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| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000317/FIN-2018410-01 | 30 September 2018 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2018001-01 | 31 March 2018 23:59:59 | Calvert Cliffs | Self-revealing | Failure to Conduct Adequate Radiation Surveys and Evaluate Potential Radiological Hazards | Aself-revealed Green non-cited violation(NCV)of Title 10 Code of Federal Regulations(10 CFR) 20.1501, Surveys and Monitoring: General, was identified when Exelon failed to perform adequate surveys of the 11 reactor coolant pump bay area following the aggregation of 25 high dose-rate in-core detectors in one area of the flooded refueling cavity, which is adjacent to the pump bay. Surveys were not performed as required after radiological conditions changed and radiological hazard mitigation measures, such as locking and controlling access in accordance with Exelon procedures, were not implemented, resulting in accessible dose-rates of up to 2,000 millirem per hour(mrem/hr)in the pump bay |
| 05000317/FIN-2017004-01 | 31 December 2017 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Assessment of Fire Brigade Performance During an Announced Fire Drill | An NRC-identified Green non-cited violation (NCV) of Calvert Cliffs Nuclear Power Plant Renewed Facility Operating License DPR-53, DRP-69, Condition E, was identified for Exelons failure to adequately assess the performance of the fire brigade during an announced fire drill. Specifically, Exelon failed to properly assess the command and control performance of the fire brigade leader (FBL) which resulted in the fire drill being improperly evaluated as having met the assessment criteria. The inspectors determined that Exelons failure to properly assess fire brigade performance in accordance with OP-AA-201-003, Fire Drill Performance, Revision 16, was a performance deficiency. Exelon has entered this issue into their corrective action program (CAP) as action request (AR) 04094397The inspectors reviewed IMC 0612, Appendix B, Issue Screening, issued on September 7, 2012, and determined the issue is more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and adversely affected its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to properly evaluate the performance of the fire brigade and correct identified deficiencies adversely affects the fire brigades ability to protect against the effects of a fire. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on October 7, 2016, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power issued on June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) since it involved fire brigade training requirements, the fire brigade demonstrated the ability to meet the required times for fire extinguishment for the fire drill scenario, and the finding did not significantly affect the fire brigades ability to respond to a fire. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Self-Assessment, because Exelon did not conduct a self-critical and objective assessment of the fire brigades performance. Specifically, Exelon failed to conduct a self-critical and objective assessment of the FBLs performance during the fire drill described above. |
| 05000317/FIN-2016004-01 | 31 December 2016 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Inspection of Caulking, Seals, and Expansion Barriers in the Auxiliary Building | Green. The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR), Appendix B, Criterion XVI, Corrective Action, for Exelons failure to identify conditions adverse to quality at CCNPP. Specifically, several safety related auxiliary building caulking, seals, expansion joints, and penetration barriers were found by the inspectors or revealed themselves by water intrusion events to be degraded. The inspectors determined that Exelons failure to identify degradation of several auxiliary building caulking, seals, and expansion joints was a performance deficiency that was reasonably within its ability to foresee and correct and should have been prevented. Exelons immediate corrective actions included performing operability determinations on degraded barriers, and repair of the degraded barriers. Exelon entered these issues into its corrective action program (CAP) as action request (AR) 02715188, AR 02715199, AR 02716543, AR 02725901, and AR 02564655. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, issued on May 6, 2016, and determined the issue is more than minor because it adversely affected the Human Performance attribute, of the Auxiliary Building Area, of the Barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, and found it was sufficiently similar to Example 3.k, in that significant programmatic deficiencies were identified that could have led to worse outcomes. Specifically, several inspection programs designed to identify degraded barriers, caulking, seals, and expansion joints in safety related auxiliary building barriers, had not been performed, or had been performed inadequately. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on October 7, 2016, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power issued on June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) since, the only safety related degradation represented by the finding is of the radiological barrier function provided for the auxiliary building. The inspectors determined that the cause of the finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon did not complete the baseline inspection required by AMBD-0026 within the 10 years preceding entry of Units 1 and 2 into their respective periods of extended operation as specified in CNG-CM-6.01. Additionally, inspections conducted under AMBD-0052, and 0-013-49-O-18M were inadequate in that they failed to identify degradation of the barriers as described above. (H.8) |
| 05000317/FIN-2016008-01 | 30 September 2016 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | A violation of low to moderate safety significance (White) for CCNPP Unit 1 and a violation of very low safety significance (Green) for CCNPP Unit 2 were identified by Exelon and are violations of NRC requirements. The violations were licensee identified during the process of converting to a risk-informed performance-based fire protection program under National Fire Protection Association Standard 805. The NRC screened the issues and determined that it warranted enforcement discretion per Section 9.1 of the NRC Enforcement Policy, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and satisfy the criteria of the NRC Enforcement Policy for being dispositioned as non-cited violations. The violations are described in Section 1R05.06 of this inspection report. |
| 05000317/FIN-2016003-01 | 30 September 2016 23:59:59 | Calvert Cliffs | NRC identified | Deficient Design Control of Air Pressure Available for Unit 1 Component Cooling Water Air Operated Valves | The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for Exelons failure to establish measures to assure that the design basis was correctly translated into specifications affecting safety related functions of air operated valves (AOV). Specifically, when implementing a design change, Exelon failed to verify the air pressure supplied to AOVs in the component cooling (CC) water system was adequate to ensure that the valves would have performed their safety function to close during certain specific accident conditions. The inspectors determined that Exelons failure to verify ECP-15-000213 ensured that air pressure supplied to safety related Unit 1 CC heat exchanger (HX) outlet AOVs was sufficient to support their safety function of closing during a design basis accident (DBA) was a performance deficiency that was reasonably within its ability to foresee and correct and should have been prevented. Exelons immediate corrective actions included conducting an engineering evaluation that demonstrated the operability of the CC system in the degraded condition and increasing the air pressure supplied to the CC HX outlet valves to ensure the valves are capable of fully closing during a DBA. Exelon entered this issue into its corrective action program (CAP) as action request (AR) 02680281. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected the design control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, and found it was sufficiently similar to Example 3.j, in that the design analysis deficiency resulted in a condition where reasonable doubt existed regarding the operability of the Unit 1 CC HX outlet valves. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, issued on June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) since, the finding did not involve an actual open pathway in the physical integrity of reactor containment. The inspectors determined that the cause of the finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelons AOV program, as implemented by ER-AA-410, Air Operated Valve Implementing Program, Revision 2, did not require that complete, accurate, and up-to-date documentation on the CC HX outlet valves design be maintained. (H.7) |
| 05000317/FIN-2016002-02 | 30 June 2016 23:59:59 | Calvert Cliffs | NRC identified | Failure to Report Conditions as Required by 10 CFR 50.73 | The inspectors identified a Severity Level IV, NCV of 10 CFR 50.73(a)(2) for Exelons failure to report within 60 days of discovery, a condition that could have prevented the fulfillment of the safety function of the service water (SRW) system needed to mitigate the consequences of an accident. Additionally, Exelon failed to report within 60 days of discovery, a single condition that caused two trains of the SRW system, a system designed to mitigate the consequences of an accident, to become inoperable. Exelon entered the issue into their CAP as IR 02688409 and on July 20, 2016, submitted LER 05000317/2016-004-00, High Energy Line Break Barrier Breached Due to Human Performance Error Causing Both Service Water Trains to be Inoperable. The inspectors determined that Exelons failure to report a single condition that caused the inoperability of two trains of SRW and may have prevented SRW from fulfilling its design functions to mitigate the consequences of an accident within 60 days of discovering the condition was a violation of 10 CFR 50.73(a)(2), and could have impacted the regulatory process. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and the NRC Enforcement Policy, revised February 4, 2015, and determined the violation is of SL-IV because it is most similar to example 6.9.d.9 of the NRC Enforcement Policy, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73, which is a SL-IV violation. The inspectors determined that the violation did not have a cross-cutting aspect because it involved the traditional enforcement process only. |
| 05000317/FIN-2016403-02 | 30 June 2016 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2016301-01 | 30 June 2016 23:59:59 | Calvert Cliffs | NRC identified | SRO upgrade candidates performing RO duties without completing the requalification program | The NRC identified a Severity Level IV NCV of 10 CFR 50.54(i-1) and 10 CFR 55.59(a)(1), in that Exelon reduced the scope of the requalification program without NRC approval by allowing two SRO upgrade candidates, who were not current in requalification training, to perform RO duties with their qualifications lapsed. Immediate actions by Exelon included relieving one of these individuals from duty and replacing him with a qualified operator, suspending the RO qualifications for both SRO upgrade candidates, and initiating an apparent cause evaluation (CR 02648066). The inspectors determined that this violation was associated with a minor deficiency because the failure to follow training and qualification procedures had no safety impact. However, this violation impacted the regulatory process in that these licensed operators performed licensed duties while in non-compliance with their licenses. According to the Enforcement Policy, operators being in noncompliance with a condition stated on their licenses could be a Severity Level III violation. However, because no operational issues resulted from these individuals performance, the NRC determined that a Severity Level IV violation was more appropriate. In accordance with Inspection Manual Chapter 0612, because this violation involved traditional enforcement and does not have an underlying technical violation that would be considered more than minor, a cross-cutting aspect was not assigned to this violation. |
| 05000317/FIN-2016403-01 | 30 June 2016 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2016201-01 | 30 June 2016 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2016002-01 | 30 June 2016 23:59:59 | Calvert Cliffs | NRC identified | Scaffolding Impairs Fire Sprinkler Systems in Safety Related Fire Areas | The inspectors identified a Green, NCV of CCNPP Renewed Facility Operating License for Units One and Two, paragraph 2.E for Exelons failure to maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report (UFSAR). Specifically, Exelon installed scaffolding in safety related areas not in accordance with approved procedures and, therefore, impaired fire sprinkler systems that were required by the approved fire protection program without establishing approved contingency measures. The inspectors determined that Exelons impairment of fire sprinkler systems by installing scaffolding with dimensions exceeding those approved in Exelon procedure MA-AA-716-025 was a performance deficiency that was within Exelons ability to foresee and prevent. The performance deficiency led to the violation of CCNPP Renewed Facility Operating License, paragraph 2.E, because Exelon failed to maintain in effect all provisions of the approved fire protection program. Exelons immediate corrective actions included stationing continuous fire watches and removal of the scaffolding deck boards which were impairing the fire sprinkler systems. Exelon entered these issues in to their corrective action program (CAP) as issue reports (IR): 02642463, 02642549, 02642844, 02644495, 02647104, 02647454, and 02647455. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon installed scaffolding that exceeded the allowed dimensions in MA-AA-716-025 and impaired the function of fire sprinkler systems in areas containing safety related equipment. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix F, The Fire Protection SDP Worksheet issued on September 20, 2013 and determined the finding to be of very low safety significance (Green) because, in all cases of impairment, the fire sprinkler systems were still capable of protecting their intended targets or were still capable to suppress fires such that no additional equipment important to safety would have been affected. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon failed to properly implement procedure MA-AA-716-025, Scaffold Installation, Modification, and Removal Request Process, Revision 11, which limits scaffolding dimensions and locations when installing scaffolding in safety related areas. (H.8) |
| 05000318/FIN-2016002-03 | 30 June 2016 23:59:59 | Calvert Cliffs | Self-revealing | Failure to Implement Engineering Change Procedures Results in Plant Trip | The inspectors documented a self-revealing, Green finding for Exelons failure to implement procedures for engineering changes. Specifically, Exelon failed to address the full scope and critical parameters associated with a modification to a steam generator feed pump (SGFP). As a result, the 22 SGFP turbine pedestal studs were improperly torqued, resulting in the SGFP shifting, becoming misaligned, and eventually resulting in the failure of the turbine to pump coupling. This resulted in the unexpected tripping of the 22 SGFP on December 1, 2015, and operators inserting a manual reactor trip as required by procedure. The inspectors determined that Exelons failure to properly implement procedures CNG-CM-1.01-1003, Design Inputs and Change Impact Screen, Revision 00601, Attachment 12; CNG-CM-1.01-2000, Scoping and Identification of Critical Components, Revision 00201; and CNG-FES-007, Preparation of Design Inputs and Change Impact Screen, Revision 00010 was a performance deficiency that was a performance deficiency that was within Exelons ability to foresee and prevent. Exelons corrective actions included, replacing the failed coupling, verifying the torque on the 21 SGFP using a HYTORCTM, and developing an adverse condition monitoring plan for Unit 1s SGFPs. Exelon conducted a root cause evaluation (RCE) and developed corrective actions to preclude repetition (CAPR) including implementation of Exelon procedure HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Review, Revision 007 and conducting critical parameters and rigor training for engineering personnel including the expectations for three pass reviews and verification of assumptions. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and IMC 0612, Appendix E, Examples of Minor Issues and determined the issue is more than minor because it was associated with the Design Control Attribute of the Initiating Events Cornerstone and adversely impacted the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in a reactor trip from full power on December 1, 2015. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, issued on June 19, 2012 and determined the finding to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon failed to develop and maintain complete and accurate engineering change packages (ECP), work orders (WO), and maintenance procedures.(H.7) |
| 05000317/FIN-2016001-01 | 31 March 2016 23:59:59 | Calvert Cliffs | NRC identified | Issue of concern Regarding Characterization and Acceptance of a Relevant Indication in Pressurizer to Nozzle Dissimilar Metal Weld | An unresolved item (URI) was identified by the inspectors relating to an issue of concern involving Exelons acceptance and characterization of the relevant indication in weld 4-SR-1006-1 during prior refuel outages. Additional information is required to determine whether a performance deficiency, which is more than minor, exists. Description. Based on a review of Exelon letter dated February 25, 2016, the inspectors preliminarily concluded the relevant indication in weld 4-SR-1006-1 was incorrectly accepted during prior refuel outages and was not in conformance with ASME Code Section XI, Article IWA-3000. Additional inspection, including review of Exelons root cause analysis of this issue, is warranted to determine whether a performance deficiency, which is more than minor, exists related to characterization and acceptance of a relevant indication in weld 4-SR-1006-1. (URI 05000317/2016001-01, Issue of Concern Regarding Characterization and Acceptance of a Relevant Indication in Pressurizer to Nozzle Dissimilar Metal Weld) |
| 05000317/FIN-2015004-04 | 31 December 2015 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | 10 CFR 55.21 and 10 CFR 55.33 state, in part, that licensed operators are required to have a physical examination every two years to ensure that their medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety. As part of licensed operator medical evaluations, screening questions to identify potentially disqualifying medical conditions are required as specified in ANSI/ANS-3.4-1983, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants. Contrary to this requirement, as a result of Exelons medical examination audit completed August 8, 2014, Exelon identified nine (9) licensed operators who were given an incomplete health questionnaire during their biennial medical examination. The questionnaire failed to request information about seven (7) potentially disqualifying health conditions from ANSI/ANS-3.4-1983 during a biennial medical examination. The omission of these seven potentially disqualifying conditions from the questionnaire resulted in an incomplete medical examination. Exelon identified that the cause was an incorrect revision to the sites medical examination process procedure. The revision issue was corrected in a subsequent revision and the audit documented that the nine licensed operators all completed medical evaluations with the correct screening questions within the next 18 months. The results of the medical examination audit were documented in IR 2423783. This violation is subject to traditional enforcement because of the potential impact upon regulatory process because the operators medical conditions are reviewed by the NRC when issuing or renewing operator licenses. The inspectors determined that this issue meets the criteria for a Severity Level IV violation using example 6.4.d.1(c) from the NRC Enforcement Policy because the operators who potentially did not meet ANSI/ANS-3.4, Section 5, due to an incomplete medical examination, subsequently were found to meet the health requirements for licensing. This is of very low safety significance because no incorrect regulatory decision was made as a result of the incomplete medical questionnaire and because no changes to license restrictions were required. |
| 05000317/FIN-2015007-01 | 31 December 2015 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Verification of Offsite Power Operability Limit | The team identified a finding of very low safety significance involving a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, "Design Control," because Exelon did not ensure the operability of offsite power in design calculations. The team determined that non-conservative assumptions caused the results of the voltage calculation to predict higher 4160 Volts, Alternating Current (VAC) switchgear post-trip voltage levels than those which could occur with existing controls. Specifically, the team found that Exelons calculation assumed a 3.2 percent switchyard voltage drop upon main generator trip, which did not bound the 5 percent alarm setting provided by the Transmission System Operator Security Analysis application. The team also determined that Exelon used a non-quantitative evaluation, which could not be verified, to adjust design basis calculation results in order to show that during a design basis event the 4160 VAC bus voltage would recover in time to reset the degraded voltage relay prior to the transient degraded voltage relay (TUR) tripping (causing a loss of offsite power). The team could not determine if offsite power would be lost during the event because these assumptions could not be validated. Exelon entered the issue into the corrective action program and performed preliminary computer modeling of the current plant configuration that showed offsite power was operable. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone design control attribute and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and was similar to Example 3j in Appendix E of the NRC IMC 0612. Using the NRC IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because it was a design deficiency confirmed not to result in the loss of operability or functionality. This finding was not assigned a cross-cutting aspect because it was a historical design issue not indicative of current performance. |
| 05000317/FIN-2015004-03 | 31 December 2015 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | 10 CFR 55.25 states, in part, that if an operator develops a permanent physical or mental condition that causes the operator to fail to meet the requirements of 10 CFR 55.21, the facility licensee shall notify the Commission within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c) which states that the regional administrator shall be notified if a licensed operator develops a permanent disability or illness. Contrary to these requirements, as the result of Exelons medical examination audit completed August 8, 2014, Exelon identified four cases in which a change in licensed operator medical conditions were not communicated to the NRC within the required 30 days. The results of the medical examination audit were documented in IR 2423780 and subsequent notifications were made to the NRC. This violation is subject to traditional enforcement because of the potential impact upon the regulatory process for issuing restrictions to operators licenses. The inspectors determined that this issue meets the criteria for a Severity Level IV violation using example 6.4.d.1(a) from the NRC Enforcement Policy because no incorrect regulatory decision was made as the result of the failure of the licensee to report within 30 days. This is of very low safety significance because after NRC review of the subsequent notifications, no changes to license restrictions were required. |
| 05000317/FIN-2015004-02 | 31 December 2015 23:59:59 | Calvert Cliffs | NRC identified | AFAS Channel Inoperable due to Valve Misposition | The inspectors documented a self-revealing Green NCV of TS 5.4.1.a for Exelons failure to implement procedures as required by RG 1.33, Appendix A, Section 8, Procedures for Control of Metering and Testing Equipment and for Surveillance Tests, Procedures, and Calibrations, during maintenance which resulted in a manual isolation valve (1HVFW-1804) being incorrectly placed in the closed position. This human performance error isolated the number 12 steam generator (SG) wide range level transmitter (1LT1124C) and subsequently rendered the auxiliary feedwater actuation system (AFAS) sensor channel ZF inoperable for 33 hours and 39 minutes, a condition prohibited by TS 3.3.4, Engineered Safety Features Actuation System (ESFAS) Instrumentation. The inspectors determined that the failure to properly implement procedure STP M-525AT-1 and place 1HVFW-1804 in its required position was a performance deficiency that was reasonably within Exelons ability to foresee and prevent. Upon identification, Exelon staff entered this issue into their CAP as condition report (CR)-2014-003320. Exelons immediate corrective action was to enter TS 3.3.4.A, to determine and correct the cause, and to retest the system for proper operation. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected the configuration control attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon operated with manual isolation valve, 1HVFW-1804 closed which resulted in the inoperability of the AFAS sensor channel ZF for approximately 33 hours and 39 minutes. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, issued on June 19, 2012, the inspectors determined that a detailed risk evaluation was necessary to disposition the significance of this finding because the finding represented an actual loss of function of at least a single train of AFAS for greater than its TS allowed outage time. A regional SRA performed a detailed risk evaluation. The finding was determined to be of very low safety significance (Green) because the redundant AFAS sensor was operable and functional to ensure actuation of the system if it had been required, therefore there was no loss of the system function. Additionally, the unit was in Mode 3 with very low decay heat levels during the time the ZF sensor channel was determined to be inoperable and plant procedures exist to manually start the AFW system if failure of automatic actuation were to occur. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not stop when faced with an uncertain condition about the position of 1HVFW- 1804. Specifically, personnel conducting the second verification did not appropriately question the position of isolation valve 1HVFW-1804 because of the higher experience level of the personnel conducting the first verification. |
| 05000317/FIN-2015004-01 | 31 December 2015 23:59:59 | Calvert Cliffs | NRC identified | Failure to Implement Procedures for the Control of Hazard Barriers During Maintenance | The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1.a for Exelons failure to implement procedures as required by Regulatory Guide (RG) 1.33, Appendix A, Section 1, Administrative Procedures, during replacement of the 11 service water (SRW) pump motor, resulting in the SRW pump room door, a high energy line break (HELB) barrier, being impaired. This rendered the safety-related equipment protected by the HELB barrier inoperable. The inspectors determined that the failure to properly implement Exelon procedures EN-1-135, Control of Barriers, Revision 00202, and CC-AA- 201, Plant Barrier Control Program, Revision 11, was a performance deficiency that was reasonably within Exelons ability to foresee and prevent. Upon identification, Exelon staff entered this issue into their corrective action program (CAP) as issue report (IR) 2586773. Exelons immediate corrective actions included halting of impairing hazard barriers without considering the degraded barriers effect on equipment operability. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelons actions in blocking open the HELB barrier resulted in a condition where structures, systems, and components (SSCs) necessary to mitigate the effects of a HELB may not have functioned as required; therefore, the reliability of these protected SSCs was adversely impacted. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, issued on June 19, 2012, the inspectors determined that a detailed risk evaluation was necessary to disposition the significance of this finding because the finding represented a loss of the SRW system. A regional Senior Reactor Analyst (SRA) performed a detailed risk evaluation using an exposure interval of 10 minutes as the maximum time the condition was allowed in the plant. Using these inputs yielded an initiating event frequency of 4E-9/year. From discussions with the inspectors, the analyst confirmed a list of affected equipment. The analyst bounded the scenario by assuming all mitigating equipment would be lost which gave a maximum change in core damage frequency of 4E-9/year. Since the bounded change in core damage frequency was less than 1E-6, the finding was determined to be of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Work Management, because Exelon did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. Specifically, Exelons process for planning and controlling maintenance did not identify the applicability of Exelon procedure CC-AA-201. |
| 05000317/FIN-2015007-02 | 31 December 2015 23:59:59 | Calvert Cliffs | NRC identified | Failure to Verify AC Equipment Operability at Design Loading and Voltage Levels | The team identified a finding of very low safety significance (Green) involving a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, because Exelon failed to verify, in design basis calculations, that all required Class 1E alternating current (AC) components would perform their safety functions during design basis events. Specifically, the team found multiple examples where Exelon failed to ensure AC equipment operability and functionality at maximum postulated loading levels and minimum allowable voltage levels. Specifically, the team found that during design basis events several transformers exceeded their manufacturers rating and Exelon had not performed an analysis that demonstrated voltage trip setpoints of the degraded voltage relays would ensure adequate voltage was available to supplied equipment. Exelon entered this issue into the corrective action program and performed preliminary analysis to show that there was reasonable assurance that equipment remained operable. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone design control attribute and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and was similar to Example 3j in Appendix E of the NRC IMC 0612. Using the NRC IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because it was a design deficiency confirmed not to result in the loss of operability or functionality. The team did not identify a cross-cutting aspect with this finding because it did not represent current performance. |
| 05000317/FIN-2015010-01 | 31 December 2015 23:59:59 | Calvert Cliffs | NRC identified | Untimely Actions to Test or Inspect DFO Check Valves Relied on for Safety | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not assure that conditions adverse to quality were promptly corrected. Specifically, from November 2012, until October 28, 2015, Exelon did not ensure that diesel fuel oil (DFO) transfer system header check valves DFO-146 and DFO-148 were properly tested or inspected to ensure they would perform their safety function. This issue was previously documented as a NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, in inspection report 05000317, 318/2013003. The inspectors determined that not promptly correcting a condition adverse to quality previously documented in an NCV was a performance deficiency that was within Exelons ability to foresee and prevent. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the safety function of DFO-146 and DFO-148, to close on the failure of a fuel oil storage tank to prevent draining the unaffected tank had never been verified though test or inspection since initial plant construction; therefore, reasonable doubt exists whether the valves remained capable of performing that function. The inspectors evaluated the significance of this finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The 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, flooding, or severe weather event. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance Procedure Adherence because Exelon staff did not follow station processes, procedures, and work instructions. Specifically, Exelon staff did not ensure corrective action due date extensions and cancellations were justified, evaluated for adverse consequences, and presented to the Management Review Committee (MRC) as required by station procedures. As a result, corrective actions to restore compliance were not completed in a timely manner. |
| 05000317/FIN-2015003-02 | 30 September 2015 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | 10 CFR 74.19 (c), Recordkeeping, states, in part that, each licensee who is authorized to possess special nuclear material (SNM), shall conduct a physical inventory of all SNM in its possession, under license, at intervals not to exceed 12 months. Contrary to this, on May 22, 2015, Exelon identified that the 2014 SNM inventory had not been completed by the end of August 2014, as was required since the 2013 SNM inventory was completed in August 2013. The 2014 SNM inventory was started on August 26, 2014, and was completed on October 6, 2014. Exelon subsequently self-identified that inventories of nine locations had exceeded 12 months although all SNM was accounted for by October 6, 2014. The inspectors determined that this finding was of very low safety significance (Green), because the finding did not represent an actual loss of SNM and the performance of an inventory in June 2015, as part of the corrective actions, was completed satisfactorily. The inspectors determined that Exelon correctly evaluated the finding and developed appropriate corrective action as documented in Exelons CAP as IR02504484. |
| 05000317/FIN-2015003-03 | 30 September 2015 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | TS 5.4.1.a states, in part, that written procedures shall be established and maintained covering the applicable procedures recommended in RG 1.33, Revision 2, Appendix A, February 1978, of which Section 9 specifies procedures for performing maintenance. The vendor technical manual specifies the need to conduct routine lube oil sample analysis and Exelon procedure MA-AA-716-006, Control of Lubricants Program, Revision 11, directs the performance of sampling in accordance with specific site approved procedures. Contrary to the above, following the June 17, 2015, failure of the 1A EDG surveillance test, Exelon identified that appropriate procedural guidance did not exist for the processing of 1A EDG engine lube oil samples. On June 17, 2015, during surveillance testing of the 1A EDG, Exelon secured the engine due to high lube oil filter differential pressure. The engine lube oil filters were determined to be clogged due to engine coolant contamination of the engine lube oil system caused by leakage past O-rings on one engine cylinder piston. Investigation determined that monthly engine lube oil samples were not provided to the vendor for analysis from February May 2015 due to the extended absence of the regular lubrication specialist and lack of procedural guidance for processing of lube oil samples once they were obtained. Subsequent analysis of these samples revealed that the engine lube oil had elevated potassium levels which is indicative of lube oil contamination by engine coolant. The inspectors evaluated the issue using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings AtPower, which determined that the finding was of very low safety significance (Green) because the safety function was not lost and the 1A EDG was not considered inoperable for greater than its TS limiting condition for operation allowed outage time. The inspectors determined that Exelon correctly evaluated the finding and developed appropriate corrective action as documented in Exelons CAP as IR02517365. |
| 05000317/FIN-2015003-01 | 30 September 2015 23:59:59 | Calvert Cliffs | NRC identified | Failure to Establish and Maintain Procedures for the Operation of the Diesel Fuel Oil System | The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1.a for Exelons failure to adequately establish and maintain procedures as required by Regulatory Guide (RG) 1.33, Appendix A, Section 3, Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems. The inspectors determined that Exelons failure to adequately establish and maintain a procedure for the operation of the diesel fuel oil (DFO) supply system was a performance deficiency. Exelon entered this issue into their corrective action program (CAP) as issue report (IR) 02541107. Exelons immediate corrective actions included halting of opening of 0-DFO-108, 21 Fuel Oil Storage Tank (FOST) to Auxiliary Boilers Isolation, and initiating an evaluation to determine the seismic adequacy of the piping downstream of 0-DFO-108. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately establish and maintain procedure Operating Instruction (OI)-21D, Fuel Oil Storage and Supply, Revision 10, for the operation of the DFO supply system resulted in the alignment of the safety-related 21 FOST to nonsafety-related/non-seismically qualified piping thus rendering the 21 FOST inoperable. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 4, External Events Screening Questions, issued on June 19, 2012, the inspectors determined that a detailed risk evaluation was necessary to disposition the significance of this finding because the loss of the 21 FOST would degrade two or more trains of a multi-train system or function. A regional Senior Reactor Analyst (SRA) performed a detailed risk evaluation and determined the finding to be of very low safety significance (Green). The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Exelon failed to adequately evaluate relevant external operating experience. Specifically, Exelon failed to evaluate for systems where non-seismically qualified piping could be connected to safety-related tanks as was described in Information Notice (IN) 2012-01, Seismic Considerations Principally Issues Involving Tanks. (P.5). |
| 05000317/FIN-2015002-02 | 30 June 2015 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Maintenance Instructions for Replacement of the Units 1 and 2 Containment Air Cooler Starters | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to include appropriate quantitative acceptance criteria for determining the auxiliary contacts and mechanical interlocks were properly installed and adjusted when the Units 1 and 2 containment air coolers (CAC) starters and contactors were replaced during plant modifications. The starter and contactors with associated mechanical interlocks and auxiliary contacts provide the necessary electrical coordination to shift the CACs from fast to slow speed during a safety injection actuation signal (SIAS). The starter and contactor replacements occurred from July 2002 to July 2004. The inspectors determined that Exelons failure to include appropriate quantitative acceptance criteria for determining the auxiliary contacts and mechanical interlocks were properly installed and adjusted when the Units 1 and 2 CAC starters and contactors were replaced during plant modifications is a performance deficiency. Exelon entered this issue into their corrective action program (CAP) as IR02408755, completed an apparent cause evaluation (ACE), and completed corrective action work orders (WO) to adjust all associated starters and contactors auxiliary contacts. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it is associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Mitigating Systems Screenings Questions, issued on June 19, 2012, and determined a detailed risk evaluation was required for the actual loss of function of the 13 CAC for greater than its technical specification (TS) allowed outage time. A regional Senior Reactor Analyst performed a detailed risk evaluation using the Calvert Cliffs Standardized Plant Analysis Risk (SPAR) Model for Calvert Cliffs Unit 1, Version 8.27, for internal events and determined the finding to be of very low safety significance (Green). The inspectors determined that the finding did not have a cross-cutting aspect because the issue was not indicative of current licensee performance. |
| 05000317/FIN-2015002-01 | 30 June 2015 23:59:59 | Calvert Cliffs | NRC identified | Failure to Properly Ship Category 2 Radioactive Material - Quantity of Concern | The inspectors identified a Green NCV of 10 CFR 71.5, Transportation of Licensed Material, and CFR 172, Subpart I, Safety and Security Plans. Specifically, Exelon personnel shipped a Category 2 radioactive material quantity of concern (RAM-QC) on public highways to a waste processor without adhering to a transportation security plan. Prior to shipment, Exelons staff failed to recognize that the quantity of radioactive material met the definition RAM-QC. The inspectors determined that Exelons failure to ship material as a Category 2 RAM-QC was a performance deficiency. Exelon entered this issue into their CAP as IR02481678 and corrective actions included revising the shipping procedure to reflect the appropriate Department of Transportation requirements for shipment of Category 2 radioactive material. Additionally, Exelon implemented a formal process for reviewing pending regulatory changes for impacts to operations and support activities by the implementation of Exelon Procedure LS-AA-110, Commitment Management, Revision 10, in September 2014. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it is associated with the program and process attribute of the Public Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. In accordance with IMC 0609, Appendix D, "Public Radiation Safety Significance Determination Process," issued on February 12, 2008, the inspectors determined the finding to be of very low safety significance (Green) because Exelon had an issue involving transportation of radioactive material, but it did not involve: (1) a radiation limit that was exceeded; (2) a breach of package during transport; (3) a certificate of compliance issue; (4) a low level burial ground nonconformance; or (5) a failure to make notifications or provide emergency information. The inspectors determined that the finding did not have a crosscutting aspect because the issue was not indicative of current licensee performance because Exelon successfully implemented its transportation security plan in shipping three Category 2 RAM-QC packages in 2014. |
| 05000317/FIN-2015001-01 | 31 March 2015 23:59:59 | Calvert Cliffs | NRC identified | Component Cooling Operated in Unanalyzed Condition | The inspectors identified a Green NCV of Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.5, Component Cooling (CC) System, and 3.0.3, because Exelon operated Units 1 and 2 CC systems in an unanalyzed condition on 18 occasions and operated in a condition prohibited by TS on two occasions within the last three years. The inspectors determined that Exelons operation with both CC loops inoperable and the subsequent failure to place the unit in Mode 5 within 37 hours as required by TS is a performance deficiency. Exelon entered this issue into their corrective action program (CAP) as IR02439913. Exelons immediate corrective actions included the submission of event notification (EN) 50752 and prohibiting operation of the CC system in a configuration outside of that specified in the TS bases while further analysis was conducted. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the station operated with two CC loops unable to perform their safety function of maintaining component cooling heat exchanger (CCHX) outlet temperatures at or below 120F. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, issued on June 19, 2012, the inspectors determined that a detailed risk evaluation was necessary to disposition the significance of this finding because the finding represented a loss of a system and/or function. The detailed risk evaluation considered that the deficiency could have, under some ultimate heat sink temperature conditions, resulted in the CCHX outlet temperatures exceeding the design analyzed limit of 120F following the recirculation actuation signal (RAS) during a loss of coolant accident (LOCA). The Senior Reactor Analyst performed a bounding significance determination by conservatively assuming a complete loss of safety function for the CCHXs for the applicable limited exposure time. Emergency operating procedures also had contingencies for a postulated loss of the CC function which directed the re-alignment of a containment spray (CS) pump to ensure adequate safety injection is maintained. This evaluation determined the issue was of very low safety significance (Green). The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Design Margins, because Exelon did not operate and maintain equipment within design margins. Specifically, Exelon operated the CC system outside its design safety-related specification, resulting in an operating condition prohibited by TS. |
| 05000317/FIN-2015001-02 | 31 March 2015 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Risk Management Action for LOCI Sequencer Maintenance | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4) because Exelon did not implement adequate risk management actions (RMA) during the replacement of the loss of coolant incident (LOCI) sequencer for the safety-related 11 4KV (kilovolt) bus in accordance with station procedures. The inspectors determined that Exelons failure to establish adequate RMAs during the performance of LOCI sequencer maintenance activities in accordance with CNG-OP-4.01- 1000 is a performance deficiency. Exelons immediate corrective actions included entering this issue into their CAP as IR02444523 The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it was associated with the procedure quality 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, without adequate RMAs per station procedure CNG-OP-4.01-1000, the capability of the 0C alternate alternating current (AAC) diesel generator (DG) to perform its safety function of powering the 11 4KV bus was adversely impacted. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, and noted that this issue is sufficiently similar to examples 7.e and 7.f, in that, Exelon was required, under plant procedures, to establish RMAs or additional RMAs. The inspectors, with the assistance of a Region I Senior Reactor Analyst, evaluated this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued on May 19, 2005. Using Appendix K, Flowchart 2, Assessment of RMAs, the inspectors determined that the finding was of very low safety significance (Green) based upon the short duration exposure time (approximately one hour). Specifically, comparing the licensees calculated Yellow (1E-5) annualized risk for this maintenance evolution to the actual (1E-4/year X 1 year/8760 hours = 1E-8) incremental risk increase places the risk of this finding below the Incremental Core Damage Probability (ICDP) > 1E-6 threshold, resulting in a very low safety significance (Green). The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Work Management, because Exelon did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, Exelon failed to adequately plan, control, and execute the LOCI sequencer maintenance activity by establishing adequate RMAs that would have provided alternate success paths for maintaining the safety function of the out of service structures, systems, and components (SSCs) |
| 05000317/FIN-2014404-02 | 31 December 2014 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2014404-01 | 31 December 2014 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2014005-02 | 31 December 2014 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | On June 9, 2014, at 5:35 pm, Exelon observed that the 2A EDG field flash monitoring relay alarmed in the main control room. Initial investigation by Exelon incorrectly concluded that this alarm was caused by an alarm card problem and that the operability of the EDG was not impacted. On June 11, 2014, at 10:35 am, subsequent investigation by Exelon determined that the alarm was caused by a loose fuse clip which would have prevented the flashing of the 2A EDG generator field thus rendering the 2A EDG inoperable. Repairs were made, post-maintenance testing was completed, and the 2A EDG was declared operable on June 11, 2014, at 4:32 pm. An ACE concluded that the operations staff incorrectly determined that the 2A EDG was operable based on the available indications without determining the cause of the alarm. The 2A EDG was inoperable for 46 hours and 57 minutes which is a condition that required entry into TS LCO 3.8.1.B for one EDG inoperable. TS LCO 3.8.1.B requires five actions with required completion times ranging from 1 hour to 14 days. Four of these require actions were not performed within the required completion time thus TS Condition 3.8.1.J should have been entered which required the unit to be placed in Mode 3 within 6 hours and Mode 5 within 36 hours from entry into the condition. Neither action of TS LCO 3.8.1.J was taken within the required time, therefore, the condition existed for a time longer than allowed by TS. The inspectors reviewed Exelons ACE and other related documents and determined that no performance deficiency existed becaus Exelons actions in response to the 2A EDG field flash monitoring relay alarm was not inconsistent with station documents and their action were reasonable based on the information available to the operators at that time. The inspectors reviewed LER 05000318/2014-002-00 and determined that traditional enforcement applies in accordance with IMC 0612, Power Reactor Inspection Reports Section 0612-09 and 0612-13 and the Enforcement Policy, Section 2.2.4.d, because a violation of NRC requirements existed without an associated performance deficiency. This issue was considered to be a Severity Level IV NCV of TS LCO 3.8.1.J, in accordance with the Enforcement Policy, Section 6.1.d. This Severity Level IV licensee identified NCV was entered into Exelons CAP as CR-2014-006670. |
| 05000317/FIN-2014005-01 | 31 December 2014 23:59:59 | Calvert Cliffs | NRC identified | Spent Fuel Pool Cask Handling Crane 10 CFR 50.65(a)(2) Performance Not Met | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(2), because Exelon did not adequately demonstrate that the spent fuel pool cask handling crane (SFPCHC) (a)(2) performance was effectively controlled through performance of appropriate preventative maintenance. Specifically, Exelon did not identify and properly account for a maintenance rule functional failure (MRFF) of the SFPCHC in September 2013, and thereby did not recognize that the crane exceeded its performance criteria and required a Maintenance Rul (a)(1) determination. Exelon entered this issue in the corrective action program (CAP) as incident report (IR) 02422876. Exelons immediate corrective actions were to reclassify the September 2013 failure as a MRFF and conduct a Maintenance Rule (a)(1) determination on the SFPCHC. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the finding is more than minor because it is associated with the structure, system, and component (SSC) performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, following the MRFF of the SFPCHC in October 2014, Exelon personnel did not identify that the crane required a Maintenance Rule (a)(1) determination, to establish if the crane should be monitored in accordance with 10 CFR 50.65(a)(1). As a result, an excessive amount of time passed for Exelon to comply with the requirements of the Maintenance Rule. In accordance with IMC 0609.04, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, issued on June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not result in handling errors, dropped storage cask, or crane operations over the spent fuel pool that caused mechanical damage to fuel clad and a detectible release of radionuclides. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Exelon personnel failed to properly evaluate the issue that occurred in September 4, 2013 as a MRFF. |
| 05000317/FIN-2014004-01 | 30 September 2014 23:59:59 | Calvert Cliffs | NRC identified | Main Steam Line Drain Containment Isolation Valves not Scoped in In-Service Testing | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, for Exelons failure to meet the test requirements set forth in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for main steam line drains (MSLDs) and containment isolation valves (CIVs) motor operated valves (MOVs) (6611, 6612, 6613, 6615, 6620, 6621). Specifically, Exelon failed to scope the MSLD MOVs in their in-service testing (IST) program. As a result, the MOVs reliability was not ensured due to valve degradation not being trended as required in the IST program. Also, the MOV operability was in question because the valves were never tested to perform their containment isolation function. Exelon entered this issue into their corrective action program (CAP) as condition report (CR)-2014-005961. Immediate corrective actions included testing the MOVs. The inspectors determined that the failure to scope and meet the testing requirements of the OM Code for MSLD MOVs in accordance with 10 CFR 50.55a was a performance deficiency. This finding is more than minor because it was associated with the 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 (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to scope and test the MSLD MOVs in accordance with the OM Code did not ensure component reliability by monitoring valve degradation and did not provide assurance that the MSLD MOVs would perform their CIV function in order to protect the public from radionuclides releases during a steam generator tube rupture (SGTR) with a loss of offsite power event. The inspectors reviewed IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions issued June 19, 2012, and determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and the finding did not involve an actual reduction of hydrogen igniters in the reactor containment. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance. Specifically, the 2007 IST fourth year interval submittal was the last reasonable opportunity for Exelon to identify this issue. |
| 05000317/FIN-2014004-02 | 30 September 2014 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | TS 3.4.10, Pressurizer Safety Valves, requires two pressurizer safety valves to be operable during Modes 1 and 2, and in Mode 3 when all RCS cold leg temperatures are greater than 365F for Unit 1 or 301F for Unit 2. With one pressurizer safety valve inoperable, TS 3.4.10, Condition A, requires the inoperable valve to be restored within 15 minutes. If this is not able to be completed or if two pressurizer safety valves are inoperable, then TS 3.4.10, Condition B, is entered which requires the unit to be in Mode 3 within 6 hours AND the unit to be cooled down to below 365F for Unit 1 or 301F for Unit 2 within 12 hours. Contrary to the above, on March 12, 2013, Unit 2 pressurizer safety valve BNO4375, which had been installed in position 2RV200 during the previous operating cycle, was measured higher than its TS allowable value during as-found lift point testing. On February 28, 2014, Unit 1 pressurizer safety valves BN04373 and BM07952, which had been installed in positions 1RV200 and 1RV201 respectively during the previous operating cycle, were measured lower than their TS allowable value during as-found lift point testing. In both cases, the valves had been replaced with tested, operable valves prior to discovery of the as-found condition. Exelon concluded that the valve had been inoperable for a period of time greater than the allowed TS outage times specified in TS 3.4.10. Exelon entered both issues into their CAP as CR-2013-002415, CR- 2014-002236, and CR-2014-002237. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that each example was a finding of very low safety significance (Green) because the finding did not represent an actual loss of the pressurizer safety valve systems credited safety function to relieve pressure to prevent RCS pressure from exceeding 110 percent of RCS pipings design pressure. |
| 05000317/FIN-2014003-04 | 30 June 2014 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | TS 3.7.8, Control Room Emergency Ventilation System (CREVS), requires two CREVS trains operable during modes 1, 2, 3, 4, and during movement of irradiated fuel assemblies. With one CREVS train inoperable due to excessive bypass flow, TS 3.7.8, Condition E, is required to be entered. The required action is to restore CREVS train to operable status within seven days. This action was not completed within the required completion time because the issue was discovered after the required completion time had expired. Contrary to the above, one train of CREVS was inoperable from September 22, 2013, through October 3, 2014, due to the 12 PLFF discharge damper failure in its partially open position. As a result, Exelon operated in a condition prohibited by TS for approximately 4 days. Exelon entered this issue into their CAP as CR-2013-007736. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that this finding is of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room. |
| 05000317/FIN-2014003-05 | 30 June 2014 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. IEEE Standard 308, IEEE Standard for Class 1E Power Systems for Nuclear Power Generating Stations, Section 4.4, Design Basis, states, in part, The design basis shall include, as a minimum, the following: (f) the malfunctions, accidents, environmental events, and operating modes (see Table 1) that could physically damage Class 1E power systems or lead to degradation of system performance and for which provisions shall be incorporated. Table 1 includes natural phenomena such as Wind, Hurricane, and Tornado. Contrary to the above, from 1996 until May 2013, a design vulnerability was identified which resulted in a degradation of 1A EDG under certain environmental conditions (Sustained high winds); however, Exelon failed to incorporate this vulnerability into the stations design basis documents and system operating instructions and make provisions to address this required design basis element. Exelon entered this issue into the CAP as CR 2012-000511 and CR 2013-004310, implemented appropriate compensatory actions, and implemented a permanent plant modification (ECP-13-000510) in the spring of 2014 to address this design vulnerability. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, Question 1. Specifically, the finding is a deficiency affecting the design or qualification of a mitigating SSC, in which the SSC maintains its operability or functionality. |
| 05000317/FIN-2014003-01 | 30 June 2014 23:59:59 | Calvert Cliffs | NRC identified | Main Steam Line Drain Containment Isolation Valves not Scoped in IST | An unresolved item (URI) was identified by the inspectors relating to an issue regarding the failure of Exelon to scope main steam line drains (MSLDs) and CIVs motor operated valves (MOVs) (6611, 6612, 6613, 6615, 6620, and 6621) into their inservice testing (IST) program. Description: The inspectors identified an issue of concern involving Exelons scoping of MSLD MOVs into the IST program. The MSLD MOVs are normally open valves with the ability to be remotely-operated from the main control room. The MSLD MOVs are classified as CIVs per UFSAR, Figure 5-10, Containment Structure Isolation Valve Arrangement, Sheet 24 and 25. This figure classifies the main steam penetrations as Type III, and requires the valves to be closed to perform their CIV function. UFSAR, Section 5.2, Isolation System, Subsection 5.2.2, System Design, defines a Type III penetration as a line not directly connected to the reactor coolant system (RCS) or the containment structure atmosphere that has at least one valve, either a check valve or a remotely-operated valve, outside of the containment structure. These valves are classified as American Society of Mechanical Engineers (ASME) Code, Class 2, per drawing 60740, Sheet 0001, Steam Line Drainage System, Revision 39, and M-601, Piping Class Summary Sheets, Revision 49. The ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004, Subsection ISTA, General Requirements, Section ISTA-1100, Scope, states in part, Section IST establishes the requirements for pre-service and IST and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. These requirements apply to: a) pumps and valves that are required to perform a specific function in shutting down the reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident. 10 CFR 50.55a(f)(1), Codes and Standards, requires the establishment of OM Code IST test requirements to components which are classified ASME Code Class 1, 2 and 3. The inspectors require additional information from Exelon to determine if there is a performance deficiency which is more than minor. Specifically, the revision to calculation CA06453, Steam Generator Tube Rupture Accident Using Source Terms; calculation referenced in April 6, 1988, memo from D. S. Elkins to B. B. Mrowca, Impact of the Main Steam Drain Line on the 10CFR100 Limits of the Steam Generator Event; and for CCNPP to research which standard for the design of CIVs the plant was licensed to (equivalent to ANSI N271-1976, Containment Isolation Provisions for Fluid Systems.) The issue is identified as (URI 05000317/318/2014003-01, Main Steam Line Drain Containment Isolation Valves not Scoped in In-Service Testing Program.) |
| 05000318/FIN-2014003-02 | 30 June 2014 23:59:59 | Calvert Cliffs | NRC identified | Inaccurate EAL Threshold Values Incorporated into Site EAL Scheme Change | The inspectors documented a licensee-identified apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2), which preliminarily has been determined to be of low to moderate safety significance (White). Specifically, 10 CF 50.54(q)(2) requires a licensee to develop and maintain an emergency plan which meets the requirements of 10 CFR 50.47(b), and 10 CFR 50, Appendix E. Contrary to thi requirement, from October 11, 2013, through March 4, 2014, CCNPP failed to maintain i effect an emergency plan that met the standards in 10 CFR 50.47(b)(4) and 10 CFR 50 Appendix E, Section IV.B.1 for Unit 2. CCNPP did not maintain an adequate standar emergency level scheme because inaccurate effluent radiation monitor thresholds wer incorporated into Table R-1, Effluent Monitor Classification Threshold. During th replacement of the Unit 2 main steam line radiation monitors (MSLRMs), CCNPPs staf inaccurately calculated the associated emergency action levels (EALs) effluent threshol values for Alert, Site Area Emergency, and General Emergency, and incorporated thes thresholds into Table R-1. This error could have resulted in an over-classification of a event and at the general emergency level potentially resulted in an unnecessary protective action recommendation and could cause offsite response organizations to implement unnecessary protective actions. Exelon identified the issue, entered it into their corrective action program (CAP), implemented appropriate compensatory actions, and initiated corrective actions to revise the EAL table. The inspectors determined the finding no longer presents an immediate safety concern since appropriate compensatory actions have been implemented. The failure to maintain the EAL threshold values in Table R-1 of the site approved emergency plan was a performance deficiency that was within the Exelon staff ability to foresee and correct and should have been prevented. Using IMC 0612, Appendix B, Issue Screening, the performance deficiency was determined to be more than minor because it impacted the procedure quality attribute of the Emergency Preparedness cornerstone and adversely impacts the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, an EAL change was improperly implemented, which could result in an over-classification of an event and at the general emergency level potentially result in unnecessary protective action recommendations and movement of the public. The inspectors utilized IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, to determine the significance of the performance deficiency. The performance deficiency is associated with the emergency classification planning standard and is considered a risk significant planning standard (RSPS) function. This performance deficiency impacts the following required planning standard and RSPS function: 10 CFR 50.47(b)(4), Emergency Classification System. The inspectors were directed by the SDP to compare the performance deficiency with the examples in Section 5.4, 10 CFR 50.47(b)(4), Emergency Classification System, to evaluate the significance of this performance deficiency. Using Table 5.4-1, Significance Examples 50.47(b)(4)," the inspectors determined that the performance deficiency matched an example of a degraded RSPS function, which would be assessed as White. Specifically, the example states, in part, that the performance deficiency would be assessed White if the EAL classification process would result in an over-classification that would lead to off-site response organizations implementing, by procedure, unnecessary protective actions for the public. This condition should also be considered met if the licensee would make a protective action recommendation to the off-site response organizations because of the over-classification. The inspectors determined that the cross-cutting aspect that contributed most to the root cause is H.12, Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction techniques. Specifically, Exelon staff did not independently validate the new EAL threshold values prior to revising and implementing the EAL scheme change. |
| 05000317/FIN-2014003-03 | 30 June 2014 23:59:59 | Calvert Cliffs | NRC identified | Inadequate EAL Initiating Condition HA3.1 | The inspectors identified a Green NCV of 10 CFR 50.54 (q)(2) and 10 CFR 50.47(b)(4) because Exelon did not maintain the emergency plan to adequately meet the standards in 10 CFR 50.47(b)(4). Specifically, Exelon failed to include Unit 1 and Unit 2 component cooling (CC) rooms under EAL initiating condition HA3.1. As a result, an Alert declaration would have not been made during a hazardous gas event in a vital area. Exelon entered this issue into their CAP as condition report (CR)-2014-004683. Immediate corrective actions included revising EAL initiating condition HA3.1 to include the CC rooms and verify that there are no other areas that need to be included in EAL HA3.1. The failure to update the EAL scheme the site approved emergency plan following a plant modification was a performance deficiency that was within the Exelon staff ability to foresee and correct and should have been prevented. Using IMC 0612, Appendix B, Issue Screening, the performance deficiency was determined to be more than minor because it impacted the procedure quality attribute of the Emergency Preparedness cornerstone and adversely impacts the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, a plant modification was completed which required operators to be able to enter the CC room in order to bring the plant to cold shutdown and the EAL scheme was not updated to reflect this change. The inspectors utilized IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, to determine the significance of the performance deficiency. The performance deficiency is associated with the emergency classification planning standard and is considered a RSPS function. This performance deficiency impacts the following required planning standard and RSPS function: The inspectors were directed by the SDP to compare the performance deficiency with the examples in Section 5.4, 10 CFR 50.47(b)(4), Emergency Classification System, to evaluate the significance of this performance deficiency. The inspectors determined that the EAL was ineffective because it, in and of itself, no longer resulted in a timely and accurate declaration of an Alert for the initiating condition. Utilizing Figure 5.4.1, an ineffective EAL where an Alert would not be declared when required would screen as a Green finding. This finding has a cross-cutting aspect in the area of Human Performance, Change Management, because Exelon personnel didnt use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, Engineering personnel did not ensure that the impact to the Emergency Plan was adequately evaluated as a result of the permanent plant change engineering change package (ECP)-11-000983 (H.3). |
| 05000317/FIN-2014002-03 | 31 March 2014 23:59:59 | Calvert Cliffs | Self-revealing | Inadvertent Loss of RCS Inventory During Lowered Inventory Conditions | The inspectors identified a self-revealing NCV of Technical Specification (TS) 5.4.1, Procedures, for the failure of Constellation Energy Nuclear Group, LLC (CENG) personnel to adequately implement procedures associated with a local leak rate test (LLRT). Specifically, CENG personnel did not isolate the letdown line in accordance with surveillance test procedure (STP)-O-108D-1, Containment Penetration Local Leak Rate Tests, prior to draining the piping in preparation for an LLRT on chemical and volume control system (CVCS) containment isolation valves. This resulted in inadvertently draining 150 gallons from the reactor coolant system (RCS) while the reactor vessel was in a lowered inventory condition. Immediate corrective actions included entering this issue into their corrective action program (CAP), performing a prompt investigation, and conducting a safety stand-down. In addition, an apparent cause evaluation will be performed to determine any additional corrective actions. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to isolate the letdown line prior to draining resulted in he loss of 150 gallons of RCS inventory and challenged the critical safety function of inventory control while in a lowered inventory condition. Operator actions were required to identify and isolate the leak to prevent further inventory loss. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, issued February 28, 2005, and determined that the issue screened to Green (very low safety significance). Specifically, the inspectors determined that adequate mitigating capability remained available and the finding did not represent a loss of control of RCS level due to less than 2 feet of inventory loss when not in midloop. As a result, a Phase 2 quantitative assessment was not required and the issue screened to Green. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Teamwork, because CENG individuals and work groups did not adequately communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, a detailed shift turnover between dayshift and nightshift LLRT operators was not completed to ensure that the oncoming operators were aware of the letdown system configuration (H.4). |
| 05000317/FIN-2014002-01 | 31 March 2014 23:59:59 | Calvert Cliffs | Self-revealing | 11 and 12 AFW Pumps Inoperable due to Valves Misposition | The inspectors identified a self-revealing problem consisting of NCVs of TS 3.7.3, Auxiliary Feedwater System, and TS 5.4.1, Procedures, because CENG Operations personnel did not adhere to procedures which resulted in a valve mispositioning event that inadvertently rendered the 11 and 12 turbine driven auxiliary feedwater (AFW) pumps inoperable for approximately 12 hours, a condition prohibited by TSs. Specifically, on February 7, 2014, operators did not perform draining of 11 turbine driven AFW pump steam supply drain line as stated in Operating Instruction (OI)-32A, Auxiliary Feedwater System, resulting in two main steam (MS) drain valves being left opened. With the drain valves open, an actual auxiliary feedwater actuation system (AFAS) signal would have resulted in steam blowing down into the room via the sump and causing room temperatures to exceed 130F, the maximum temperature allowed in the room to protect the pump air cooled bearings. Immediate corrective actions included restoring the proper AFW system valve lineup, entering this issue into their CAP, returning the valves to their normal position on Unit 1, and ensuring that similar valves were in the correct position on Unit 2. Planned corrective actions include conducting an apparent cause evaluation to understand the apparent and contributing causes of this event and determine additional corrective actions. The problem is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, Operations personnel lost configuration control of valves MS-225 and MS-228 resulting in the inoperability of the 11 and 12 AFW pumps for approximately 12 hours. The inspectors evaluated the problem using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, issued June 19, 2012, and determined that the problem represented an actual loss of function of at least a single train for greater than its TS allowed outage time which required a detailed risk evaluation. The senior reactor analyst performed a detailed risk assessment utilizing the CCNPP Unit 1 simplified plant analysis risk model version 8.2.1 and determined that the problem is of very low safety significance (Green). Specifically, given a 12 hour exposure period with both turbine driven AFW pumps assumed to fail-to-run, the change in the internal events core damage frequency (CDF) was calculated to be in the high 10-8 range (Green). This problem has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because CENG personnel did not follow processes, procedures, and work instructions. Specifically, after draining the 11 AFW pump mud leg, CENG plant operators did not restore MS-225 and MS-228 to their required position as stated in procedure OI-32A (H.8). |
| 05000318/FIN-2014002-02 | 31 March 2014 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Compensatory Actions for Out of Service Letdown Radiation Monitor | The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54, Conditions of Licenses, paragraph (q)(2), because CENG did not maintain the Emergency Plan to adequately meet the standards in 50.47(b)(4). Specifically, following the removal of the Unit 2 letdown radiation monitor for maintenance on October 28, 2013, CENG did not establish adequate compensatory measures to ensure that a fuel clad degradation emergency action level (EAL) could be assessed in a timely manner as discussed in the Emergency Plan. This could have resulted in an unnecessary delay in the recognition of a Notice of an Unusual Event (NOUE) EAL declaration for elevated coolant reactivity. Immediate corrective actions included restoring the proper valve lineup, entering this issue into their CAP, and implementing compensatory actions, which included the use of a portable radiation monitor with appropriate alarm setpoints to initiate action to sample the RCS to determine if the specified reactor coolant activity limits are exceeded. Planned corrective actions include restoration of the Unit 2 letdown radiation monitor. This finding is more than minor because it was associated with the emergency response organization performance attribute of the Emergency Preparedness (EP) cornerstone and affected the cornerstones objective to ensure that CENG is capable of implementing adequate measures to protect public health and safety in the event of a radiological emergency. Specifically, the failure to establish compensatory actions beyond the normal RCS sampling frequency could have resulted in exceeding an NOUE EAL threshold for a degraded fuel clad and the condition not becoming known until the next normal RCS sample or upon further fuel clad degradation requiring escalation under other EALs. In accordance with IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued February 24, 2012, the inspectors determined the finding is of very low safety significance (Green). Utilizing IMC 0609, Appendix B, the inspectors determined that the finding is associated with an aspect of the Emergency Plan related to the EAL Classification Scheme 10 CFR 50.47(b)(4). The inspectors determined that the EAL was ineffective because it, in and of itself, no longer resulted in a timely and accurate declaration for the initiating condition. Utilizing Figure 5.4.1, the impact of the ineffective EAL is that a NOUE would be declared in a timely manner, which screens as a Green finding. In addition, the finding is similar to a Green finding in Table 5.4.1, Significance Examples 50.47(b)(4), in that the EAL classification process is not capable of classifying an Alert or NOUE in a timely and accurate manner. This finding has a cross-cutting aspect in the area of Human Performance, Work Management, because CENG personnel adequately implement a work process that included the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, Operations and Chemistry personnel did not ensure that the assigned tasks were adequate to compensate for the increased in nuclear risk associated with having the letdown radiation monitor out of service (H.5). |
| 05000317/FIN-2014403-01 | 31 March 2014 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | |
| 05000317/FIN-2013007-01 | 31 December 2013 23:59:59 | Calvert Cliffs | NRC identified | Failure to Adequately Ensure Cable Protection and Eliminate Potential Secondary Fires | Calvert Cliffs Nuclear Power Plant Operating License conditions 2.E. for both Unit 1 and Unit 2, require that Calvert Cliffs Nuclear Power Plant, LLC, shall implement and maintain in effect all provisions of the approved fire protection program as described in the approved fire protection program (FPP) as described in the Updated Final Safety Analysis Report (UFSAR). UFSAR Section 9.9.1 states in part that the FPP has been developed in accordance with the documents listed in Section 9.9.12 (References 1 through 19). References 4 and 5 are the Interactive Cable Analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 1 and Unit 2 respectively. Assumption 5 in the ICA Manual for Units 1 and 2 states: If a fire causes electrical shorts or overloads, protective devices are assumed to function properly except as affected by the postulated fire. Contrary to the above, on June 18, 2013, Constellation identified seven Unit 1 cables and three Unit 2 cables which were not coordinated, i.e. the cables were undersized and could overheat due to fire induced faults causing secondary fires. The cables traversed multiple fire areas where secondary fires could render additional redundant or alternate safe shutdown equipment unavailable. The violation was historical and occurred when Calvert Cliffs completed its first safe shutdown analysis. Constellation is in transition to NFPA 805 and, therefore, this licensee-identified violation was evaluated in accordance with the criteria established in Section 9.1 of the NRC Enforcement Policy, Enforcement Discretion for Certain Fire Protection issues (10 CFR 50.48). Specifically, because all of the criteria were met, the NRC is exercising discretion and not issuing a violation for this issue. |
| 05000317/FIN-2013005-02 | 31 December 2013 23:59:59 | Calvert Cliffs | NRC identified | Preconditioning of Containment Air Coolers Emergency Outlet Valves | The inspectors identified an NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XI, Test Control, because CENGs in-service test (IST) procedures did not provide instructions to preclude preconditioning of the containmen air cooler (CAC) emergency outlet valves. Specifically, STP-O-065B-2, 21 SRW Subsystem Operability Test, was written such that a full stroke of the CAC emergency outlet valves was allowed prior to performance of the IST stroke time testing of the valves in the open direction. As a result, the 21 CAC emergency outlet valve, 2-CV-1582, was preconditioned during the last four surveillance tests performed on the valve and the 24 CAC emergency outlet valve, 2-CV-1593, was preconditioned during three of the last four surveillance tests performed on the valve. Immediate corrective actions included entering this issue in the CAP. Corrective actions included revising STP-O-065B to prevent future preconditioning of all the CAC emergency outlet valves. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, preconditioning of the CAC emergency outlet valves prior to performing IST stroke time testing could mask valve degradation. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system, and component (SSC), did not represent a loss of system function, did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-TS trains of equipment, designated as having high safety significance in accordance with the maintenance rule program, for greater than 24 hours. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Resources, because CENG staff failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, CENG staff did not provide a complete and accurate procedure that would preclude preconditioning of the CAC emergency outlet valve during in-service testing (H.2(c)). |
| 05000317/FIN-2013005-01 | 31 December 2013 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Emergency and Abnormal Operating Procedures for the Loss of the 21 DC Bus | The inspectors identified an NCV of Technical Specification (TS) 5.4.1 Procedures, because Constellation Energy Nuclear Group (CENG) failed to maintai adequate guidance in Emergency Operating Procedure (EOP) 8, Functional Recover Procedure, and/or Abnormal Operating Procedure (AOP) 7J, Loss of 120 Volt Vital Alternating Current (AC) or 125 Volt Vital Direct Current (DC) Power. Specifically, EOP-8 and/or AOP-7J did not contain adequate instructions to cross-tie the 480 volt AC vital buses to restore the 120 volt AC vital buses during a loss of offsite power (LOOP) event concurrent with a single failure of the 21 125 volt DC bus. As a result, the engineered safety features actuation system (ESFAS) and auxiliary feedwater actuation system (AFAS) would inadvertently actuate on both units if the 120 volt AC vital buses were not restored within a specified period of time. CENG staffs immediate corrective actions included entering this issue into their corrective action program (CAP). Corrective actions planned include revising AOP-7J to add in steps to cross-tie the 480 volt AC vital buses. The finding is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and 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, following a LOOP concurrent with a failure of the 21 DC bus, inadvertent ESFAS and AFAS actuations would occur on both units if power is not restored to the vital 120 volt AC buses. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to stable shutdown condition. The inspectors determined that this finding did not have cross-cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance. Specifically, the inspectors determined that this was a legacy procedure issue and did not note any recent reasonable opportunities for CENG personnel to identify this issue. |
| 05000317/FIN-2013004-01 | 30 September 2013 23:59:59 | Calvert Cliffs | NRC identified | Inadequate Post-Maintenance Test Associated with an Atmospheric Dump Valve | The inspectors identified an NCV of Technical Specifications 5.4.1, Procedures, for the failure of Constellation Energy Nuclear Group (CENG) personnel to establish, implement, and maintain maintenance requirements associated with No. 21 atmospheric dump valve (ADV). Specifically, CENG personnel failed to perform an adequate postmaintenance test (PMT) in accordance with the work instructions for the No. 21 ADV following maintenance and prior to its return to service. As a result, the valve was returned to service in a condition where its containment isolation function was inoperable. Immediate corrective actions included entering this issue into the corrective action program (CAP). Additional corrective actions taken or planned include training Maintenance shop personnel on writing condition reports (CRs) for all failed PMTs and for Operations to ensure that work orders involving ADVs include post-maintenance operability tests for containment closure. The finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the No. 21 ADV was returned to service in a condition where its containment isolation function was inoperable. In addition, the finding is similar to IMC 0612, Appendix E, Example 5.b, in that, the system was returned to service prior to resolution of the degraded condition. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment. Specifically, there was no loss of steam generator tube integrity. Also, the finding did not involve an actual reduction of hydrogen igniters in the reactor containment. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, CAP component, because CENG staff did not ensure that issues potentially impacting nuclear safety were promptly identified, fully evaluated, and that actions are taken to address safety issues in a timely manner, commensurate with their safety significance. Specifically, CENG staff did not implement a CAP with a low threshold for identifying issues such as writing a CR following the identification that the ADV was degraded. |
| 05000317/FIN-2013004-02 | 30 September 2013 23:59:59 | Calvert Cliffs | Licensee-identified | Licensee-Identified Violation | On February 17, 2013, while Unit 2 was in Mode 3 during a refueling outage, CENG personnel identified a pinhole leak at the upper packing leakoff line cap seal weld of pressurizer spray valve 2CV-100F, which constituted RCS pressure boundary leakage. Technical Specifications limiting condition for operation 3.4.13.a, RCS Operational Leakage, limits pressure boundary leakage during plant operation to zero. With any RCS pressure boundary leakage, the technical specifications require the operating unit to be in Mode 3 within 6 hours and to be in Mode 5 within 36 hours. Contrary to the above, based on review of boric acid walkdown data, RCS pressure boundary leakage existed sometime after the last boric acid walkdown conducted in Unit 2 2011 refueling outage and continued during operation for a time longer than allowed by the technical specifications. The inspectors determined that no performance deficiency existed because CENG satisfactorily tested the component using appropriate non-destructive testing prior to installation, identified the boundary leakage through the use of an prescribed monitoring program (boric acid leakage monitoring) and the monitoring frequency was appropriate for the system location (component location inside containment is inaccessible during reactor operation). The inspectors reviewed LER 2013-001-00 and determined that traditional enforcement applies in accordance with IMC 0612, Section 0612-09 and 0612-13 and Enforcement Policy, Section 2.2.4.d, because a violation of NRC requirements existed without an associated significance determination process performance deficiency. This issue was considered to be a Severity Level IV NCV of Technical Specifications limiting condition for operation 3.4.13.a, in accordance with Enforcement Policy, Section 6.1.d. In addition, the inspectors also evaluated this finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors screened the issue and determined that RCS leakage is considered a loss of coolant accident initiator, and evaluated it using the Initiating Event criteria in Appendix A. Assuming worst case degradation, the leakage would not result in exceeding the technical specifications limit for identified RCS leakage (10 gallons per minute) nor would the leakage have likely affected other mitigation systems resulting in a total loss of their safety function. This severity level IV licensee-identified NCV was entered into CENGs CAP as CR-2013-001245. |
| 05000317/FIN-2013202-01 | 30 September 2013 23:59:59 | Calvert Cliffs | NRC identified | Security | |
| 05000317/FIN-2013403-01 | 30 June 2013 23:59:59 | Calvert Cliffs | NRC identified | Security |