departure from a method of evaluation approved by the NRC and required NRC review and approval prior to implementation was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding had the ability to impact the regulatory process. The performance deficiency was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation. In accordance with the NRC Enforcement Manual, risk insights from Inspection Manual Chapter 0609, Significance Determination Process, are used in determining the significance of 10 CFR 50.59 violations. Using the Inspection Manual Chapter 0612, Appendix B, Issue Screening, the team determined the finding adversely affected the Barrier Integrity Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the team determined the finding required a detailed risk evaluation because the finding involved the reactor coolant system boundary. A Senior Reactor Analyst performed the evaluation and determined the finding had very low safety significance (i.e., Green) because the NRC performed calculations and did not determine that the licensees Pressure-Temperature limits had or would have expired or been invalid; therefore, the change in risk was negligible. Since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. The finding does not have a cross-cutting aspect because cross-cutting aspects are not assigned to traditional enforcement violations.
05000416/FIN-2013004-032013Q3Grand GulfFailure to Review Temporary Modifications by Operations Personnel During TurnoverThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure 02-S-01-4, Shift Relief and Turnover, Revision 42. Specifically, the licensee failed to ensure proper turnover of the status of temporary modifications installed in the plant was being conducted by operations staff during turnover. The inspectors determined that the operations staff was required by Attachment III of that procedure to review the TMs log prior to taking the shift. The inspectors interviewed the operations staff and asked if the TMs were reviewed prior to taking shift that day. The staff member stated he had not and when asked about Attachment III of the turnover procedure, he was not familiar with that attachment of the procedure. The inspectors interviewed additional operations staff members about the review of temporary modification status during turnover, and they also indicated they had not reviewed temporary modification during turnover. As a corrective action, the licensee added copies of Attachment III of the shift turnover procedure to the operations staff turnover book to ensure TMs were reviewed during shift turnover. The licensee entered this issue into the corrective action process under Condition Reports CR-GGN-2013-04481 and CR-GGN-2013-05955. The failure to review temporary modifications by operations personnel during turnover in accordance with station procedures 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 more significant safety concerns. Specifically, operators not reviewing the status of TMs installed in the plant during turnover could result in a loss of configuration control of plant equipment that could result in an improper response by operators to plant events. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Attachment 4, Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure, or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allowed outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. The inspectors determined the apparent cause of this finding was that licensee personnel were not using Attachment III of the operations turnover procedure. Therefore, the finding has a cross-cutting aspect in human performance area associated with work practices in that the licensee management did not provide proper oversight to ensure a proper turnover was being conducted by operations personnel.
05000416/FIN-2013004-042013Q3Grand GulfFailure to Maintain Design Control of the Power Supplies for the Emergency Switchgear and Battery Room Fire DampersThe inspectors reviewed a self-revealing Green non-cited violation of Facility Operating License Condition 2.C (41), Fire Protection Program, involving the failure to maintain design control of the power supplies for the emergency switchgear and battery room fire dampers. During a surveillance of the division 2 carbon dioxide Fire Damper Actuation System, ten division 1 switchgear and battery room cooler fire dampers were inadvertently closed. Electricians investigated and found that a common ground existed between the division 1 and 2 emergency switchgear and battery room damper control panels. The common ground was determined to originate from a factory installed ground strap connecting the negative terminal to the ground/neutral on the emergency switchgear and battery room damper control power supplies. The licensee reviewed plant drawings and determined that the ground strap on the power supplies should have been removed prior to installation due to this being designed as a non-grounded system. As an immediate corrective action, the licensee removed the factory installed ground straps and restored the system to operable status. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-03827. The failure to verify a new power supply was a like-for-like replacement of the original power supply to ensure the replacement power supply did not alter the design of the damper control system was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Attachment 4, Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had an adverse effect on the fixed fire suppression systems. The inspectors assigned a low degradation rating due to the fact that the automatic fire suppression systems performance and reliability was minimally impacted by the inspection finding. Since the finding was assigned a low degradation rating, it screened as being of very low safety significance (Green). The apparent cause of this finding was the procurement engineering evaluation did not verify the replacement power supplies met the design requirements to be compatible with the unique design of the emergency switchgear and battery room damper control system. Therefore, the finding had a cross-cutting aspect in the area of human performance, work practices component because the licensee failed to properly perform a procurement evaluation in accordance with station procedures.
05000416/FIN-2013004-052013Q3Grand GulfFailure to Implement the Offsite Dose Calculation ManualInspectors identified three examples of a non-cited violation of Technical Specification 5.5, Programs and Manuals, for failure to maintain and implement requirements of the offsite dose calculation manual (ODCM). Specifically, the licensee failed to: (1) adequately document and justify ODCM changes, (2) approve licensee initiated changes to the ODCM, and (3) implement the radiological effluent controls for liquid releases. The violation was entered into the licensees corrective action program as Condition Report CR-GGN-2013- 05039, and the licensee is evaluating the issue to determine the proper corrective action. Failure to implement the requirements of the offsite dose calculation manual is a performance deficiency. This performance deficiency is more than minor because it affected the Public Radiation Safety Cornerstone attribute of program and process because the failure to adequately justify and approve offsite dose calculation manual changes resulted in 49 liquid effluent releases, contrary to the licensees Offsite Dose Calculation Manual, Revision 37, requirements. Using Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined this to be a violation of very low safety significance (Green). The violation was in the effluent release program but was not a substantial failure to implement the effluent program, and the dose to the public did not exceed the 10 CFR Part 50 Appendix I criterion or 10 CFR 20.1301(e) limits. The violation had a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to ensure the individuals preparing and reviewing offsite dose calculation manual changes had sufficient knowledge of the effluent release control system, its components, and its function to adequately evaluate the impact of the change.
05000416/FIN-2013004-062013Q3Grand GulfFailure to Include Some Solid Radwaste Released in the 2012 Regulatory Guide 1.21 Annual Effluent ReportInspectors identified a non-cited violation of Technical Specification 5.6.3 because the licensee failed to include in the 2012 Annual Radiological Effluent Release Report some solid radioactive waste released to an offsite waste processor. The failure to include in the 2012 Annual Radiological Effluent Release Report all solid radioactive waste released to an offsite waste processor was a performance deficiency, contrary to Technical Specification 5.6.3. The violation was determined to be more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process and adversely affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation, in that some licensed radioactive material, which left the Grand Gulf Nuclear Station, was unaccounted for. Using Inspection Manual Chapter 0609, Appendix D, \"Public Radiation Safety Significance Determination Process,\" dated February 12, 2008, the inspectors determined the violation to be of very low safety significance because, although it was a radioactive material control issue, it was not a transportation issue, and it did not result in public dose greater than 0.005 rem. The violation had a cross-cutting aspect in the human performance area, work control component because the licensee did not appropriately coordinate work activities by incorporating actions to address the need for work groups to communicate and coordinate with each other during activities in which interdepartmental coordination was necessary to assure human performance.
05000416/FIN-2013004-072013Q3Grand GulfFailure to Follow Alarm Response Steps to Restore the TSE Following MaintenanceThe inspectors reviewed a Green self-revealing finding for the failure to follow Procedure 04-1-02-1H13-P680-9A, TSE INFL OFF, Revision 36; in that operations personnel did not verify steps were followed per this alarm response procedure prior to returning the turbine thermal stress evaluator (TSE) to service following maintenance activities. The failure to follow alarm response procedure then resulted in an automatic reactor scram on July 30, 2013. Site personnel determined that the scram was caused by high reactor pressure resulting from the turbine unloading beyond the capability of the bypass valves after restoring the TSE to service following maintenance. On July 26, 2013, the control room received an alarm TSE-STU CAB FAIL. The licensee failed to determine the correct cause of the alarm due to inadequate troubleshooting. Therefore, when the maintenance was completed and the TSE was returned to service, the turbine started to unload resulting in a reactor scram due to reactor vessel high pressure. The immediate corrective actions included determining the cause of the scram and taking actions to restore equipment prior to plant startup. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-04943. The failure to follow alarm response steps to restore the TSE following maintenance is a performance deficiency. Specifically, Procedure 04-1-02-1H13-P680-9A, TSE INFL OFF, Revision 36, step 4.1 requires operational personnel to ensure that the TSE is functioning correctly following maintenance prior to restoring to service. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the issue has a very low safety significance (Green) because it only caused a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that the apparent cause of the finding was that the licensee did not troubleshoot to validate the cause for alarm TSE STU Cab Failure in accordance with station troubleshooting procedures. Therefore, the finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee did not use the troubleshooting process effectively.
05000416/FIN-2013004-082013Q3Grand GulfLicensee-Identified ViolationTitle 10 CFR 50.54(q) requires, in part, that licensees follow and maintain the effectiveness of an emergency plan that meets the requirements in the planning standards of 50.47(b). Title 10 CFR 50.47(b)(4) requires a standard emergency classification and action level scheme is in use by the nuclear facility licensee. Contrary to the above, the licensee failed to use the emergency classification and action level scheme to classify an event. Specifically, the licensee incorrectly used the emergency classification and action level scheme on May 12, 2013, by declaring a Notice of Unusual Event (NOUE) when an electrical transformer was thought to be on fire. The licensee, during a subsequent investigation, determined that the event was not a fire. This finding was more than minor because over classification potentially puts the public at risk and affected the Emergency Preparedness Cornerstone attribute of emergency response organization performance. The finding was evaluated by the Emergency Preparedness Significance Determination Process and determined to be of very low safety significance (Green) because it was a failure to comply with the NRC requirements and was not a loss of planning standard function. The planning standard function was not lost because the emergency classification and action level scheme basis has not changed. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-4156.
05000416/FIN-2013003-052013Q2Grand GulfFailure to Verify the Residual Heat Removal B System was Full of Water Within its Specified FrequencyThe inspectors identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.5.1.1 for the failure to verify the residual heat removal B system was full of water within its specified frequency. The inspectors reviewed a low pressure core injection subsystem B monthly functional test that was performed on April 10, 2013, per Procedure 06-OP-1E12-M-0002, LPCI/RHR Subsystem B Monthly Functional Test, Revision 113. The inspectors identified that the licensee failed to perform ultra sonic testing on the pipe prior to and after venting of the pipe directly upstream of the residual heat removal heat exchanger B outboard vent valve (1E12F074B). By not performing the ultra sonic testing, the licensee did not verify the residual heat removal system was full of water as required by Surveillance Requirement 3.5.1.1. Immediate corrective actions included performing the ultra sonic tests, which verified the system was full of water and satisfied the surveillance requirement. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-02847. The failure to verify the residual heat removal B system was full of water as required by Technical Specification Surveillance Requirement SR 3.5.1.1 is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstones objective of ensuring the availability, reliability and capability of systems that respond to prevent undesirable consequences. Specifically, the failure to perform the required ultra sonic testing resulted in Technical Specification Surveillance Requirement SR 3.5.1.1 not being met. Therefore, the licensee could not ensure the system would perform properly by injecting its full capacity into the reactor coolant system upon demand. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure, or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allowed outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. Through interviews with operations personnel, the inspectors determined the apparent cause of the finding was that management failed to ensure the ultra sonic test was performed. Therefore, the finding had a crosscutting aspect in the human performance area associated with the work practices component because the licensee failed to ensure supervisory and management oversight of work activities.
05000416/FIN-2013201-012013Q2Grand Gulf10 CFR 54.13 Apparent Violation for failure to provide complete and accurate information in response to RAIsThis letter refers to the evaluations by the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation (NRR) of responses dated May 25,2012, provided by Entergy Operations, Inc. (EOI), to requests for additional information (RAls) associated with the license renewal application for the Grand Gulf Nuclear Station. The responses pertained to requests by the NRC regarding the site\'s implementation of aging management activities for components included in two aging management programs. The NRC evaluation identified one apparent violation of NRC requirements that is being considered for escalated enforcement action in accordance with the NRC Enforcement Policy. The NRC evaluation determined that EOI apparently failed to provide complete and accurate information to the NRC in responses dated May 25, 2012, to RAls B.1.22-1, B.1.22-2, and B.1.41-3. The three RAls addressed several issues with the aging management activities for the Flow-Accelerated Corrosion program and the Service Water Integrity program. The license renewal application stated that both of these programs were consistent with the corresponding programs described in NUREG-1801, \"Generic Aging Lessons Learned (GALL) Report.\" However, during its reviews of operating experience associated with these programs, the staff found indications that aspects of these programs were inconsistent with corresponding programs in the GALL Report. Enforcement is not being considered for the statements in the initial license renewal application, but for the apparent incomplete or inaccurate information in the responses to RAls intended to evaluate potential inconsistencies with the programs in the GALL Report. The RAI responses were material because the NRC needed the requested information to verify that certain components would be adequately managed for erosion mechanisms so that the intended functions will be maintained consistent with the current licensing basis, as required by 10 CFR 54.21 (a)(3). Therefore, EOI appears to be in violation of 10 CFR 54.13.
05000445/FIN-2013007-112013Q2Comanche PeakFailure to Correct Design Calculations to Incorporate Technical Specification Allowed Frequency Range for the Emergency Diesel Generator in a Timely MannerThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Specifically, since May 2010, the licensee failed to correct a condition adverse to quality in a timely manner that involved updating design basis calculations for safety-related equipment to include the allowed technical specification frequency range of 2 percent for the emergency diesel generators. The finding was entered into the licensee\'s corrective action program as Condition Report CR-2013-006604. The inspectors determined that the failure to correct a condition adverse to quality in a timely manner that involved updating design basis calculations for safety-related equipment to include the allowed technical specification frequency range of 2 percent for the emergency diesel generators was a performance deficiency. The performance deficiency is more-than-minor because it was associated with the Reactor Safety, Mitigating Systems Cornerstone, Design Control attribute and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the calculations to support safety-related equipment did not include allowed technical specification frequency range for the emergency diesel generators to ensure the equipment would be capable of performing their safety-related functions. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was a deficiency affecting the design or qualification that did not result in the safety-related equipment losing operability or functionality. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to take appropriate corrective actions to address updating design basis calculations to include technical specification allowed emergency diesel generator frequency range in a timely manner, commensurate with their safety significance.
05000445/FIN-2013007-102013Q2Comanche PeakFailure to Incorporate the Refueling Water Storage Tank Vortexing Design Calculation into the Emergency Operating Procedures for Containment Spray Pump OperationThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, measures shall be establish to assure that the design basis for systems, structures, and components are correctly translated into specifications, drawings, procedures and instructions. Specifically, since 2006 and 2007, the licensee failed to appropriately incorporate the RWST vortexing design calculations 6 percent indicated level into the emergency operating procedures for switching containment spray pump suction from the RWST to the containment sump to prevent damage to the pumps. The finding was entered into the licensees corrective action program as Condition Report CR-2013-005739. The inspectors determined that the failure to appropriately incorporate the RWST vortexing design calculations 6 percent indicated level into the emergency operating procedures for switching containment spray pump suction from the RWST to the containment sump to prevent damage to the pumps was a performance deficiency. The performance deficiency is more-than-minor because it was associated with the Reactor Safety, Mitigating Systems Cornerstone, Procedure Quality attribute 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, Emergency Operating Procedure EOS-1.3A/B allowed the operators the ability to delay transfer of containment spray pump suction source which could have caused damage to the pumps due to vortexing. Using Inspection Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was not a design deficiency and did not result in the loss of operability or functionality. This finding did not have a cross-cutting aspect because the change to the procedure due to the addition of the sump strainers occurred in 2006 and 2007, and did not reflect current licensee performance.
05000445/FIN-2013007-092013Q2Comanche PeakFailure to Identify Fouling on the Emergency Diesel Generator Building Exhaust Ventilation ScreensThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, that states, in part, measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Specifically, prior to June 17, 2013, the licensee failed to establish an activity to identify fouling of the Unit 1 emergency diesel generator building exhaust ventilation screens. The finding was entered into the licensee\'s corrective action program as Condition Report CR-2013-006540. The inspectors determined that the failure to identify fouling on the Unit 1 emergency diesel generator building exhaust ventilation screens was a performance deficiency. The performance deficiency is more-than-minor because it had the potential to lead to a more significant safety concern. Specifically, the Unit 1 emergency diesel generator rooms could have insufficient exhaust flow to meet design basis temperature requirements if left uncorrected. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was not a design deficiency and did not result in the emergency diesel generators losing operability or functionality. This finding did not have a crosscutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
05000445/FIN-2013007-082013Q2Comanche PeakFailure to Provide Appropriate Acceptance Criteria for the Safety Chill Water PumpsThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, that states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, since 1994, the licensee failed to recognize that if the safety-related chilled water pumps were degraded to 90 percent of their reference value, as permitted by IST Procedures OPT-209A/B, the system may not be able to achieve the required design flowrates as stated in Calculation 1-EB-311-8. This finding was entered into the licensees corrective action program as Condition Report CR-2013-006252. The inspectors determined that the failure to ensure appropriate acceptance criteria were incorporated into test procedures for the safety chill water pumps was a performance deficiency. The performance deficiency is more-than-minor because it was associated with the Reactor Safety, Mitigating Systems Cornerstone, Design Control attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to recognize that if the safety-related chilled water pumps were degraded to 90 percent of their reference value, as permitted by IST Procedures OPT-209A/B, the system may not be able to achieve the required design flowrates as stated in Calculation 1-EB-311-8. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was not a design deficiency and did not result in the loss of operability or functionality. This finding did not have a cross-cutting aspect because Calculation 1-EB-311-8 was updated in 1994 to incorporate the uninterruptible power system fan coil units and did not reflect current licensee performance.
05000445/FIN-2013007-072013Q2Comanche PeakFailure to Provide Appropriate Acceptance Criteria and Testing Procedure InstructionsThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, 10 CFR Part 50, Appendix B, Criterion XI, Test Control, that states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, since 2001, the licensee failed to provide appropriate acceptance criteria and testing procedure instructions during modified performance tests involving Class 1E batteries for the 1-minute critical period testing data which incorporated the requirements of IEEE Standard 450-1995 to ensure the battery would meet the required design voltage for the duty cycle. The finding was entered into the licensees corrective action program as Condition Report CR-2013-005673. The inspectors determined that the failure to provide appropriate acceptance criteria and testing procedure instructions involving Class 1E batteries for the 1-minute critical period testing data during modified performance tests was a performance deficiency. The performance deficiency is more-than-minor because it was associated with the Reactor Safety, Mitigating Systems Cornerstone, Procedure Quality attribute 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, Procedure MSE-S0-5715 does not direct the technicians to record and evaluate the voltage at the end of the 1-minute critical period to ensure it does not drop below the designed minimum voltage, which would indicate the battery would not be capable of meeting the required design function. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was not a design deficiency and did not result in the loss of operability or functionality. This finding did not have a cross-cutting aspect because Calculation EE-CA-0000-5121 was implemented in 2001 and did not reflect current licensee performance.
05000445/FIN-2013007-062013Q2Comanche PeakFailure to Perform Adequate Operability AssessmentsThe inspectors identified a Green, non-cited violation, with three examples, of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that states, in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, for example 1 on February 28, 2013, for example 2 on June 5, 2013 and for example 3 on June 8, 2013, the licensee failed to follow procedure STI-442.01, Operability Determination and Functionality Assessment Program, Revision 1, Attachment 8.B page 3 of 5 which states, in part, Identify the topics that are applicable to the quick technical evaluation and include information for applicable topics within the evaluation such as: for example 1, The effect or potential effect of the degraded or nonconforming condition on the affected SSCs ability to perform its specified safety function, or for example 2, Compensatory Measures are recommended, or for example 3, Whether there is reasonable expectation of operability, including the basis for the determination. The finding was entered into the licensee\'s corrective action program as Condition Report CR-2013-006599. The inspectors determined that the failure to perform adequate operability assessments was a performance deficiency. The performance deficiency is more-than-minor because: Example 1: It was associated with the Reactor Safety, Barrier Integrity Cornerstone, Configuration Control attribute and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, shutting off of the containment spray pumps during a large break LOCA inside containment would allow containment pressure to increase. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, the inspectors determined the finding was of very low (Green) safety significance because it 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 or actual reduction in function of hydrogen igniters in the reactor containment. Example 2: It was associated with the Reactor Safety, Mitigating Systems Cornerstone, Equipment Performance attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the compensatory measures established in the first operability assessment did not ensure that offsite power would be maintained at minimum grid voltage. Example 3: It was associated with the Reactor Safety, Mitigating Systems Cornerstone, Design Control attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the operability assessment initially credited the use of the battery chargers after the emergency diesel generators restored power to the bus, without evaluating design basis for the battery chargers. For examples 2 and 3, the inspectors used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because these examples were a deficiency affecting the design or qualification that did not result in losing operability or functionality. This finding had a cross-cutting aspect in the area of human performance associated with the decision making component because the licensee failed in all three examples to conduct an effectiveness review of a safety-significant decision to verify the validity of the underlying assumptions to identify possible unintended consequences during the original operability assessments.
05000445/FIN-2013007-052013Q2Comanche PeakFailure to Analyze Effect of System Harmonics on Degraded Voltage RelaysThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that states, in part, measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to May 20, 2013, the licensee failed to assess the adverse effects of 6.9kV and 480V system harmonics on the degraded voltage relays. The finding was entered into the licensees corrective action program as Condition Report CR-2013-006230. The inspectors determined that the failure to analyze the effect of electrical system harmonics on the degraded voltage relays was a performance deficiency. The performance deficiency is more-than-minor because it was associated with the Reactor Safety, Mitigating Systems Cornerstone, Design Control attribute 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, failure to analyze the effect of electrical system harmonics on the degraded voltage relays could cause the relays to fail to actuate at the setpoint specified in Technical Specifications. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was a deficiency affecting the design or qualification that did not result in the safetyrelated equipment losing operability or functionality. This finding did not have a crosscutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
05000445/FIN-2013007-042013Q2Comanche PeakFailure to Update the FSAR for the APDGs in Accordance with Regulatory Guide 1.70-1995The inspectors identified a Severity level IV, non-cited violation of 10 CFR 50.71(e)(4), requires the UFSAR be updated, at intervals not exceeding 24 months, and states in part, the revisions must reflect all changes made in the facility or procedures described in the UFSAR. Specifically, prior to June 20, 2013, the inspectors identified the alternate power diesel generator system was not described in sufficient detail in the FSAR as required. This finding was entered into the licensees corrective action program as Condition Report CR-2013-006256. The inspectors determined that the failure to update the Final Safety Analysis Report to include the description of the APDG system in section 8.3.1 AC Power Systems was a performance deficiency. The issue is a performance deficiency because it was a failure to meet requirement, 10 CFR 50.71(e)(4), and it was within the licensees ability to correct the problem. Using Inspection Manual Chapter 0612, Appendix B, the performance deficiency was assessed through both the Reactor Oversight Process and traditional enforcement because the finding had the potential for impacting the NRCs ability to perform its regulatory function. The finding resulted in a minor performance deficiency. For traditional enforcement, the inspectors used the Enforcement Policy, in accordance with Section 6.1.d.3, and determined the violation to be a Severity Level IV, non-cited violation, because the licensee failed to update the UFSAR as required by 10 CFR 50.71(e)(4), but the lack of up-to-date information had not resulted in any unacceptable change to the facility or procedures. This violation did not have a cross-cutting aspect.
05000445/FIN-2013007-032013Q2Comanche PeakFailure To Establish 10 CFR 50.65(a)(1) Performance Goals for the APDGsThe inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(1), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, that states, in part, that the licensee shall monitor the performance or condition of structures, systems, or components, against licensee established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. Specifically, on July 26, 2012, the licensee failed to establish goals and monitor the performance of the alternate power diesel generator system to ensure the system is capable of providing the necessary electric power onto the emergency buses. The finding was entered into the licensees corrective action program as Condition Report CR-2013-006521. The inspectors determined that the failure to follow procedure to establish performance goals while performing Maintenance Rule (a)(1) monitoring to ensure the APDG system is capable and tested to meet the design basis requirements, was a performance deficiency. The performance deficiency is more-than-minor because it was associated with the Reactor Safety, Mitigating Systems Cornerstone, Equipment Performance attribute and adversely affected the cornerstone objective to ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure directs the licensee to establish performance goals on activities that address conditions which were determined to be classified as (a)(1). In accordance with Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Mitigating System Cornerstone. Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined the finding was of very low (Green) safety significance because the finding was not a design deficiency and did not result in the loss of operability or functionality. This finding had a cross-cutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure that emergency equipment is adequate and available to assure nuclear safety.
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