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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000483/FIN-2018003-01 | 2018Q3 | NRC identified | Failure to perform 10 CFR 50.59 evaluation for compensatory measures associated with stagnant, inactive loop | The inspectors identified an unresolved item related to implementation of 10 CFR 50.59, Evaluations Changes, Tests and Experiments, for the licensees failure to perform an adequate evaluation for compensatory measures for a stagnant, inactive loop. The inspectors identified an unresolved item related to implementation of 10 CFR 50.59, Evaluations Changes, Tests and Experiments, for the licensees failure to perform an adequate evaluation for compensatory measures for a stagnant, inactive loop. The licensee enacted compensatory measures to support atmospheric dump valve/turbine-driven AFW pump operability due to an issue identified for natural circulation cooldown with a faulted steam generator (i.e., inactive loop). A reduction in the Technical Specification 3.4.16 dose equivalent iodine (DEI) limit (from 1Ci/gm to 0.4Ci/gm) was imposed without a 10 CFR 50.59 evaluation and/or license amendment. Specifically, the licensee did not consider the compensatory measure of reducing Technical Specification 3.4.16 limits on DEI-131 as a change to technical specifications.The licensee considered this a temporary action that did not meet the intent of 10 CFR 50.90 for a technical specification change. | ||
| 05000483/FIN-2018003-02 | 2018Q3 | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 6 of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for combating emergencies and other significant events. The licensee established Emergency Operating Procedure (EOP) ES-0.2, Natural Circulation Cooldown, Revision 9, in part, to meet the regulatory requirement. Figure 1 of ES-0.2 allowed cooldown rates that exceeded the values used in the license basis for radiological consequence analyses and exceeded the values used in the design of the nitrogen accumulators for atmospheric steam dumps and turbine-driven auxiliary feedwater system actuations. This issue was discussed in Licensee Event Report 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown Contrary to the above, from April 29, 2008 through May 7, 2018, the licensee failed to maintain procedures for combating emergencies and other significant events. Specifically, the licensee failed to maintain EOPs for natural circulation cooldown. This performance deficiency resulted in atmospheric steam dumps and turbine-driven auxiliary feedwater systems being rendered inoperable due to depletion of the safety-related actuation nitrogen. | |
| 05000483/FIN-2018002-05 | 2018Q2 | Severity level Minor | NRC identified | Minor Violation | Contrary to Technical Specification 3.6.3, Containment Isolation Valves, the licensee failed to maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4. Specifically, the licensee failed to shut the reactor building service air header supply outer containment isolation valve KAV0118 after the fall 2017 refueling outage. As a result, isolation valve KAV0118 was left open from November 25, 2017, through January 11, 2018, which rendered the valves containment isolation function inoperable. The as-found testing demonstrated that the overall containment isolation function, for that penetration, was met with inner containment isolation valve KAV0039 in the normally shut position. Additional information can be found in Licensee Event Report 05000483/2018-001-00, Violation of 20 Technical Specification 3.6.3, Containment Isolation Manual Valve Found in Open Position (ADAMS Accession Number ML18071A208). The licensees failure to comply with Technical Specification 3.6.3, Containment Isolation Valves, and maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4 was a performance deficiency. Screening: The inspectors determined the performance deficiency was minor because it was not a precursor to a significant event, did not have the potential to lead to a more significant safety concern, did not relate to a performance indicator that would have exceeded a threshold and did not adversely impact any of the cornerstone objectives. Specifically, the as-found local leak rate testing demonstrated that containment isolation function was met with inner containment isolation valve KAV0039 in the normally shut position. Enforcement: The failure to comply with Technical Specification 3.6.3, Containment Isolation Valves, and maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. | |
| 05000483/FIN-2018002-04 | 2018Q2 | Green | H.14 | NRC identified | Failure of an Analysis of the Impact of Changes to Emergency Action Levels to Demonstrate the Changes Did Not Reduce the Effectiveness of the Emergency Plan | The inspectors identified a non-cited violation of 10 CFR 50.54(q)(3) for the failure of an analysis of the impact of changes to licensee emergency action levels to demonstrate that the changes did not reduce the effectiveness of the emergency plan. |
| 05000483/FIN-2018002-03 | 2018Q2 | Green | P.6 | NRC identified | Failure to Critique an Inaccurate Emergency Classification During a Simulator Training Scenario | The inspectors identified a non-cited violation of 10 CFR 50.47(b)(14) for the licensees failure to critique an inaccurate emergency classification made during licensed operator training. |
| 05000483/FIN-2018002-02 | 2018Q2 | Green | H.4 | NRC identified | Failure to Establish Maintenance Procedures for Doors that Provide Safety-Related Functions | The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to establish, implement, and maintain procedures associated with door maintenance. Specifically, the licensee failed to establish, implement, and maintain maintenance procedures for doors that provide safety-related functions such as ventilation pressure boundaries. As a result, 15 safety-related doors were identified that either had degraded conditions or that did not have a periodic maintenance task to inspect the doors. |
| 05000483/FIN-2018002-01 | 2018Q2 | Green | H.12 | NRC identified | Failure to Adequately Assess and Manage Risk Associated with Switchyard Work During a Planned Risk Significant Turbine-Driven Auxiliary Feedwater Pump Equipment Outage | The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at Nuclear Power Plants, for the licensees failure to adequately assess and manage risk associated with switchyard work during a planned risk significant turbine-driven auxiliary feedwater pump equipment outage. Specifically, the licensee failed to properly classify switchyard work and manage the risk as required by Procedures APA-ZZ-00322, Appendix F, Online Work Integrated Risk Management, Revision 16, and ODP-ZZ-00002, Appendix 2, Risk Management Actions for Planned Risk Significant Activities, Revision 13. |
| 05000483/FIN-2018001-01 | 2018Q1 | Green | NRC identified | Failure to Maintain Emergency Operating Procedures | The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, "Procedures," for the licensee's failure to maintain emergency operating procedures for aligning auxiliary feedwater suction sources. Specifically, the licensee added continuous action steps to emergency operating procedures that placed both motor-driven auxiliary feedwater pumps in pull-to-lock and isolated their associated recirculation lines after depleting the two non-safety-related suction sources. These actions cause two of the three safety-related auxiliary feedwater pumps to be rendered inoperable prior to aligning the safety-related suction source of essential service water which is credited in accident analysis. | |
| 05000483/FIN-2017003-01 | 2017Q3 | Green | P.3 | NRC identified | Spurious Containment Spray Pump Start | The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to implement Preventative Maintenance Basis document IC-LSELS, Load Shed and Emergency Load Sequencer (LSELS), Revision 0. Specifically, the licensee failed to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23, a Consolidated Controls 6N232 relay driver card, within the scheduled periodicity. On June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours, of which all 44 hours w ere unplanned. As immediate corrective actions, the licensee replaced the circuit card under Job 17002747, completed post -maintenance testing, and restored the system to operable status on June 30, 2017. The licensee entered this issue into the corrective action program under Condition Report 20170 3433. The failure to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 within the scheduled periodicity was a performance deficiency. This performance deficiency was more than minor , and therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, o n June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours . Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At - Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) 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. Specifically, the total duration of inoperability was 44 hours which is less 3 than the technical specification allowed completion time of 72 hours for this system. The 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 did not replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 prior to failure although this issue was documented in corrective actions ranging from April 2008 to January 2017 (P.3). |
| 05000483/FIN-2017007-04 | 2017Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 8 of Regulatory Guide 1.33, Revision 2, Appendix A, Procedures for Control of Measuring and Test Equipment and for Surveillance Tests, Procedures, and Calibrations, Part b, requires, in part, that specific procedures for surveillance tests, inspections, and calibrations, should be written (implementing procedures are required for each surveillance test, inspection, or calibration, listed in the technical specifications). Station Procedure EDP-ZZ-01114, Motor Operated Valve Program Guide, Revision 034, Section 3.6.3.b, requires, in part, that the motor-operated valve engineer document a signature analysis report within 60 days following a diagnostic test of motor operated valves. Contrary to the above, on July 17, 2016, the motor-operated valve engineer failed to generate a signature analysis report within 60 days following a recent diagnostic test of a motor-operated valve. Specifically, in May 2014, the NRC inspection team identified NCV 05000483/2014007-06, Failure to Review Motor Operated Valve (MOV) Data and Complete Analysis of the Data in a Timely Manner. This finding was entered into the licensee's corrective action program as Callaway Action Requests CARs 201402987 and 201402992. During Refueling Outage RF21 (spring of 2016), 33 motor operated valves had been tested and should have had a signature analysis report completed by the end of June 2016. On July 17, 2016, the licensee personnel recognized that they had not completed the signature analysis report for 31 of the 33 valves tested. The team evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012. The team concluded the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered no. The licensee entered this issue into their corrective action program as Condition Report CR-201606143. | |
| 05000483/FIN-2017007-03 | 2017Q3 | Green | NRC identified | Inputs to Internal Flooding Calculations Not Translated into Procedures or Instructions | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which requires, in part, that measures shall be established to assure that the design basis is correctly translated into procedures and instructions. Specifically, prior to on August 4, 2017, the licensee had design calculations that assumed operator actions to mitigate internal flooding of certain areas within specified time durations. These time requirements for the design basis flooding calculations had not been translated into any procedures or instructions. In response to this issue, the licensee performed a preliminary evaluation and determined that operator actions to support the design calculation could be performed within the time required. The licensee has entered this issue into their corrective action program as Condition Report CR-201703981. The team determined that the failure to translate operator time requirements for mitigating design basis flooding of critical areas into procedures or instructions was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to confirm that design basis inputs had been translated into procedures or instructions. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, 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; and did not result in the loss of one or more trains of nontechnical specification equipment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. | |
| 05000483/FIN-2017007-02 | 2017Q3 | Green | NRC identified | Safety Injection Piggyback Valve EJ-HV-8804A Valve Interlocks Not Tested | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which requires, in part, that 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 procedures. Specifically, prior to August 3, 2017, the licensee failed to have a program to completely test the interlock circuit for safety injection pump and recirculation suction isolation valves, EJ-HV-8804A and EJ-HV-8804B. When the licensee personnel performed a review the interlock circuits for the valves, they identified that there had been gaps in the testing. In response to this issue, the licensee investigated all of the testing activities associated with the valve interlock circuits and identified that in 2010, a comprehensive test of the circuits had been performed as the result of a modification. The licensee has entered this issue into their corrective action program as Condition Report CR-201703962. The team determined that the failure to develop and implement testing programs for verifying that the circuits for the multiple interlocks associated with safety injection valve EJ-HV-8804A would perform as designed was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to establish a testing program to verify that the valve interlock circuits for valve EJ-HV-8804A were being tested. A failure of the interlocks and an operator error could result in an inadvertent release path to the environment. 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 non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. | |
| 05000483/FIN-2017007-01 | 2017Q3 | Green | H.5 | NRC identified | Not Verifying the Operation and Timing of the Engineered Safety Feature Transformer XNB01 Load Tap Changer | The team identified a Green, cited violation of Technical Specification 5.4.1.a which requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance, requires, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with documented instructions appropriate to the circumstances. Specifically, from May 2014 through August 4, 2017, as a result of ineffective corrective action of Callaway Action Requests CAR-201402827 and CAR-201405312, the licensee failed to performed preventative maintenance procedures to verify the operation and timing of the engineered safety feature transformer XNB01 load tap changer. This violation was previously identified by the NRC and documented as NCV 05000483/2014007-01. In accordance with Section 2.3.2.a of the NRC Enforcement Policy, this finding is being cited because the licensee failed to restore compliance within a reasonable amount of time after the violation was initially identified. This finding was entered into the licensees corrective action program as Condition Report CR-201703992, VIO 05000458/2017007-01, Not Verifying the Operation and Timing of the Engineered Safety Feature Transformer XNB01 Load Tap Changer. The team determined that the failure to implement maintenance procedures to periodically verify transformer XNB01 load tap changer operation and time testing was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failures to perform, periodic verification of the operation and time testing of the load tap changer could result in adverse operation of the load tap changer during a design basis event. If the load tap changer did not operate correctly, the safety-related buses may not have adequate voltage to reset the degraded voltage relay, thus spuriously disconnecting from the offsite power source. In accordance with Inspection Manual Chapter 0609, Appendix A, 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 non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The finding had a cross-cutting aspect in the area of human performance, work management, because the licensee failed to plan, control, and execute work activities such that nuclear safety is the overriding priority. Specifically, the licensee did not plan and execute the testing of the transformer XNB01 load tap changer in a timely manner (H.5). |
| 05000483/FIN-2017002-02 | 2017Q2 | Severity level IV | NRC identified | Failure to Analyze the Effect of Changes to Maintaining the Gaitronics System | Severity Level IV. The inspectors identified a Severity Level IV non- cited violation for the licensees failure to perform an analysis of a change to processes supporting the emergency preparedness program that demonstrated the change did not reduce the effectiveness of the emergency plan in accordance with the requirements of 10 CFR 50.54(q)(3). There were no immediate safety concerns associated with this violation because less than 10 percent of the public address speakers were determined to be degraded or non- functional. This issue has been placed in the licensees corrective action system as Condition Report 201702343. The failure to perform an analysis of the effect of changes in processes supporting emergency preparedness is a performance deficiency within the licensees ability to foresee and correct. The finding was more than minor because the finding was associated with the Facilities and Equipment Cornerstone attribute and adversely affected the Emergency Preparedness Cornerstone objective. The finding was assessed using traditional enforcement because the licensees failure to perform a required analysis impacted the regulatory process . The finding was evaluated using the NRCs Enforcement Policy, dated November 1, 2016, Section 6.6(d) , and was determined to be a Severity Level IV violation because the violation did not affect radiological assessment or offsite notification. Traditional enforcement violations are not assessed for cross -cutting aspects. | |
| 05000483/FIN-2017403-01 | 2017Q2 | Severity level IV | NRC identified | Security | ||
| 05000483/FIN-2017002-01 | 2017Q2 | Green | H.11 | Self-revealing | Failure to Follow Motor Control Center Procedure | Green . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Procedure MPE-ZZ-QS001, Cleaning and Inspection of Motor Control Centers, Revision 34. On May 2, 2017, the licensee failed to ensure contactors operated freely per step 7.6.8 during reassembly of motor control center NG08F for the essential service water cooling tower by pass valve EFHV0066. As a result, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. As immediate corrective actions, the licensee replaced the starter assembly under Job 17001973, completed testing including electrically cycling valve EFHV0066, and restored the system to operable status on May 4, 2017. The licensee entered this issue into the corrective action program under Condition Report 201702418. The failure to follow Procedure MPE-ZZ-QS001 was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it adversely affected the configuration control 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, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) 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. Specifically, the total duration of 3 inoperability was approximately 57 hours which is less than the allowed completion time of 72 hours for this system. The finding had a cross-cutting aspect in the area of human performance associated with challenge the unknown because the licensee failed to stop when faced with uncertain conditions. Specifically, the maintenance technician encountered resistance when manually operating the contactors, signed off the step as complete, and later rationalized the decision with the supervisor aft er completing the work (H.11 ). |
| 05000483/FIN-2017403-02 | 2017Q2 | Green | H.11 | Licensee-identified | Security | |
| 05000483/FIN-2017403-03 | 2017Q2 | Green | Licensee-identified | Security | ||
| 05000483/FIN-2017403-04 | 2017Q2 | Green | NRC identified | Security | ||
| 05000483/FIN-2017403-05 | 2017Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000483/FIN-2017403-06 | 2017Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000483/FIN-2017404-01 | 2017Q2 | Green | H.1 | NRC identified | Security | |
| 05000483/FIN-2017404-02 | 2017Q2 | H.13 | NRC identified | Security | ||
| 05000483/FIN-2017001-01 | 2017Q1 | Severity level Enforcement Discretion | NRC identified | Enforcement Action EA-17-050, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliances | Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, Criterion 2, Design Bases for Protection Against Natural Phenomena, states, in part, that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes. Criterion 4, Environmental and Dynamic Effects Design Basis, states, in part, that structures, systems and components important to safety shall be appropriately protected against dynamic effects including missiles which may result from events and conditions outside the nuclear power unit. As part of their response to Regulatory Issue Summary 2015-06, Tornado Missile Protection, the licensee performed a review of protection against tornado-generated missiles required by the current licensing basis. During the review, on March 13, 2017, the licensee identified a portion of the diesel generator fuel oil system that could be susceptible to tornado missiles. The licensee identified a low-probability scenario where one or more tornado-generated missiles could impact the emergency fuel oil truck connection lines on the south wall of the diesel generator building. The two non-safety-related connection lines are each connected to the safety-related normal fuel oil transfer lines via a tee connection and a normally closed isolation valve. Direct impact by a tornado-generated missile parallel to either trains connection line could impart a load on the tee connection to the normal fuel oil line that had not been evaluated. Failure of the tee connection would result in the associated diesel generator being incapable of performing its safety function. Relevant Enforcement Discretion Policy On June 10, 2015, the NRC issued Enforcement Guidance Memorandum (EGM) 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance (Adams Accession Number ML15111A269). The EGM referenced a bounding generic risk analysis performed by the NRC staff that concluded that tornado missile vulnerabilities pose a low risk significance to operating nuclear plants. Because of this, the EGM described the conditions under which the NRC staff may exercise enforcement discretion for noncompliance with the current licensing basis for tornado-generated missile protection. Specifically, if the licensee could not meet the technical specification required actions within the required completion time, the EGM allows the staff to exercise enforcement discretion provided the licensee implements initial compensatory measures prior to the expiration of the time allowed by the limiting condition for operation. The compensatory actions should provide additional protection such that the likelihood of tornado missile effects are lessened. The EGM then requires the licensee to implement more comprehensive compensatory measures within 60 days of issue discovery. The compensatory measures must remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. In addition, the issue must be entered into the licensees corrective action program. Because EGM 15-002 listed Callaway as a Group A plant, enforcement discretion will expire on June 10, 2018. Because the EGM did not provide for enforcement discretion for any related underlying technical violations; any associated underlying technical violations will be assessed through the enforcement process. Licensee Actions The licensee declared both diesel generators inoperable, complied with the applicable technical specification action statements, initiated a condition report, invoked the enforcement discretion guidance, implemented prompt compensatory measures, and returned the systems, structures, and components to an operable status. The licensee instituted compensatory measures intended to reduce the likelihood of tornado missile effects. These included verifying that guidance was in place for severe weather procedures, abnormal and emergency operating procedures, and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX), that training on these procedures was current, and that a heightened level of awareness of the vulnerability was established. As an additional compensatory measure, the licensee placed concrete blocks adjacent to the piping penetrations to provide a greater level of protection from tornado generated missiles. NRC Actions The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance in EGM 15-002. The inspectors also evaluated whether the measures as implemented would function as intended and were properly controlled. The inspectors verified through inspection that the EGM 15-002 criteria were met in each case. Therefore, the staff determined that it was appropriate to exercise enforcement discretion and not take enforcement action for the required actions of Technical Specification 3.8.1, AC Sources Operating, provided the non-compliances are resolved by June 10, 2018 (EA-17-050). The inspectors did not fully review the underlying circumstances that resulted in the technical specification violations. As stated in EGM 15-002, violations of other requirements which may have contributed to the technical specification violations will be evaluated independently of EGM implementation. The inspectors will verify restoration of compliance and assess the underlying circumstances during future inspection activities. | |
| 05000483/FIN-2016004-01 | 2016Q4 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, Section 9.a, requires, in part, that maintenance should be properly pre-planned and performed in accordance with documented instructions appropriate to the circumstances. Contrary to the above, on October 19, 2016, the licensee failed to properly pre-plan and perform a post-maintenance test in accordance with documented instructions appropriate to the circumstances. Specifically, the post-maintenance test for work performed on valve EFHV0066, the essential service water to ultimate heat sink cooling tower train B bypass valve, did not include a seat leak test, which would be necessary for the work performed. As a result, on November 17, 2016, operators discovered this valve leaking by at approximately 3900 gallons per minute. The licensee subsequently determined that the safety function of the ultimate heat sink would not be adversely affected with leakage up to 4100 gallons per minute. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012. The inspectors concluded the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered no. The licensee entered this issue into their corrective action program as Condition Report 201608791. | |
| 05000483/FIN-2016002-01 | 2016Q2 | Green | H.14 | NRC identified | Failure to Account for Water Hammer Stresses in Essential Service Water System Calculations | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for the essential service water pipe stresses caused by pressure fluctuations of the known column closure water hammer phenomenon. The licensee failed to properly account for essential service water piping membrane stress and impact loads as required by the 1974 ASME Code, Section III, paragraphs ND-3112.4 and ND-3111. Specifically, the licensees design calculations for the essential service water system did not account for the pressure fluctuations caused by a known column closure water hammer phenomenon that occurs during a loss of off-site power or load sequencer testing. The licensee completed a prompt operability determination assuring the system was operable under the current conditions and was completing engineering evaluations of the data collected to demonstrate the operability of the system under design conditions. The licensee entered this issued into the corrective action program as Callaway Action Requests 201603472 and 201603819. The inspectors determined that the licensees failure to account for the pressure fluctuations caused by a known column closure water hammer phenomenon in the design calculations for the essential service water system 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 and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, when the licensee recognized that the column separation water hammer phenomenon was occurring in the essential service water system, they only applied the forces to the containment coolers, not the entire system (H.14). |
| 05000483/FIN-2016002-02 | 2016Q2 | Green | H.14 | NRC identified | Failure to Meet Applicable ASME Code Requirements for Repairs to Components in the Essential Service Water System | The inspectors identified a non-cited violation of 10 CFR 50.55a, Codes and Standards, for the licensees failure to repair various ASME Code Class 3 components in accordance with ASME Code, Section XI requirements. Specifically, the licensee did not follow the applicable ASME Code requirements when making repairs to various components in the ASME Code Class 3 essential service water system. The licensee reasonably determined the essential service water system remained operable, and completed the necessary repairs and testing to restore compliance with ASME Code. The licensee entered this issue into their corrective action program as Callaway Action Requests 201603640 and 201604282. The inspectors determined that the programmatic failure to repair various ASME Code Class 3 components in the essential service water system in accordance with ASME Code 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 and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. Specifically, the licensee performed a historical system health review and reasonably determined the essential service water system remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant system leaks, and the appropriate repairs and testing were completed on the affected components. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee failed to ensure training of the personnel was adequate to recognize that the repair of the leaks constituted repairs in accordance with ASME Code, Section XI and thus failed to include the necessary ASME testing requirements in the work performance packages to ensure adequate performance of an activity which affected testing of a safety-related modification/repair to risk-significant systems, and thereby ensure nuclear safety (H.9). |
| 05000483/FIN-2016002-03 | 2016Q2 | Green | H.14 | NRC identified | Failure to Adequately Evaluate Operability for a Degraded Condition | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an adequate operability assessment when a degraded or nonconforming condition was identified. Specifically, after the licensee identified that a severe water hammer transient would occur following a loss of off-site power, the licensee generated an operability evaluation that relied on judgement and inaccurate information which failed to establish a reasonable expectation of operability. Following questions from inspectors the licensee determined that this judgement was not correct and performed a new evaluation to ensure operability of the essential service water system. The licensee entered this issue into their corrective action program as Callaway Action Request 201605488. The licensees failure to properly assess and document the basis for operability when a severe water hammer occurred in the essential service water system was 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 cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, severe water hammer transients in the essential service water system due to a loss of off-site power, result in a condition where structures, systems, and components necessary to mitigate the effects of accidents may not have functioned as required. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensees use of unsupported judgement and incorrect data resulted in an evaluation that failed to demonstrate a reasonable expectation of operability (H.14). |
| 05000483/FIN-2016002-04 | 2016Q2 | Green | H.1 | NRC identified | Failure to Promptly Correct Conditions Adverse to Quality | During an NRC inspection conducted June 6-30, 2016, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, from November 2010 through June 2016, the licensee failed to promptly correct a condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues which were previously identified by the NRC as non-cited violation 05000483/2010006-01. The failure to resolve these issues resulted in subsequent safety-related equipment failures. This violation is associated with a Green Significance Determination Process finding The inspectors identified a cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to take timely corrective action for a previously identified condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues that were previously identified by the NRC as non-cited violation 05000483/2010006-01 and the failure to resolve these issues resulted in subsequent safety-related equipment failures. The licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into their corrective action program as Callaway Action Request 201604440. The licensees failure to take timely and adequate corrective actions to correct a condition adverse to quality was 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 cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct water hammer and corrosion issue resulted in the licensee declaring safety-related room coolers and chillers inoperable until an analysis of system operability was completed. This affected their capability to respond to initiating events to prevent undesirable consequences Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of resources in the human performance area because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, by failing to address water hammer and corrosion issues, station management failed to ensure that the essential service water system was available and adequately maintained to respond during a loss of off-site power event (H.1). |
| 05000483/FIN-2016002-05 | 2016Q2 | Green | H.9 | Self-revealing | Failure to Follow Plant Foreign Material Exclusion Procedure | The inspectors reviewed a self-revealed finding for the licensees failure to follow the plant procedure for foreign material exclusion. Specifically, after finding foreign material (broken cable ties) within the main generator excitation transformer, established as a foreign material exclusion Level 2 area, the licensee failed to determine the reason for the foreign material and enter the issue into the corrective action program for resolution as required by Procedure APA-ZZ-00801, Foreign Material Exclusion, Revision 32. The licensees failure to follow the plant procedure for foreign material exclusion was 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 Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, after identifying several broken cable ties on the floor inside a foreign material exclusion Level 2 area the licensee did not determine the reason for the foreign material nor enter the condition into the corrective action program as required by Procedure APA-ZZ-00801. Because the licensee failed to understand what caused the cable tie degradation, a subsequent cable tie failure resulted in a plant trip. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the finding was determined to be of very low safety significance because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, several groups within the licensees organization were unaware the excitation transformer cabinet was classified as a foreign material exclusion Level 2 area nor the requirements if foreign material is found within the foreign material exclusion area (H.9). |
| 05000483/FIN-2016403-01 | 2016Q2 | Green | NRC identified | Security | ||
| 05000483/FIN-2016403-02 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000483/FIN-2016001-02 | 2016Q1 | Green | H.4 | NRC identified | Inadequate Operability Evaluation for Degraded Flood Mitigation Capability in Piping Penetration Room | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an adequate operability determination for safety related components located in the 1988 foot auxiliary building train B piping penetration room (room 1203) based on degraded internal flooding drain capability. Specifically, the immediate operability determination included incorrect assumptions that were not verified to support the operability determination as required by Procedure ODP-ZZ-00001, Addendum 15, Operability and Functionality Determinations, Revision 8. The immediate corrective action was to implement a compensatory measure to support operability of the equipment in room 1203. The issue was placed in the corrective action program as Callaway Action Request 201601412. The licensees failure to verify assumptions used in the immediate operability determination and ensure a sound basis for operability exists per plant procedures was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is similar to examples 3.j and 3.k in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, and if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, failure to perform adequate operability evaluations by verifying assumptions and ensuring a sound basis for operability exists may result in the failure to enter the appropriate limiting conditions of operation for technical specification equipment. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding involved the degradation of equipment specifically designed to mitigate a flooding initiating event, therefore, Exhibit 4, External Events Screening Questions, was used to complete the screening. The finding was determined to need a detailed risk evaluation because if the equipment (i.e., floor drain lines) is assumed to be completely failed or unavailable, it would degrade one or more trains of a system that supports a risk significant system or function. In consultation with the Senior Reactor Analyst, the finding was determined to be of very low safety significance because, based on the actual condition of the drains and the extent of the clogging in room 1203, an evaluation by the licensee showed that the maximum internal flooding water level in the room would not challenge the operability of any equipment needed for safe shutdown or to mitigate an accident. This finding has a team work cross-cutting aspect in the human performance cross-cutting area because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, inadequate communication between engineering and operations personnel led to the belief that a passageway existed between rooms 1203 and 1204 when it did not (H.4). |
| 05000483/FIN-2016001-01 | 2016Q1 | NRC identified | Possible Incorrect Screening of the Spent Fuel Pool Decay Heat Removal Key Safety Function | The inspectors identified an unresolved item associated with the National Fire Protection Association (NFPA) Standard 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants, non-power operations assessment. Specifically, the inspectors developed an issue of concern in that the licensee screened the potential loss of spent fuel pool cooling from further consideration for any fire event based on adequate procedural guidance and time when the procedures would not maintain the fuel in a safe and stable condition. On January 13, 2014, the licensee transitioned their fire protection program to a risk-informed, performance-based program based on NFPA Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. Paragraph 1.3.1 of NFPA Standard 805 requires licensees to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. Paragraph 1.5.1 of NFPA Standard 805 lists five nuclear safety performance criteria. These criteria provide requirements to demonstrate that fire protection features are capable of providing reasonable assurance that the plant is not placed in an unrecoverable condition in the event of a fire. For the decay heat removal nuclear safety performance criterion, the standard requires that decay heat removal shall be capable of removing sufficient heat from the reactor core or spent fuel such that fuel is maintained in a safe and stable condition. Paragraph 1.6.56 of NFPA Standard 805 defines safe and stable conditions: For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling. The licensee described how they satisfied the nuclear safety performance criteria in Calculation KC-26, Nuclear Safety Capability Assessment, Revision 1. The Nuclear Safety Capability Assessment applied to both power and non-power operations. For non-power operations, the licensee evaluated the spent fuel pool decay heat removal key safety function and determined that the spent fuel pool decay heat removal key safety function did not require a detailed review since adequate time was available, and procedural guidance was provided, for operators to respond to and mitigate a loss of spent fuel pool decay heat removal, even under full hot core offload conditions. The licensee stated that the shortest time to boil, under worst case conditions for a normal plant shutdown, was two hours. In addition, the licensee stated that all of the analyses to address a loss of spent fuel pool decay heat removal utilized a success criterion of no boiling. The licensee implemented the process outlined in Frequently Asked Question (FAQ) 07-0040, Non-Power Operations Clarifications, Revision 4, for the non-power operations assessment. This FAQ stated that licensees should conservatively assume the entire contents of a fire area are lost and document the loss of success paths. This FAQ also stated that licensees should specifically identify those areas (pinch points) that cause the loss of all success paths for a key safety function. The inspectors noted that the licensee did not perform these actions for the spent fuel pool decay heat removal key safety function because this key safety function was screened out from further consideration. If the licensee had evaluated the spent fuel pool decay heat removal key safety function using the process outlined in this FAQ, then the licensee would have assumed that both trains of spent fuel pool cooling are lost during a fire in the fuel handling building because both trains are located within the same fire area and were unprotected. This FAQ also stated that fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling), thereby eliminating a pinch point. However, the licensee stated that no fire modeling was used to eliminate the identification of pinch point fire areas as part of the non-power operations assessment performed using the process in FAQ 07-0040. In the event that a fire in the fuel handling building disabled both trains of spent fuel pool cooling, operators were expected to enter Procedure OTO-EC-00002, Spent Fuel Pool High Temperature, Revision 9, due to the increasing temperature of the spent fuel pool. This procedure provided directions for operators to restore one or both trains of spent fuel pool cooling. Since both trains of spent fuel pool cooling were assumed lost due to the fire, the operators would be unable to restore spent fuel pool cooling using this procedure. After a period of time, the spent fuel pool would begin boiling and the level would begin lowering. At this time, operators were expected to enter Procedure OTO-EC-00001, Loss of SPF/Refuel Pool Level, Revision 13. Procedure OTO-EC-00001 directed the operators to open two normally locked essential service water valves to restore and maintain spent fuel pool level. The licensees procedures allowed the spent fuel pool to reach boiling conditions prior to restoring and maintaining level. Since NFPA Standard 805 defined safe and stable conditions, in part, as fuel coolant temperature below boiling, the procedures did not maintain the fuel in a safe and stable condition. The inspectors identified an issue of concern in that the licensee screened the potential loss of spent fuel pool cooling from further consideration for any fire event based on adequate procedural guidance and time when the procedures would not maintain the fuel in a safe and stable condition. The inspectors determined that additional information is required to determine if a performance deficiency exists. Specifically, the inspectors need to determine if this scenario should have been addressed as part of the current FAQ 07-0040 guidance, or if new guidance is needed to address this type of scenario where the full core has been offloaded to the spent fuel pool. On March 31, 2016, additional guidance was requested from the Office of Nuclear Reactor Regulation via a request to review and update FAQ 07-0040. This memorandum is documented in ADAMS as Accession Number ML16091A152. The licensee entered this issue of concern into the corrective action program as Callaway Action Request 201600726. This issue of concern is being treated as Unresolved Item 05000483/2016001-01, Possible Incorrect Screening of the Spent Fuel Pool Decay Heat Removal Key Safety Function. | ||
| 05000483/FIN-2015009-04 | 2015Q4 | Green | H.14 | NRC identified | Two Examples of a Failure to Properly Designate the Significance Level of Callaway Action Requests. | The team identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to implement their corrective action program procedure. Specifically: (1) on November 20, 2014, the licensee designated the improper setting of the auxiliary feedwater flow control valve ALHV005 limit switches as Significance Level 5 (administrative close) instead of Significance Level 3 (lower tier cause evaluation) and (2) on December 9, 2014, the licensee downgraded the failure of the Modutronics card for valve ALHV0005 from Significance Level 1 (root cause analysis) to Significance Level 3 based on unverified assumptions of the failure mechanisms. Following failure of the Modutronics card for valve ALHV0005, the licensee assumed that the early failure was due to a manufacturing defect (infant mortality) without supporting data to prove this designation. The licensee entered these issues into the corrective action program as Callaway Action Requests 201506921 and 201507235. The two failures to properly designate the Significance Level of Callaway action requests constitute a performance deficiency. This finding was more than minor, and therefore, a finding because it adversely affected the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failures to properly designate the significance of the conditions precluded determining the appropriate cause determinations and extent of conditions and resulted in failure to correct the conditions before they further manifested themselves following a trip. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance, because it did not affect system design, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This finding has a human performance cross-cutting aspect in the area of conservative bias in that the decision-making did not demonstrate a conservative/prudent choice in designating the significance level of the Callaway action requests based on two cases of unverified/incorrect information (H.14). |
| 05000483/FIN-2015009-06 | 2015Q4 | Green | H.11 | Self-revealing | Failure to Have an Adequate Procedure for Testing the Torque and Thrust Values for the Auxiliary Feedwater Pump Flow Control Valves | The team reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to provide a procedure appropriate to the circumstances. Specifically, on March 4, 2014, the licensee performed Job 08505547, and had not correctly accounted for the differential pressure the valve would actually experience, and had incorrectly set and tested the close torque switch on valve ALHV0005. As a result, On November 15, 2015, during steam generator filling operations, Valve ALHV0005 failed to move in the closed direction when the torque switch opened. The incorrect close torque switch setting prevented the valve from going full closed. In response to this issue, the licensee, using Job 14005755, repaired the valve, and confirmed that the close torque switch settings were correct and successfully retested. This finding was entered into the licensees corrective action program as Callaway Action Report 201508399. The failure to establish a procedure that included a suitable instructions to set the torque switch on a motor-driven AFW valve after maintenance or testing was a performance deficiency. This finding was more than minor, and therefore, a finding because it adversely affected the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a post-maintenance testing program for the motor-driven auxiliary feedwater valve torque and thrust settings caused valve ALHV0005 not to close completely, causing the operators to take action and shut down motor-driven feedwater pump B. In accordance with 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 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 non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding had a cross-cutting aspect in the area of human performance, challenge the unknown, because the licensee did not stop and challenge that the tested differential pressure across valve ALHV0005 was significantly different than the other valves (H.11). |
| 05000483/FIN-2015009-07 | 2015Q4 | Green | H.12 | NRC identified | Failure to Identify and Correct Additional Undersized Components on Auxiliary Feedwater System Flow Control Valve Modified Modutronics Controller Cards | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality. Specifically, as of September 23, 2015, the licensee had not taken corrective action, following previous identification of undersized field current rectifier bridges, to ensure that an independent review of the modified circuit design had been completed, or that the modified cards had been subjected to a sufficient testing and qualification program. Thus, following questioning by the team, the licensee identified additional components (two other rectifier bridges) on the newly modified circuit cards that were also potentially undersized. The licensee performed an operability evaluation and concluded that the new cards were operable, based on additional circuit analysis that was performed. This issue was entered into the corrective action program as Callaway Action Request 201506874. The failure to identify and correct a condition adverse to quality was a performance deficiency. This performance deficiency is more than minor, and therefore, a finding because it adversely affected the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct design deficiencies associated with these circuit cards could have resulted in the inoperability of auxiliary feedwater control valves and their inability to operate on demand. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance, because it did not affect system design, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This finding has a human performance cross-cutting aspect in the area of Avoid Complacency, because the licensee did not thoroughly evaluate the issue to ensure that the resolutions address causes and extent of conditions. Specifically, the licensee had identified that the Modutronics cards failed because of improper design of the field current rectifier bridge, but did not plan for the possibility for other latent issues to determine if other components on the cards were adequately sized for their application (H.12). |
| 05000483/FIN-2015009-05 | 2015Q4 | Green | H.13 | NRC identified | Failure to Determine the Cause and Take Corrective Action to Preclude Repetition for the Inadequate Design of Auxiliary Feedwater Flow Control Valve Modutronics Cards. | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to determine the cause and take corrective action to preclude repetition for a significant condition adverse to quality. Specifically, on May 21, 2015, the licensee received new information that refuted the previously assumed failure mechanism for AFW flow control valve ALHV0005 documented in December 2014, but failed to initiate a new Callaway action request to document the new information and report it to appropriate levels of management. As a result, the licensee failed to identify the failure of the valve as a significant condition adverse to quality, determine the cause, initiate a prompt operability assessment, and identify corrective action to preclude repetition until valve ALHV0007 failed, for the same reason, following a reactor trip on August 11, 2015. The licensee entered this issue into the corrective action program as Callaway action request 201506846. The failure to determine the cause and take corrective action to preclude repetition for a significant condition adverse to quality when failure analysis indicated that a significant defect existed on valves ALHV0005 and ALHV0007 was a performance deficiency. This finding was more than minor, and therefore, a finding, because it adversely affected the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure of the licensee to determine the cause and take corrective action to preclude repetition for a significant condition adverse to quality when new information on the failure mechanism was received precluded determining the root cause and extent of condition and the performance of an operability determination, which resulted in failure to correct the condition before it further manifested itself following a reactor trip. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance, because it did not affect system design, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This finding has a human performance cross-cutting aspect in the area of consistent process in that the individuals that received the information concerning the failure mechanism of the Modutronics cards failed to use a systematic approach to documenting the information and communicating it to appropriate levels of management (H.13). |
| 05000483/FIN-2015009-01 | 2015Q4 | Green | H.4 | NRC identified | Failure to Verify the Suitability of the Design of the Reverse-Engineered Replacement Controller Cards for the Auxiliary Feedwater Flow Control Valves | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that the design of the replacement reverse-engineered Modutronics controller cards for the auxiliary feedwater control valves were suitable for their application. Specifically, as of August 11, 2015, the licensee failed to establish suitable interface requirements in procurement documents to Nuclear Logistics Incorporated (the vendor) and verify the adequacy of the design by either design reviews or testing. Specifically, the team identified that neither the licensee nor the vendor had performed a design review sufficient to assure that the Modutronics controller cards were suitable for their application. In addition, the licensee had not provided the vendor with sufficient information to reverse-engineer the controller cards. Lastly, neither the licensee nor the vendor performed testing sufficient to verify the adequacy of the design of the new Modutronics controller cards. As a result, the replacement cards were supplied with motor field current rectifier bridges that were undersized and marginal for their application, such that two of them failed in service, rendering these auxiliary feedwater system valves inoperable. Following performance of a root cause analysis, the licensee replaced the deficient controller cards with those of a higher current rating. The licensee initiated Callaway Action Request 201505796 to place this item into the corrective action program. The failure to ensure that the design of the replacement for the Modutronics cards was suitable for their application was a performance deficiency. This performance deficiency is more than minor, and therefore, a finding because it adversely affected the Mitigating Systems Cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, design deficiencies associated with these circuit cards resulted in the inoperability of auxiliary feedwater control valves and their ability to operate on demand. The team performed an initial screening of the finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the team determined that the finding required a detail risk evaluation because it represented the potential loss of one train of safety-related equipment (auxiliary feedwater) for greater than the technical specification allowed outage time. A Region IV senior reactor analyst performed a detailed risk evaluation in accordance with Appendix A, Section 6.0, Detailed Risk Evaluation, which determined that the finding was of very low safety significance (Green). The analyst determined that the importance of the failure of valves ALHV0005 and ALHV0007 was based on the postulated failure time of the turbine-driven auxiliary feedwater pump because this determined the position in which the valves failed. The internal events incremental conditional core damage probability was 8.17 x 10-7. The analyst also determined that the finding had only a minimal effect on external initiator risk and that the finding would not involve a significant increase in the risk of a large, early release of radiation. This finding has a human performance cross-cutting aspect in the area of teamwork, because individuals in different work groups did not appropriately communicate across organizational boundaries. Specifically, licensee personnel did not adequately communicate the design and testing requirements for the reverse engineered cards (H.4). |
| 05000483/FIN-2015004-01 | 2015Q4 | Green | H.5 | Self-revealing | Failure to Promptly Correct a Condition Adverse to Quality on the Reactor Coolant System | The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality. Specifically, after identifying boric acid deposits on the flange downstream of valve BBV0400, a reactor coolant system boundary valve, the licensee did not promptly take action to stop the reactor coolant system leakage before it worsened and caused a plant shutdown due to reactor coolant system leakage in excess of technical specification limits. The immediate corrective action was to torque the valve and flange to reduce leakage to within limits. The licensee entered this issue into their corrective action program as Callaway Action Request 201505308. The licensees failure to correct the condition adverse to quality (i.e. leakage past valve BBV0400) in a timely manner was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the reactor coolant system equipment and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to correct the reactor coolant system leakage through valve BBV0400 resulted in reactor coolant system leakage worsening and exceeding technical specification limits, and a plant shutdown. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance because after a reasonable assessment of degradation, it could not: 1) result in exceeding the reactor coolant system leak rate for a small loss of coolant accident, or 2) have likely affected other systems used to mitigate a loss of coolant accident resulting in a total loss of their function. This finding has a cross-cutting aspect in the work management component of the human performance cross-cutting area because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee initially planned to address the reactor coolant leakage six months after the issue was identified, and then moved it out an additional three months, failing to prioritize the work commensurate with its safety significance. |
| 05000483/FIN-2015004-02 | 2015Q4 | Green | Self-revealing | Failure to Properly Establish and Maintain a Plant Procedure for Preparation for Refueling Outages | The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1, Procedures, for the licensees failure to establish, implement, and maintain a procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, Procedure ODP-ZZ-00022, Operations Preparation, Performance, and Restoration from Refueling Outages, did not provide adequate guidance to ensure a blind flange located on the reactor coolant system was properly reinstalled resulting in reactor coolant system leakage into containment. The immediate corrective action taken by the licensee was to replace the gasket with a Flexitallic gasket and torque the flange. Additionally, the licensee implemented repetitive maintenance tasks in their work management program to identify flanges removed during an outage and to torque them properly upon reinstallation. The licensee entered this issue into their corrective action program as Callaway Action Request 201505702. The licensees failure to properly establish and maintain Procedure ODP-ZZ-00022 was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, Procedure ODP-ZZ-00022, did not provide adequate guidance to ensure the blind flange located downstream of valve BBV0400 was properly reinstalled. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance because after a reasonable assessment of degradation, it could not: 1) result in exceeding the reactor coolant system leak rate for a small loss of coolant accident, or 2) have likely affected other systems used to mitigate a loss of coolant accident resulting in a total loss of their function. This finding does not have a cross-cutting aspect because the performance deficiency is not representative of current licensee performance, in that the inadequate instructions were added to the procedure in 2003. | |
| 05000483/FIN-2015004-03 | 2015Q4 | Green | H.4 | Self-revealing | Failure to Follow Plant Procedure for Unit Reliability Team | The inspectors reviewed a self-revealing finding for the licensees failure to follow plant procedures for the unit reliability team. Specifically, after delaying a modification to the plants turbine control system, no compensatory measures were implemented to minimize or prevent failure of the system due to aging of the system beyond its evaluated service life as required by plant Procedure APA-ZZ-00549, Appendix E, Unit Reliability Team Operations. The licensees failure to follow the plant procedure for the unit reliability team was 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 Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, as no compensatory measures were implemented after the digital upgrade to the turbine control system was deferred from the spring 2013 refueling outage to the spring 2016 refueling outage, the turbine control system malfunctioned causing a runback of the turbine and downpower transient on the plant. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the teamwork component of the human performance cross-cutting area because the licensee did not ensure that individuals and work groups communicate across organizational boundaries to ensure nuclear safety is maintained. Specifically, the outage leadership team identified the need for the compensatory measures, but did not communicate the priority nor the effect on nuclear safety to site leadership to gain the resources needed to implement these measures. |
| 05000483/FIN-2015009-02 | 2015Q4 | Green | Self-revealing | Failure to Have an Adequate Procedure for Calibration of the Auxiliary Feedwater Pump Flow Control Valve Potentiometer | The team reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to prescribe activities affecting quality using procedures appropriate to the circumstances. Specifically, on November 18, 2009, the licensee revised Procedure MTE-ZZ-QA033, MOVATS UDS (motor operated valve actuator test system universal diagnostic system) Testing of Torque Controlled Modutronics Limitorque Motor Operated Rising Stem Valves, Revision 3, to incorporate a second method of valve testing, and introduced an error in bypassing a test of the Modutronics board setup feedback potentiometer. As a result, on July 23, 2015, the actuator misinterpreted the actual position of the valve, which subsequently failed to open when operators attempted to open the valve following a forced reactor shutdown. In response to this issue, the licensee has reviewed all maintenance and test activities that could affect the potentiometer and has revised the appropriate procedures. This finding was entered into the licensees corrective action program as Callaway Action Request 201505332. The failure to provide a procedure appropriate to the circumstances for an auxiliary feedwater system flow control valve was a performance deficiency. This finding was more than minor, and therefore, a finding because it adversely affected the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to provide a procedure appropriate to the circumstances to set up an auxiliary feedwater system flow control valve feedback potentiometer resulted in its inability to operate manually on demand. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance, because it did not affect system design, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. The valve would have automatically throttled auxiliary feedwater flow to approximately 300 gpm on demand. This finding did not have a cross-cutting aspect because the procedure revision resulting in the inadequate procedure was issued in 2009, and previous opportunities to correct the procedure occurred in 2010. Thus, this performance deficiency was not indicative of current licensee performance | |
| 05000483/FIN-2015009-03 | 2015Q4 | Green | Self-revealing | Failure to Have an Adequate Post-Maintenance Test for Setting the Motor-Driven Auxiliary Feedwater Flow Control Valve Modutronics Potentiometer | The team reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for failure to ensure that testing demonstrated that structures, systems, and components will perform satisfactorily in service. Specifically, on October 24, 2014, the licensee failed to establish a suitable post-maintenance test program to demonstrate that the motor-driven auxiliary feedwater flow control valve Modutronics potentiometer had been set correctly after maintenance. The testing consisted of stroking the valve full open or full closed, and did not consider step changes in valve positioning and did not confirm the potentiometer feedback settings during valve positions that were not full open or full closed. In response to this issue, the licensee performed another calibration of the potentiometer, focusing on the potentiometer position during the valve stroke. The new post-maintenance test included opening the valve in discreet step changes to test the valve position feedback potentiometer. This finding was entered into the licensees corrective action program as Callaway Action Request 201505332. The failure to establish a suitable post-maintenance test program to demonstrate that the motor-driven auxiliary feedwater flow control valve Modutronics potentiometer would be set correctly after maintenance or testing was a performance deficiency. This finding was more than minor, and therefore, a finding because it adversely affected the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a post-maintenance testing program for the motor-driven auxiliary feedwater valve Modutronics potentiometer to verify that the potentiometer was set correctly, resulted in valve ALHV0011 failing to open when operators initiated a signal to place the valve in an open position. In accordance with 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 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 non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. | |
| 05000483/FIN-2015003-04 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation | Title 10 of the Code of Federal Regulations 55.46(c), Plant-Referenced Simulators, requires, in part, that a plant-referenced simulator must demonstrate expected plant response to operator input and to transient and accident conditions to which the simulators have been designed to respond. Contrary to the above, on December 12, 2013, and March 23, 2015, the simulator failed to demonstrate expected plant response to operator input and to transient and accident conditions to which the simulator has been designed to respond. Specifically, during simulator post-event testing on those dates, the simulator did not correspond in direction of change of all monitored plant parameters and, in one case, the letdown portion of the chemical and volume control system automatically isolated when this did not occur in the reference plant. The violation was of very low safety significance because it dealt with identified simulator modeling deficiencies that did not negatively impact operator performance in the actual plant during reportable events. The licensee entered this issue into their corrective action program as Callaway Action Report 201504406. | |
| 05000483/FIN-2015404-01 | 2015Q3 | Green | NRC identified | Security | ||
| 05000483/FIN-2015003-03 | 2015Q3 | Green | H.14 | NRC identified | Unauthorized Non-Routine Maintenance on a Sealed Source Device | The inspectors identified a non-cited violation of Callaway Plants License No. NPF-25, Condition 2.B.(3), for the licensee performing non-routine maintenance on a J.L. Shepherd calibrator without license authorization. The licensee documented this issue in their corrective action program as Corrective Action Request 201505175. Their immediate corrective action was to secure the calibration source and review their procedural requirements. Performing non-routine maintenance on a J.L. Shepherd calibrator without a license authorization is a performance deficiency. This finding is more than minor because the performance deficiency adversely affects the Occupational Radiation Safety Cornerstone, in that, if the licensee performs non-routine maintenance on radiologically risk significant sources without being specifically authorized or trained on how to perform the non-routine maintenance, an uncontrolled high radiation area could result. 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 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 a conservative bias cross-cutting aspect in the area of human performance, because individuals did not use decision making practices that emphasized prudent choices over those that were simply allowable, or ensure a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, licensee staff assumed that they could perform any type of maintenance on the calibrator without verifying that their license authorized those activities (H.14). |
| 05000483/FIN-2015003-02 | 2015Q3 | Green | H.11 | NRC identified | Failure to Follow Operability Determination Procedure | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow their operability determination procedure. Specifically, when an auxiliary feedwater control valve failed to operate from the main control room, the licensee failed to evaluate the operability of the component in accordance with Procedure ODP-ZZ-00001, Addendum 15, Operability and Functionality Determinations. The immediate corrective action taken by the licensee was to evaluate the operability of the flow control valve. After determining that the equipment was inoperable, the licensee entered the required technical specification condition and performed the required technical specification actions. The licensee entered this issue into their corrective action program as Callaway Action Request 201502708. This performance deficiency is more than minor and, therefore, a finding, because, if left uncorrected, it has the potential to lead to a more significant safety concern if safety-related systems are not properly evaluated for operability. The finding affects the Mitigating System Cornerstone because the performance deficiency is related to the auxiliary feedwater systems ability to conduct short-term decay heat removal. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance because it did not affect system design, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This finding has a cross-cutting aspect of challenge the unknown in the human performance cross-cutting area because the licensee did not stop when faced with uncertain conditions. Specifically, rather than declaring the system inoperable and allowing the process to evaluate the condition, the licensee declared the system operable without fully understanding the failure mechanism (H.11). |
| 05000483/FIN-2015003-01 | 2015Q3 | Green | NRC identified | Failure to Conduct Simulator Testing and Maintenance In Accordance with ANSI/ANS-3.5-2009 | The inspectors identified a finding with four examples for failing to conduct and evaluate simulator performance testing in accordance with the standards of ANSI/ANS-3.5-2009. Specifically, the licensee failed to do the following: set the instantaneous main turbine load reduction to 50 percent as supported by design basis data in the 2014 performance of Transient (11), Maximum Design Load Rejection include the evaluation of parameter pressurizer temperature in the 30 percent, 50 percent, and 80 percent power Steady-State Performance Test as specified in accordance with the standard, Appendix B, Section B.3.1 include the evaluation of parameter secondary heat balance data in the 30 percent, 50 percent, and 80 percent power Steady-State Performance Test as specified in accordance with the standard, Appendix B, Section B.3.1 replicate the dynamic functioning of annunciators on the simulator panels used during normal, abnormal, off-normal, and emergency evolutions, or to identify and correct noticeable differences in accordance with the standard, Sections 4.2.1.2 and 4.2.1.4 The licensee initiated corrective action documented in Callaway Action Requests 201504760, 201504759, 201504418, and 201504355. The licensees failure to conduct and evaluate performance testing in accordance with the ANSI/ANS-3.5-2009 standard as endorsed by Regulatory Guide 1.149, Revision 4, was the performance deficiency. The performance deficiency is more than minor because it adversely impacted the human 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. Additionally, if left uncorrected, the performance deficiency could have become more significant in that not correcting noticeable differences between the simulator and the reference plant can both leave the potential for negative training of licensed operators and call into question the ability to conduct valid licensing examinations with the simulator. Using Manual Chapter 0609, Significance Determination Process, Attachment 4, Tables 1, 2, and 3 worksheets; and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), Flowchart Block #14, the finding was determined to have very low safety significance (Green) because it dealt with deficiencies associated with simulator testing, modification, and maintenance and there was no evidence that the plant-referenced simulator does not demonstrate the expected plant response or have uncorrected modeling and hardware deficiencies related to the examples. The examples supporting this finding involved actions taken with the simulator testing and maintenance program before the present performance period. Therefore, no cross-cutting aspect is assigned to the finding. | |
| 05000483/FIN-2015002-03 | 2015Q2 | Severity level Enforcement Discretion | NRC identified | Inverter NN11 Inadvertently Transferred to its Alternate AC Source | On June 9, 2014, the Callaway Plant was in Mode 1 operating at 100 percent rated thermal power when, during a maintenance activity, inverter NN11 unexpectedly transferred from its normal direct current (dc) source to its bypass alternating current (ac) source. This inverter provides power to the NN01 bus which is one of four vital 120 Vac instrument buses at the Callaway Plant. The transfer of inverter NN11 to its bypass source was caused by a momentary loss of power to bus SB038 which supports instrumentation and controls for systems such as the reactor trip system and the engineered safety feature actuation system. This momentary loss of power caused the following plant impacts: Control rod insertion 612 steps, with an associated pressurizer level and pressur perturbation and subsequent Xenon transien Opening of valve BNLCV0112D, centrifugal charging pump A suction fro refueling water storage tank isolation valve, due to momentary loss of th associated volume control tank level channe Actuation of the steam generator environmental allowance modifier circuit resulting in resetting of the low level setpoint trip from 17 to 21 percent narro range leve Numerous momentary partial trip actuation The NRC inspectors responded to the control room and verified that the plant system responded as designed and that the operators stabilized the plant in accordance wit plant procedures Investigation identified a loose mounting screw that secures disconnect switch NN01-11 to NN01. Maintenance work in the area of the loose termination led to a momentary interruption of power to cabinet SB038, which appeared as a fault condition to the inverter, thus causing the inverter to transfer to its alternate power source. The cabinet, bus, and inverter are seismically qualified and are required to be capable of performing their design basis accident functions following a safe shutdown earthquake. With the degraded electrical termination, which existed for an extended period of time before discovery and repair, the inverter and SB038 loads would not have been capable of performing their design basis function following a safe shutdown earthquake, thus rendering the components inoperable. The direct cause of this event was inadequate thread engagement of the screw securing disconnect switch NN01-11 to the NN01 bus. However, the presence of threads in the switch mounting hole (which is not intended to engage with the bus bar termination screw) introduced the potential for binding during screw installation. The detail of this mounting configuration is not identified on plant drawings of the cabinet or switch provided by the vendor and nothing in the work control process required a detailed comparison of the switch to the work procedures and, as such, it was reasonable that this potential vulnerability was not identified and addressed in the procedure or pre-job walkdown. During the actual installation of the screw, the screw appeared flush and tight with the switch mounting board, meeting the requirements of the work package. The equipment was successfully post-maintenance tested and technical specification surveillance tested for a period of 6 years. There was also no industry or vendor operating experience describing this vulnerability. Based on this information, the inspectors concluded that no performance deficiency existed since it was not reasonable for Callaway Plant personnel to foresee and correct this condition. The licensees root cause analysis determined that the root cause of the event was that work instructions did not include direction to remove the back panel cover of the cabinet to support alignment and thread engagement of the mounting screws during switch installation. Corrective actions taken by the licensee included changes to job planning aids and the maintenance procedures associated with the replacement of this type of switch. The inspectors determined during their review of Licensee Event Report 2014-003-01 that traditional enforcement applies in accordance with Inspection Manual Chapter 0612, Appendix B, Figures 1 and 2, Issue Screening, Inspection Manual Chapter 0612, Section 9, and NRC Enforcement Policy, Section 2.2.4.d, because a violation of NRC requirements existed without an associated Reactor Oversight Process performance deficiency. This issue is considered to be a Severity Level IV violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, based on a conservative bounding evaluation performed using Callaways SPAR model which determined the condition was of very low safety significance (Green) and was similar in significance to NRC Enforcement Policy example 6.1.d.2. This issue was entered into Callaway Plants corrective action program as Callaway Action Request 201403898. Licensee Event Report 2014-003-01 was submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Callaway Technical Specification 3.8.7, Inverters Operating, based on the period of past inoperability of the NN11 inverter and SB038 loads. The inspectors reviewed the licensees submittal and determined that the report included the potential safety consequences and necessary corrective actions, but it did not thoroughly document the event, in that the effects on the plant from the inverter transfer to its alternate ac power source were not described. The licensee entered the licensee event report completeness issue into their corrective action program as Callaway Action Request 201504217 and initiated a corrective action to submit a revision of the licensee event report at a later date. Because it was not reasonable for the licensee to have been able to foresee and correct the condition that caused the switch failure, the NRC determined that no performance deficiency existed. Thus, the NRC is exercising enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and is not issuing enforcement action for the violation (EA-15-152). Further, because the licensees action and/or inaction did not contribute to this violation, it will not be considered in the assessment process or the NRCs reactor oversight process action matrix. This licensee event report is closed. These activities constitute completion of one event follow-up sample, as defined in Inspection Procedure 71153. |