Semantic search

This page or action requires JavaScript to work. Please enable JavaScript in your browser or use a browser where it is supported, so that functionality can be provided as requested. For further assistance, please have a look at the noscript help page.

Search Clear all entries

	Quarter	Site	Title	Description
05000289/FIN-2017004-01	2017Q4	Three Mile Island	Failure to correct degraded control rod connections	The inspectors documented a self-revealing finding involving the failure to follow LS-AA-125, Corrective Action Program, Revision 14. Specifically, the licensee failed to take appropriate corrective actions to correct degraded control rod drive mechanism cable connections identified during a 2010 stuck rod event. This resulted in a rod drop event on October 10, 2017, that caused a turbine runback to 55 percent and required a plant shutdown to repair. As an immediate corrective action, the licensee replaced the Bendix 7-pin electrical connector for the control rod drive mechanism (CRDM) and performed extent of condition visual and resistance checks on the other CRDM cables. The issue was entered into their corrective action program (CAP) as issue report (IR) 04061160.The performance deficiency is more-than-minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, a transient resulting from a dropped rod challenged the critical safety function of reactivity control. The inspectors determined that this finding was of very low safety significance (Green) since it did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because despite indications of degradation during inspections in 2013 and 2015, the site failed to ensure that a resolution addressed the cause commensurate with its safety significance (P.2).
05000334/FIN-2017004-01	2017Q4	Beaver Valley	Inadequate Control of Entry into High Radiation Areas	A self-revealing, very low safety significance NCV of Technical Specification (TS) 5.7.1 for failure to control a high radiation area (HRA) was identified. On November 8, 2017, during independent spent fuel storage installation (ISFSI) dry cask loading campaign activities, the failure of multiple barriers resulted in a worker gaining access to an HRA while signed onto an incorrect radiation work permit (RWP) and a subsequent dose rate alarm. Specifically, a worker signed on to an incorrect RWP during a break, and did not recognize that the surveyed work area dose rates were higher than the RWP setpoints. Additionally, radiation protection personnel controlling access to the HRA failed to ensure that the worker was on the correct RWP per plant procedure requirements for a subsequent entry into anHRA. This resulted in the worker entering an HRA under the incorrect RWP and receiving a dose rate alarm of 1,070 millirem per hour. Upon receiving a dose rate alarm, the worker backed away from the area and reported the issue to radiation protection personnel. FENOCs immediate corrective actions included putting the work in a safe condition, performing follow-up surveys, and verifying remaining personnel trip tickets to ensure all individuals were on the correct RWP. FENOC entered the issue into their corrective action program (CAP) as condition report (CR) 2017-11206.The failure to control access to an HRA is a performance deficiency that was within FENOCs ability to foresee and correct and should have been prevented. The performance deficiency is more than minor because it is associated with the Program and Process attribute (Procedures) of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine reactor operation. Specifically, the failure of multiple barriers resulted in a worker gaining access to an HRA while signed on to an incorrect RWP and receiving a dose rate alarm. IMC 0612, Appendix E, Section 6, Health Physics, General Screening Criteria, states that a performance deficiency involving more than one barrier or the loss of a significant barrier would be classified as a more-than-minor performance deficiency. Using IMC 0609,Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low significance (Green) because: (1) it was not an as low as reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding was a human performance cross-cutting aspect associated with avoiding complacency because FENOC failed to ensure individuals recognize and plan for the possibility of mistakes and ensure individuals implement the appropriate error reduction tools, even when expecting a successful outcome (H.12)
05000334/FIN-2017003-01	2017Q3	Beaver Valley	Operability Determinations and Functionality Assessments	Inspection Scope The inspectors reviewed operability determinations for the following degraded or non- conforming conditions based on the risk significance of the associated components and systems: Unit 1 Anchor Darling double disk gate valves evaluation resulting from NRC Information Notice 2017- 03 on July 13, 2017 Unit 1 fire protection system functionality during a fire water header break on July 20, 2017 Impact on Unit 1 SSST 1A from nearby fire water header break on July 20, 2017 Unit 1 EDG exhaust piping not protected from tornado- generated missiles on July 25, 2017 Unit 1 degraded main steam valve room high energy line break door on July 26, 2017 Unit 2 inoperable DRPI impact on verifying operability of control rod F10 on August 25, 2017 Unit 1 EDG 1 -2 building exhaust damper missing louver on September 22, 2017 The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject SSC remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS s and UFSAR to FENOCs evaluations to determine whether the SSCs were operable. The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability , the inspectors determined whether the measures in place would function as intended and were properly controlled by FENOC. 11 b. Findings 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the applicable regulatory requirements and the design basis for SSCs are correctly translated into specifications, drawing, procedures, and instructions. Contrary to the above, FENOC failed to correctly translate the design basis for protection against tornado generated missiles into their specifications and procedures. Specifically, FENOC did not adequately protect Unit 1 EDG s exhausts from tornado generated missiles. FENOC documented the condition adverse to quality in their CAP under condition report 2017 -07550 and took immediate compensatory actions. The inspectors evaluated FENOCs immediate compensatory measures, which included verifying that procedures are in place and training is current for performing actions in response to a tornado. Because this violation was identified during the discretion period covered by Enforcement Guidance Memorandum 15- 002, Revision 1, Enforcement Discretion for Tornado Missile Protection Non- compliance (ML16355A286) and because FENOC has implemented compensatory measures, the NRC is exercising enforcement discretion and is not issuing enforcement action and is allowing continued reactor operation
05000289/FIN-2017003-01	2017Q3	Three Mile Island	Licensee-Identified Violation	The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.Technical specification 4.1.4, Operational Safety Review, requires each remote shutdown system function shown in Table 3.5-4 shall be demonstrated operable by the performance of the following check, test, and calibration. The technical specification surveillance requirement 4.1.4.b states that the licensee shall verify each required control circuit and transfer switch is capable of performing the intended function in accordance with the licensees surveillance frequency control program, in this caseevery refueling interval. Contrary to SR 4.1.4.b, from January, 1987, until September 2017, Exelon did not verify that each required control circuit on the Unit 1 remote shutdown panel was capable of performing the intended function. Specifically, Exelon did not test four of the required six relays for the B EDG either by operation of the components or by performance of a continuity check. Exelons corrective action included entering this issue into the CAP as issue reports 4020064 and 4047426, developing a remote shutdown system testing procedure for the B EDG system, and the completion of a risk evaluation as required by surveillance requirement 4.0.2. The inspectors determined that the finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. It is of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, since the missed surveillance did not impact the ability to reach safe shutdown.
05000219/FIN-2017003-01	2017Q3	Oyster Creek	Inadequate Augmented Offgas System Procedure Resulted in a Manual Scram	A self -revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the augmented offgas (AOG) system operation procedure as required by NRC Regulatory Guide 1.33, Quality Assurance Requirements (Operation), Appendix A, Section 7, Procedures for Control of Radioactivity. Specifically, Exelon procedure 350.1, Augmented Offgas System Operation, did not include adequate guidance for placing the AOG system into a recycle or shutdown configuration following a system trip. Without this guidance, Operations personnel failed to ensure the correct configuration of the AOG system following a partial trip of the system which resulted in degraded main condenser vacuum and a subsequent manual reactor scram on July 3, 2017. This issue was entered into the corrective action program as issue report 4028402. The corrective actions included placing the AOG system in the correct configuration and revising the AOG system operation procedure to provide guidance for verifying proper alignment of the AOG system when the system is in recycle or shutdown. The inspectors determined the performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of Procedure Quality and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to establish an adequate procedure for verifying proper alignment of the AOG system following a full or partial trip of the system resulted in the AOG inlet valve being left in the open position, which allowed demineralized water to be siphoned from the flame arrestor tank and slowly fill the offgas hold- up pipe. This caused a degradation of main condenser vacuum and resulted in operators inserting a manual reactor scram on July 3, 2017. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was a transient initiator that did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition, and therefore was of very low safety significance (Green). The finding had a cross- cutting aspect in the area of Human Performance, Avoid Complacency , because Exelon failed to recognize and plan for the possibility of mistakes or latent errors and implement appropriate error reduction tools by verifying the AOG system was properly aligned following a system trip ; instead , Operations personnel relied upon using a procedure that did not contain adequate guidance to place the AOG system in the correct configuration following a system trip (H. 12)
05000334/FIN-2017002-01	2017Q2	Beaver Valley	Licensee-Identified Violation	The following violation of very low safety significance (Green) was identified by FENOC and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV . TS 3.7.8, "Service Water System", requires two service water trains to be operable. There is no associated action provided for both trains inoperable. LCO 3.0.3 states, in part, that when an LCO is not met and an associated action is not provided, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Act ion shall be initiated within one hour to place the unit, as applicable, in M ODE 3 within 7 hours. Contrary to the above, on August 20, 2015 and August 31, 2015 , FENOC had both trains of service water inoperable for greater than 7 hours while performing the service water full flow test and did not place Unit 2 in Mode 3. FE NOC entered this issue into the CAP as CR 2017- 04023. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings . Because the finding represented a loss of function of a system, a detailed risk evaluation was performed. A Region I senior reactor analyst used the BVPS Unit 2 Standardized Plant Analysis Risk Model version 8.5 to perform the evaluation. A seismic initiating event frequency was obtained from the Risk Assessment of Operational Events Handbook Volume 2, External Events. A surrogate loss -of-offsite - power event was used applying the seismic initiating event frequency for BVPS with a train of service water being failed with no recovery assumed. The finding was determined to be of very low safety significance (Green) because the limited exposure time in this configuration resulted in a change in core damage frequency in the 1E -10/yr range. The dominant core damage sequence was a seismic event with failure of the EDG .
05000219/FIN-2017002-01	2017Q2	Oyster Creek	Inadequate Assessment of Degraded Fuel Oil Filter Impact to Emergency Diesel Generator Operability	The inspectors identified a finding associated with Exelon procedure OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess the No. 2 emergency diesel generator operability with a degraded fuel oil filter. Specifically, Exelon did not adequately assess the capability of the emergency diesel generator to perform its function during its credited duration time of 72 hours. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 3999576 and IR 3990799 and subsequently replaced the fuel oil filter. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. This issue was also similar to Example 3j of IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of the No. 2 emergency diesel generator and additional analysis was necessary to verify operability. The inspectors evaluated the finding using Exhibit 2, Mitigating System Screening Questions, in Appendix A to IMC 0609, Significance Determination Process. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate the issue associated with the degraded fuel oil filter and its impact to the No. 2 emergency diesel generator operability (P.2).
05000289/FIN-2017008-01	2017Q1	Three Mile Island	Failure to Correct Deficiency in Implementing Controls for Pre-Staging Material in the Reactor Building	Green. The inspectors identified a finding of very low safety significance involving a non- cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action Program," because Exelon did not effectively correct a condition adverse to quality regarding the implementation of controls for pre-staging of materials in the reactor building. Specifically, Exelon did not effectively implement corrective actions regarding the control of pre-staging materials in the reactor building during power operations, which resulted in unsecured prohibited material in a location that had the potential, during a large break loss of coolant accident (LOCA), to be transported to and impact the emergency core cooling system (ECCS) sump. Exelon documented this finding in issue reports 2608560 and 2578255. Corrective actions include Exelon to establish a focus team, led by the maintenance manager, to ensure pre-outage loading of the reactor building is conducted in accordance with requirements and directly supervised by Exelon personnel. The performance deficiency is rnore than minor because, if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, without proper controls implemented, materials may be pre-staged in the reactor building in a quantity or configuration that may render the ECCS sump inoperable. The inspectors evaluated the finding against the Mitigating System Cornerstone using Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," and Appendix A, "The Significance Determination Process for Findings At-Power," Exhibit 2, and determined this finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, Field Presence, because Exelon senior managers did not ensure the oversight of work activities by supplemental personnel (H.2).
05000289/FIN-2017008-02	2017Q1	Three Mile Island	Licensee-Identified Violation	The following licensee-identified violation of NRC requirements was determined to be of very low safety significance and meets the NRC Enforcement Policy criteria for being dispositioned as a Non-Cited Violation. Technical Specifications 6.8., "Procedures and Programs," requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix 'A' of Regulatory Guide 1.33, Revision 2, 1978. Regulatory Guide 1.33, Revision 2, "Quality Assurance Program Requirements," Appendix A, requires administrative procedures for access to containment. Exelon Administrative Procedure 1015, Revision 7, "Equipment Storage Inside Class I Building," requires that no equipment shall be stored, placed, or staged inside a Class I Building without an approved Equipment Storage Data Sheet (ESDS). It further states, in part, that within the reactor building materials such as plastic sheeting must be fastened/secured in such a way as to prevent them from being washed into the reactor building sump post-LOCA. Contrary to the above, between October 27, 2015, and October 28, 2015, Exelon did not properly implement a procedure related to the staging of equipment in preparation for a Three Mile Island, Unit 1 refueling outage. Specifically, on October 28, 2015, Exelon performed a reactor building loading walkdown to review the equipment staged for the upcoming outage. During the walkdown, Exelon noted that several items staged in the reactor building were not in accordance with TMI Procedure AP 1 015. Items inappropriately stored included loose plastic, light stands, light bulbs, a Knaack locker box, and bolt cutters. Exelon immediately removed the prohibited items from the reactor building and documented the condition in IR 2575255. The finding is more than minor because it was associated with the availability and reliability attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loose plastic had the potential to adversely impact the ECCS by compromising the recirculation suction flow path due to blockage of the suction strainer. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," and Appendix A, "The Significance Determination Process for Findings At- Power," Exhibit 2, and determined this finding screened as very low safety significance (Green) because, based on inspector review of a technical debris evaluation (ACIT 4 2578255-08) by Exelon, the finding did not represent an actual loss of function of a system.
05000334/FIN-2017001-01	2017Q1	Beaver Valley	Failure to Follow the ASME OM Code for a Failed Relief Valve Set Pressure Test	Severity Level IV. The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (CFR) 50.55a(z), Alternatives to codes and standards requirements, for FENOCs failure to obtain prior authorization for implementing an alternative to the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). Specifically, until prompted by the inspectors, FENOC did not submit to the NRC and receive an alternative to the ASME OM Code requirement to not test the residual heat removal (RHR) relief valve, RV-1RH-721, during a recent refueling outage for Unit 1 when the charging system letdown relief valve, RV-1CH-203, failed to lift within three percent of set-pressure. FENOCs immediate corrective actions included performing a prompt operability determination, submitting a relief request, and entering the issue into the corrective action program (CAP) as condition report (CR) 2017-03937. The inspectors determined that this violation impacted the ability of the NRC to perform its regulatory oversight function, and was therefore subject to traditional enforcement. Section 2.2.1.c of the Enforcement Policy states that failure to receive prior NRC approval for changes in licensed activities when required is an example of impacting the ability of the NRC to perform its regulatory oversight function. After considering the factors in Section 2.2.1.c of the Enforcement Policy, the inspectors determined that the performance deficiency was a Severity Level IV violation because the change implemented by FENOC would likely be approved by the NRC. Because this violation involves the traditional enforcement process and does not have an associated finding that is more than minor, the inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Appendix B.
05000334/FIN-2017001-02	2017Q1	Beaver Valley	Operability Determinations and Functionality Assessments	10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawing, procedures, and instructions. Contrary to the above, FENOC failed to correctly translate the design basis for protection against tornado-generated missiles into their specifications and procedures. Specifically, FENOC did not adequately protect Unit 1 and Unit 2s main steam safety and atmospheric dump valve exhausts from tornado-generated missiles. Additionally, FENOC did not adequately protect Unit 2s component cooling pumps and spent fuel from tornado-generated missiles by failing to include in their procedures actions for closing the tornado doors in the event of a tornado. The inspectors evaluated FENOCs immediate compensatory measures, which included verifying that procedures are in place and training is current for performing actions in response to a tornado. Because this violation was identified during the discretion period covered by Enforcement Guidance Memorandum 15-002, Revision 1, Enforcement Discretion for Tornado Missile Protection non-compliance (ML16355A286) and because FENOC has implemented compensatory measures, the NRC is exercising enforcement discretion, is not issuing enforcement action, and is allowing continued reactor operation.
05000334/FIN-2016004-01	2016Q4	Beaver Valley	Failure to Follow Procedure Results in an Inoperable A River Water Train	A self-revealing NCV of Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for FENOCs failure to assure that activities affecting quality were accomplished in accordance with procedures. Specifically, FENOC failed to follow NOP-OP-1001, Clearance/Tagging Program, and clearance 1W11-30-MNM-002 when removing the clearance for the A bay of the main intake structure. This resulted in disabling the automatic start capability of the standby C river water pump and made the A river water train inoperable and unavailable. FENOCs immediate corrective action was to rack the breaker for the A river water pump to the disconnect position, which cleared the annunciator and restored operability to the A train of river water. FENOC entered this issue into their corrective action program (CAP) as condition report (CR) 2016-14253. The performance deficiency is more-than-minor because it is associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOC incorrectly racked the A river water pump breaker onto the 1AE 4160 volts alternating current (VAC) safety bus while the C river water pump was already racked onto the bus. This caused the A train of river water to be inoperable and unavailable because the automatic start capability of the C pump was disabled. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, or an actual loss of function of one non-technical specification train designated as high safety significance. This finding has a cross-cutting aspect in Human Performance, Avoid Complacency, because the operators did not plan for the possibility of mistakes and did not implement appropriate error-reduction tools (H.12).
05000334/FIN-2016004-02	2016Q4	Beaver Valley	Licensee-Identified Violation	The following licensee-identified violation of NRC requirements was determined to be of very low safety significance and meets the NRC Enforcement Policy criteria for being dispositioned as a NCV. Radioactive material shipment B-4655, was made from Beaver Valley on May 5, 2016, to ResinSolutions in Erwin, TN. During a self-assessment performed by the FENOC staff on November 3, 2016, it was identified that the scaling factors used to determine the hard-to-detect nuclides listed on the manifest (NRC Form 540) for shipment B-4655 were incorrect. The scaling factors used to manifest the shipment were not for the waste stream shipped. Recalculation of the isotopic values using the correct waste stream scaling factors resulted in different numeric values for multiple radionuclides in the shipment, but did not cause a change in the proper shipping name, packaging, or labeling. 10 CFR 71.5 requires, in part, that radioactive materials be transported with an accurate shipment manifest. Contrary to the above, on May 5, 2016, FENOC transported radioactive materials with a shipment manifest that incorrectly stated that the radiological activity of the package was higher than the actual activity. FENOC documented this issue in CR 2016-13071, and provided a corrected shipment manifest to the recipient of the material. In accordance with IMC 0609, Appendix D, "Public Radiation Safety Significance Determination Process," the finding was determined to be of very low safety significance (Green) because FENOC had an issue involving transportation of radioactive material, but it did not involve a radiation limit that was exceeded, a breach of package during transport, a certificate of compliance issue, a low level burial ground nonconformance, or a failure to make notifications or provide emergency information.
05000289/FIN-2016004-01	2016Q4	Three Mile Island	Licensee-Identified Violation	Technical specification 3.2.12.1, "LTOP Protection", requires when the reactor vessel head is installed and indicated reactor coolant system temperature is 313F, high pressure injection pump breakers shall not be racked in unless injection valves (MU-V16A/B/C/D and MU-V217) are closed with their associated breakers open and that pressurizer level is maintained 100 inches, or restore pressurizer level to 100 inches within 1 hour. Contrary to technical specification 3.2.12.1, during reactor coolant system filling with the vessel head installed and temperature < 313F, high pressure injection pump breakers were racked in while pressurizer level was >100 inches for greater than 1 hour. The condition existed for 2 hours and 49 minutes until recognized by the operating crew when questioned by a senior reactor operator trainee, at which time the crew took immediate actions to reduce pressurizer level <100 inches within 1 hour. Additional corrective actions included crew remediation, additional main control room supervisory oversight, and procedure changes. Exelon entered this issue into the corrective action program as issue report 3949713. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix G, Shutdown Operations, Attachment 1, Exhibit 4, since the finding did not represent an inadvertent safety injection and did not render the power-operated relief valve (LTOP Protection) unavailable or degraded.
05000219/FIN-2016004-01	2016Q4	Oyster Creek	E EMRV Failureto Stroke Due to Incorrect Reassembly	The NRC identified a preliminary White finding and associated apparent violation of Technical Specification 6.8.1, Procedures and Programs, and Technical Specification 3.4.B, Automatic Depressurization System, because Exelon failed to implement a procedure related to the maintenance of safety related equipment. Specifically, Exelon personnel did not follow electromatic relief valve (EMRV) reassembly instructions that required personnel to reinstall previously removed lock washers from the E EMRV cut-out switch lever. The incorrect reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, which led to the E EMRVs failure to perform its safety function. This resulted in one inoperable EMRV for greater than the Technical Specification allowed outage time. The issue was entered into the corrective action program as issue report 2722109, and Exelons immediate corrective actions include installing new cut-out switch lever plates with increased clearances, replacing star lock washers with split ring lock washers for additional clearance, and verifying the five EMRV solenoid actuators being installed into the drywell following the most recent refueling outage were correctly assembled. The finding is more than minor because it adversely affects the human performance quality attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the missing lock washers due to the incorrect EMRV lever plate reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, causing the cut-out switch lever to become bound in the energized position. This led to the E EMRVs failure to perform its safety function. The inspectors screened this issue for safety significance in accordance with Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and determined a detailed risk evaluation was required because the E EMRV had potentially failed or was unreliable for greater than the Technical Specification allowed outage time. A detailed risk evaluation concluded that the increase in core damage frequency (CDF) related to the failure of the E EMRV is 5.4E-6/year; therefore, this finding was preliminary determined to have a low to moderate safety significance (White). Due to the nature of the failure, no recovery credit was assigned. The dominant core damage sequences involve loss of main feedwater events with operator errors resulting in failure to make-up to the 4 isolation condensers or otherwise maintain reactor vessel level and the loss of reactor pressure vessel depressurization capability (due to common cause failure of the remaining four EMRVs). The finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not follow station processes. Specifically, Exelon did not follow written instructions when reassembling the E EMRV. The missing lock washers resulted in excessive friction between the solenoid frame and cut-out switch lever, causing the cut-out switch lever to become bound in the energized position, which led to the E EMRVs failure to perform its safety function. (H.8)
05000334/FIN-2016003-01	2016Q3	Beaver Valley	Failure to Identify Conditions Adverse to Quality Leads to Inoperable Emergency Bus Degraded Voltage Relays	The inspectors identified an NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, for FENOCs failure to assure that a condition adverse to quality was promptly identified and corrected. Specifically, FENOC failed to promptly identify and correct a negative trend in setpoint drift and as found dropout voltage values in the AB 27N model 411T6375HF 4160 volts alternating current (VAC) and 480 VAC emergency bus degraded voltage relays. FENOCs immediate corrective actions included recalibrating or replacing the relays and entering the issue into their corrective action program (CAP) as condition report (CR) 2016-12018. The performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOCs failure to promptly identify and address a negative trend in dropout voltage setpoint drift and as found values resulted in the reduced reliability of safety related bus degraded voltage relays (seven surveillance failures and inoperable degraded bus relays between 2011 and 2016). Inoperable emergency bus degraded voltage relays could lead to damage of safetyrelated equipment during a loss of offsite power. This finding is of very low safety significance (Green) because it does not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, an actual loss of function of one non-technical specification trains designated as high safety significant, and did not involve a loss or degradation of equipment designed to mitigate a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of Problem Identification and Resolution, Trending, because FENOC did not periodically analyze the results of the degraded voltage relay surveillances to provide early indication of a declining trend (P.4).
05000289/FIN-2016003-01	2016Q3	Three Mile Island	Emergency Diesel Generator Internal Flooding Risk Not Evaluated	The inspectors identified an NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelon did not ensure the availability of the emergency diesel generator (EDG) following a seismic event. The inspectors reviewed the TMI licensing basis for internal flooding, associated evaluations and conditions reports, and walked down safety-related structures system and components (SSCs). During this review the inspectors determined that non-seismic piping failures in the EDG room were not properly evaluated. Specifically, the inspectors determined that pressurized fire water pipes in both EDG rooms were not classified as safety-related or seismically qualified. The inspectors reviewed Exelons evaluation of the potential failure of the pipe, as assumed in the TMI design and licensing basis, and determined that operator actions were credited to mitigate the pipe failure in order to prevent water from affecting the operation of the EDGs. The inspectors determined that these operator actions could not be performed prior to water from the pipe break impacting the operation of the EDGs. Following identification of the issue, Exelon entered this issue into their corrective action program and performed an analysis on the structural loading on the fire water piping during a safe shutdown earthquake and concluded that the piping would not break during the design basis event and, therefore, the EDGs remained operable. The inspectors reviewed the analysis and found it reasonable. The inspectors determined the failure to adequately evaluate the effects of a pipe failure in the EDG room in accordance with the design and licensing basis was a performance deficiency. The performance deficiency is considered more than minor because it is associated with the Protection Against External Factors 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 (i.e., core damage). Additionally, the performance deficiency is considered more than minor in accordance with Manual Chapter 0612, Appendix E - Question 3K, in that there was a reasonable doubt of operability for the EDGs requiring engineering calculations and analysis to resolve. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined the finding to be of very low safety significance (Green) because the finding was determined to be a design or qualification deficiency that did not result in an inoperability. No cross-cutting attribute is assigned to this finding because the performance deficiency was not indicative of Exelons current performance. Specifically, this issue was last identified and reviewed by Exelon in issue report 1201424 in 2010.
05000334/FIN-2016002-01	2016Q2	Beaver Valley	Procedure Change Results in Failure to Maintain the Design Basis for the Service Water System	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, for FENOCs failure to assure that the regulatory requirements and design basis for the Unit 2 service water system were correctly translated into procedures. Specifically, FENOC implemented a procedure revision in 2002 that inappropriately removed the step to declare the Unit 2 service water system inoperable while the non-seismic standby service water system is aligned to it. FENOCs immediate corrective actions included issuing instructions that prohibit planned testing of or swapping to the standby service water system and revising procedure 2OST-30.1A. FENOC entered the issue into their CAP as condition report (CR) 2016-01710. The performance deficiency is more-than-minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOCs revision to 2OST-30.1A in 2002 resulted in reduced reliability of the service water system while connected to the standby service water system for over ten hours on February 1, 2016, and nine hours on April 3, 2014. This finding was of very low safety significance (Green) because it did not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, an actual loss of function of one non-technical specification trains designated as high safety significant, and did not involve a loss or degradation of equipment designed to mitigate a seismic, flooding, or severe weather initiating event. This finding does not have a cross-cutting aspect because it is not representative of current performance. The inadequate review of revision 17 to 2OST-30.1A was an isolated instance that occurred over 14 years ago. Furthermore, the most recent NRC inspection of Changes, Tests, or Experiments and Permanent Plant Modifications, performed in 2013, and the Component Design Basis Inspection, performed in 2014 did not document any findings related to procedure changes. (Section 1R15)
05000412/FIN-2016002-02	2016Q2	Beaver Valley	Inadequate Compensatory Measures to Ensure the Effectiveness of an EAL	The inspectors identified an NCV of 10 CFR 50.54(q)(2) for FENOCs failure to follow and maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR 50.47(b)(4). Specifically, following the failure of the area radiation monitor (ARM) for the Unit 2 primary auxiliary building 773 elevation on April 23, 2016, FENOC did not establish adequate compensatory measures to ensure the effectiveness of the emergency action level (EAL) for loss of control of radioactive material, RU2. FENOCs immediate corrective actions included establishing appropriate compensatory measures for RU2, communicating the standards of EAL compensatory measures to radiation protection technicians verbally and via narrative logs, and entering this issue into their CAP as CR 2016-05975. The performance deficiency is more-than-minor because it is associated with the Facilities and Equipment attribute of the Emergency Preparedness cornerstone, and adversely affected the cornerstone objective to ensure that FENOC is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, FENOCs failure to establish adequate compensatory measures for an out-of-service ARM could have resulted in exceeding a NOUE EAL threshold for a loss of control of radioactive material without the condition being recognized until further degradation in the level of plant safety occurs. This finding was determined to be of very low safety significance (Green) since it was example of an ineffective EAL, such that a notification of unusual event (NOUE) would not be declared or would be declared in a degraded manner. This finding has a cross-cutting aspect in Human Performance, Documentation, because FENOC did not ensure that plant activities are governed by comprehensive procedures (H.7).
05000334/FIN-2016002-03	2016Q2	Beaver Valley	Failure to Appropriately Utilize Multiple and Diverse Indications Results in Plant Transient	A self-revealing finding of NOP-OP-1002, Conduct of Operations, was identified for FENOCs failure to adequately implement operator fundamentals. Specifically, operators did not appropriately utilize multiple and diverse indications when making the decision to isolate electro-hydraulic control (EHC) to a Unit 1 main turbine governor valve. This resulted in an unanticipated reactor power reduction of 2.7 percent. FENOCs immediate corrective actions included re-opening the governor valve, verifying proper system response, and entering this issue into their corrective action program (CAP) as CR 2015-08263. The performance deficiency is more-than-minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Additionally, example 4.b from IMC 0612 Appendix E details that a performance deficiency is more-than minor if it causes a reactor trip or other transient. This finding was determined to be of very low safety significance (Green) since it did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding has a cross-cutting aspect in Human Performance, Challenge the Unknown, because individuals did not consult the system expert when confronted with an unexpected condition (H.11).
05000219/FIN-2016002-01	2016Q2	Oyster Creek	Inadequate Maintenance Procedure associated with Reactor Recirculation Pump Seal	A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the reactor recirculation pump (RRP) reassembly maintenance procedures as required by NRC Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, the RRP reassembly procedure, 2400-SMM-3226.03, Reactor Recirculation Pump Mechanical Seal Rebuild Using CAN-2A Parts, did not provide critical dimensional checks for the locking plate and seal adjusting cap. This led to the incorrect reassembly of the D RRP. Exelon entered this issue into their corrective action program as issue report 2663436. The corrective actions included repairing the D RRP and revising RRP maintenance procedures to include critical dimensional information. This finding is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operation. Specifically, the incorrect reassembly of the D RRP created a leakage path, which led to an unexpected increase in reactor coolant system (RCS) unidentified leakage. As a result, the operators inserted a manual scram on April 30, 2016. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined that this finding is a transient initiator that did not contribute to both the likelihood of 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, and therefore was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding since it was not representative of current Exelon performance. Specifically, in accordance with IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and were not considered representative of present performance.
05000289/FIN-2016001-01	2016Q1	Three Mile Island	Deficient Design Control of ECCS Level Transmitter Instrument Line Heat Trace Causes Freezing and Inoperability	A self-revealing NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, was identified for failure to establish and implement adequate design control measures to assure that the borated water storage tank (BWST) was capable of performing its design function to mitigate a design basis loss of coolant accident (LOCA) event. Specifically, Exelon made a modification to the BWST level indicator safety grade heat trace circuit that placed the circuit in an unapproved electrical configuration, which failed to prevent instrument line freezing during cold weather periods, contrary to its safety-function to maintain BWST level indication operable in cold weather. This adversely impacted the availability of a BWST level indication necessary for operators to reliably perform a critical design basis manual action. Exelon documented these issues in issue reports 2609417 and 2611119. Immediate corrective actions included replacement of the affected heat trace and completion of a compatible modification to its electrical configuration. This performance deficiency was more than minor because it was associated with the design control attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding was similar to example 2.f in Appendix E of IMC 0612, in that failure to properly maintain cold weather protection equipment for the BWST level transmitters resulted in DH-LT-809 becoming inoperable. The finding was of very low safety significance (Green) because it did not affect design or qualification, did not represent a loss of system function, did not cause at least one train of BWST level instrumentation to be inoperable for greater than its Technical Specification limiting condition of operation (LCO) allowed outage time, and did not involve external event mitigation systems. The finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because station personnel did not follow the heat trace procedure, which did not allow the two types of heat trace to be spliced together.
05000289/FIN-2016001-02	2016Q1	Three Mile Island	Licensee-Identified Violation	On February 6, 2016, while making preparations to perform procedure 1303-11.45, PORV Setpoint Check, a senior operator identified that the assigned risk for this planned maintenance activity was inaccurate. Specifically, the risk for the maintenance activity was Yellow, not Green, as originally determined. The reason for the inaccurate risk was due to not previously recognizing the pressurizers block valve (RC-RV-2) would be rendered inoperable during the maintenance activity. This condition could result in failure to operate the pressurizers power operated relief valve. The failure to accurately assess the risk of the power operated relief valve setpoint check was a performance deficiency that was within the licensees ability to identify and correct. The inspectors noted that this maintenance activity had an inaccurate risk assessment for at least the past three years. This performance deficiency was a violation 10 CFR Part 50.65(a)(4), which requires, in part, the licensee to assess and manage the increase in risk that may result from the proposed maintenance activity. Contrary to the above, Exelon failed to accurately assess the risk for the power operated relief valve setpoint check over the past three years. The issue was more than minor because it was associated with the configuration control attribute of the initiating systems cornerstone and it adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, screening criteria. The finding screened to Green because the incremental core damage probability of failing to operate RC-RV-2 is less than 1.00 10-6 per year during the short period which the valve is rendered inoperable during each performance of this maintenance activity. Exelon has entered this issue into its corrective action program (issue report 2622859) and revised the risk assigned to this maintenance activity. Because this finding is of very low safety significance and had been entered into Exelons corrective action program, this violation is being treated as a Green, licensee-identified NCV, consistent with section 2.3.2 of the NRCs Enforcement Policy.
05000334/FIN-2016001-01	2016Q1	Beaver Valley	Failure to Properly Evaluate Control Room Envelope Test Results	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XI, Test Control, for FENOCs failure to properly evaluate the test results of the Control Room Envelope (CRE) unfiltered air in-leakage test performed in December 2015. Specifically, the test results exceeded the acceptance criteria specified in the test procedure and required further engineering evaluation to determine if the control room emergency ventilation system (CREVS) could meet its specified safety function. The inspectors identified that the engineering evaluation of the test results did not account for all of the in-leakage and resulted in a reasonable doubt of operability of CREVS. FENOCs immediate corrective action was to re-evaluate the December 2015 calculation and verify that CREVS remained operable with the increased in-leakage. FENOC entered the issue into their corrective action program, condition report (CR) 2016-03836. The performance deficiency is more-than-minor because it is associated with the human performance attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect from radionuclide releases caused by accidents or events. Specifically, FENOCs evaluation did not account for in-leakage from the non-tested portions of the control room radiological barrier, and therefore, did not provide reasonable assurance that the control room dose would not exceed five rem during an uncontrolled release of radioactivity. Additionally, this issue is similar to example 3j and 3k of IMC 0612 Appendix E, Examples of Minor Issues, in that FENOCs December 2015 engineering evaluation failed to adequately account for CRE in-leakage and resulted in a reasonable doubt of the operability of CREVS. The inspectors determined that this finding was of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because FENOC did not take a conservative approach to decision making, particularly when the in-leakage information was incomplete (H.14).
05000219/FIN-2016001-01	2016Q1	Oyster Creek	Failure to Identify a Slower than Normal Scram Time of a Control Rod Drive	The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify and correct a condition adverse to quality. Specifically, Exelon did not identify that the scram time test result for control rod drive 18-47 was beyond the analyzed scram time, which resulted in a degraded control rod drive. Exelon entered this issue into their corrective action program. Immediate corrective actions included fully inserting the control rod drive and developing a casual analysis to determine the degraded condition. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency affected the reliability of control rod drive 18-47 to perform its safety function due to a slower than normal scram time. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), when the SSC maintained its operability or functionality. Therefore, the inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because Exelon did not identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, Exelon did not identify that the actual scram time of control rod drive 18-47 was beyond the analyzed scram time, resulting in a degraded control rod drive.
05000219/FIN-2016001-02	2016Q1	Oyster Creek	Failure to Use Respiratory Protection as Required in RWP/ALARA Plan for Drywell Head Reassembly	A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs was identified for Exelons failure to use respiratory protection, as required in the radiation work permit (RWP)/as low as reasonably achievable (ALARA) plan 14-406 for drywell head reassembly work on October 2, 2014. The radiation protection (RP) supervisor overseeing this work removed the respiratory protection requirement for this work contrary to the RWP/ALARA requirement and without engineering approval. As a result, two workers received an unplanned intake of radioactive material that resulted in unintended internal dose. Upon identification of the intake, Exelon stopped work on this task and subsequently reinstituted the respiratory protection requirements to complete the remaining work and entered this event into their corrective action program as issue report 2390111. This finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone to ensure adequate protection of the worker from radiation exposure. Specifically, without the use of respiratory protection two workers received unintended internal dose. The inspectors evaluated the finding using inspection manual chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that this finding is of very low safety significance (Green), because it did not result in an overexposure as defined by 10 CFR 20.1201, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. This finding has a cross-cutting aspect in Human Performance, Procedural Adherence, because Exelon did not follow procedures and work instructions. Specifically, RP supervision instructed the workers that respiratory protection was not required contrary to the applicable RWP/ALARA plan.
05000219/FIN-2016001-04	2016Q1	Oyster Creek	Licensee-Identified Violation	From 2010 to 2014, Oyster Creek made a total of four shipments of radioactive material which contained category 2 quantities of radioactive material. Oyster Creek did not implement a transportation security plan for any of these shipments, which is contrary to the requirements of 49 CFR 172, Subpart I, Safety and Security Plans. This performance deficiency adversely affected the Public Radiation Safety cornerstone attribute of Program and Process based on inadequate procedures associated with the transportation of radioactive materials. The finding was determined to be of very low safety significance (Green) because the transportation of radioactive material issue 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. This issue was documented in the Exelons corrective action program as IR 2484646. Corrective actions included contracting with a vendor to receive regular, prompt notifications of potentially applicable rule changes in the Federal Register.
05000219/FIN-2016001-03	2016Q1	Oyster Creek	Inadequate Instructions for the Flexible Coupling Hose Preventative Maintenance Resulting in an Inoperable Emergency Diesel Generator	The inspectors identified a preliminary White finding and associated apparent violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Exelon did not appropriately prescribe instructions or procedures for maintenance on the emergency diesel generator (EDG) No. 1 cooling water system to ensure the EDG cooling flexible coupling hose was maintained to support the EDG safety function. Specifically, Exelon did not have appropriate work instructions to replace the EDG cooling flexible coupling hoses every 12 years as specified by Exelons procedure and vendor information. As a result, the flexible coupling hose was in service for approximately 22 years and subjected to thermal degradation and aging that eventually led to the failure of EDG No. 1 during operation on January 4, 2016. As a consequence of this inappropriate work instruction issue, Exelon violated Technical Specification 3.7.C because EDG No. 1 was determined to be inoperable for greater than the technical specification allowed outage time of seven days. Exelons immediate corrective actions included entering the issue into their corrective action program (issue reports 2607247 and 2610027), replacing of the EDG No. 1 and No. 2 flexible coupling hoses, and initiating a failure analysis to determine the causes of the failed flexible coupling hose. This finding is more than minor because it is associated with the equipment performance 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, the ruptured flexible coupling hose caused the failure of EDG No. 1 to perform its safety function. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, this finding required a detailed risk evaluation (DRE) because EDG No. 1 was inoperable for greater than the technical specification allowed outage time. The DRE estimated the increase in core damage frequency was 7E-6, or White (low to moderate safety significance) for this finding. This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance.
05000219/FIN-2015004-01	2015Q4	Oyster Creek	Preconditioning of the Standby Liquid Control Relief Valves	The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XI, Test Control, because Exelon conducted unacceptable preconditioning of the standby liquid control (SLC) relief valves prior to American Society of Mechanical Engineers (ASME) code testing. Specifically, Exelon performed a SLC system functional test prior to performing the SLC relief valve as-found testing. Exelons immediate corrective actions included completing the as-found test prior to the functional test. Exelon entered this issue into their corrective action program (CAP) as issue report 2566036 to track the resolution of the issue. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, if left uncorrected, the performance deficiency could have the potential to lead to a more significant safety concern. Specifically, completion of the functional test prior to the replacement of the SLC relief valves masks the actual as-found condition by solidifying the valve internals. As a result, the as-found condition of the SLC relief valves have not been conducted and in the worst case scenario, could open below the design setpoint, which would divert flow back to the liquid poison tank instead of into the vessel to shut down the reactor during an anticipated transient without scram (ATWS) condition. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because the structure, system or component (SSC) maintained its operability. 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 did not evaluate the effect of performing the SLC system functional test prior to conducting the ASME code as-found test on the SLC relief valves.
05000219/FIN-2015004-02	2015Q4	Oyster Creek	Inadequate Problem Identification and Resolution Leading to Degradation of EPR Causing a Reactor Scram	A self-revealing finding was identified because Exelon did not adequately identify and correct conditions, per LS-AA-120, Issue Identification and Screening Process, that led to degradation of the electric pressure regulator (EPR) wiring, which resulted in an uncontrolled rise in reactor pressure and subsequent reactor scram on average power range monitor (APRM) Hi-Hi Flux. Specifically, Exelon failed to generate issue reports to document degraded EPR wiring that was previously identified, and therefore did not take corrective action prior to a reactor scram. Planned corrective actions include reinforcing with station personnel that an issue report is required when issues are identified. This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely impacted its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. In accordance with IMC 0609, Attachment 4 and Exhibit 1 of Appendix A, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not cause both 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 there is no cross-cutting aspect associated with this finding since it is not representative of current Exelon performance. Specifically, in accordance IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and considered not representative of present performance.
05000289/FIN-2015004-01	2015Q4	Three Mile Island	Failure to Trend Vibration Data for Safety Related River Water Pump	The inspectors identified a finding of very low safety significance involving an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B Criterion XVI, Corrective Action Program, because Exelon did not identify and correct a condition adverse to quality on the B nuclear river water pump (NR-P-1B). Specifically, Exelon did not properly evaluate an adverse vibration trend on NR-P-1B, which resulted in exceeding its in-service test (IST) required action level and declared inoperable on October 10, 2015. Exelon entered the condition into their corrective action program (CAP) as issue report 2568763 and emergently replaced the pump, engaged the vendor for short and long term design and material changes to correct the vibration, and created process and peer check corrective actions to ensure all vibration data is reviewed timely and trends are addressed commensurate with their safety significance. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the elevated vibrations reduced the reliability and capability of NR-P-1B to perform its safety function. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, and determined this finding to be of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification (TS) allowed outage time and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the station did not thoroughly evaluate the elevated vibration data such that the issue was addressed before NR-P-1B became inoperable (P.2).
05000334/FIN-2015004-01	2015Q4	Beaver Valley	Inadequate Maintenance Rule Monitoring of the Auxiliary Feedwater System	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR) 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, for FENOCs failure to monitor the performance of the Unit 1 auxiliary feedwater (AFW) system against licensee-established goals. Specifically, FENOC did not identify and properly account for a maintenance preventable functional failure (MPFF) of the turbine driven auxiliary feedwater (TDAFW) pump, which demonstrated that performance of the Unit 1 AFW system was not being effectively controlled through appropriate preventive maintenance. FENOCs immediate corrective actions included entering this issue into their corrective action program, re-evaluating and classifying the TDAFW pump failure as a MPFF, performing a 10 CFR 50.65 (a)(1) evaluation of the Unit 1 AFW system, and placing the system in (a)(1) status. The performance deficiency was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, example 7.d from IMC 0612 Appendix E details that a performance deficiency is more than minor if equipment performance problems were such that effective control of performance through appropriate preventive maintenance under (a)(2) could not be demonstrated. This finding was determined to be of very low safety significance (Green) since it was not a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), it did not represent the loss of a system and/or function, it did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its technical specifications allowed outage time, and it did not represent an actual loss of a non-technical specification equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in Human Performance, Avoid Complacency, because FENOC failed to consider the extent of condition and their causes following the failure of the Unit 1 TDAFW pump on January 6, 2014 (H.12).
05000289/FIN-2015003-01	2015Q3	Three Mile Island	Internal Flooding Licensing Basis Commitment Not Met	The inspectors identified a finding because Exelon failed to meet a commitment made during original licensing to mitigate an internal flooding event. Specifically, Exelon committed to making changes to the fire water supply system to mitigate the impact of a pipe rupture in the auxiliary building. The inspectors identified that the commitment actions were not completed and no changes to the commitment were identified. The inspectors determined that the failure to perform the modifications to the fire service system, as committed to the NRC in a letter dated November 10, 1972, was a performance deficiency that was reasonably within its ability to foresee and correct. Exelon documented the issue in issue report 2544387, performed an immediate operability evaluation, and developed corrective actions to restore compliance with the commitment. The inspectors determined that the performance deficiency is associated with the Mitigating Systems cornerstone attribute of protection against external factors (internal flood hazard) and is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency adversely impacted the operators ability to detect and mitigate a fire service system pipe rupture in the safety related auxiliary building. The inspectors utilized IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, to determine the significance of the performance deficiency. The inspectors determined the finding to be of very low safety significance (Green) because the finding is not a design or qualification deficiency, does not represent a loss of system safety function or loss of a single train for greater than its allowed technical specification time, does not result in the loss of a high safety-significant maintenance rule train and does not involve the loss of function to mitigate internal flooding events. The finding is not assigned a cross-cutting aspect because the performance deficiency occurred during original plant construction and is not indicative of current plant performance.
05000219/FIN-2015003-01	2015Q3	Oyster Creek	Non-Conservative Temperature Input in the Electromatic Relief Valve Voltage Drop Calculation	The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelons measures for verifying the adequacy of design of the electromatic relief valve (EMRV) voltage drop calculation were inadequate. Specifically, non-conservative temperature inputs were used for the safety related EMRV direct current voltage drop calculation, which reduced the margin of available voltage to the EMRV solenoids. Exelon entered this issue into the corrective action program for resolution as issue report 2522756, and corrective actions included revising the calculation to include the correct temperature values and conduct an extent of condition of other voltage drop calculations that could have similar temperature values. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, lower voltage to the EMRV solenoid at higher temperatures could affect the reliability and capability of the EMRV to perform its design function. In addition, the performance deficiency is determined to be more than minor because it is similar to example 3.j of NRC IMC 0612, Appendix E, Example of Minor Issues, in that as a result of the calculation errors and the magnitude of the decrease of margin, there was a reasonable doubt on the operability of the component. The inspectors evaluated the finding using 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding is not assigned a cross-cutting aspect because it is not reflective of current performance. Specifically, the last time Exelon had an opportunity to evaluate this issue was in 2010 when Exelon identified that the EMRV solenoid voltage had low margin.
05000334/FIN-2015003-01	2015Q3	Beaver Valley	Failure to Correct a Low Oil Level in the Condensate Pump Motor	A self-revealing finding was identified for FENOCs failure to correct a low oil level in the lower motor bearing of the Unit 1 A condensate pump in accordance with NOP-LP- 2001, Corrective Action Program. Specifically, FENOC incorrectly cancelled the work order to add oil to the A condensate pump motor and installed a placard on the oil level sight glass with incorrect minimum and maximum oil levels. This led to the motor bearing failure, which caused the pump to trip on overcurrent, and required the operators to insert a manual reactor trip. FENOC entered the issue into their correct action program, condition report (CR) 2015-05256. The performance deficiency was more-than-minor because it was associated with the human performance attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, NOP-LP-2001, section 4.2.3, states that condition report/correct action owners should ensure that actions are developed to resolve the primary cause identified in the condition report. Instead of correcting the low oil level in the motor, FENOC cancelled the work order to add oil. This subsequently caused the operators to trip the plant when the condensate pump motor bearing overheated and the motor tripped on overcurrent. The inspectors determined that this finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. This finding has a crosscutting aspect in the area of Human Performance, Consistent Process, because FENOC did not seek input from the appropriate work group (engineering) prior to cancelling the work order to add oil to the condensate pump motor (H.13)
05000334/FIN-2015008-01	2015Q2	Beaver Valley	Failure to Initiate a Condition Report for an Adverse Condition	A Green self-revealing finding of NOP-LP-2001, Corrective Action Program, was identified after FENOC failed to generate a condition report for a condition adverse to quality. Specifically, FENOC did not initiate a condition report when a lifted lead was identified during preventative maintenance and installation of the Unit 1 main transformer. As a result, corrective actions were not taken and this led to an unplanned downpower from 100 percent to 15 percent reactor power on January 31, 2014. The performance deficiency was more-than-minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. This finding was determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Field Presence, because FENOC failed to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel (H.2).
05000334/FIN-2015002-01	2015Q2	Beaver Valley	Failure to Utilize Respiratory Protection as Specified by the Radiation Work Permi	The inspectors identified a self-revealing NCV of Technical Specification 5.4.1, Procedures, for FENOCs failure to utilize respiratory protection, as required by the applicable radiation work permit (RWP), for entry into the 722-foot elevation of the solid radioactive waste building on March 12, 2014. This resulted in the unplanned internal exposure of one worker. Immediate corrective actions included reestablishing RWP controls of the area and entering this issue into their corrective action program as condition report 2015-06636. The inspectors determined that the performance deficiency is more than minor because it affected the Program and Process attribute of the Occupational Radiation Safety cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors evaluated the finding using NRC Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined the finding to be of very low safety significance (Green) because it was not related to as low as (is) reasonably achievable (ALARA), did not result in an overexposure or a substantial potential for overexposure, and did not compromise the licensee's ability to assess dose. The finding has a cross-cutting aspect of Human Performance, Conservative Bias, in that individuals did not use decision making-practices that emphasized prudent choices over those that are simply allowable. Specifically, a radiation protection technician did not use conservative decision making practices and make prudent choices when entering an area with unknown radiological conditions. Examples of non-conservative decision making included: failure to wear respiratory protection when entering into unknown radiological conditions, the failure to complete and evaluate an air sample prior to entry, and not taking into account the adverse radiological conditions of the adjoining area above (735 foot elevation). (H.14)
05000412/FIN-2015002-02	2015Q2	Beaver Valley	Failure to Perform Maintenance in accordance with Licensee Maintenance Process	A self-revealing finding was identified for FENOCs failure to perform maintenance on the Unit 2 feedwater heater drain system in accordance with FENOCs maintenance process, NOP-WM-4006, Conduct of Maintenance. Specifically, FENOC did not adjust the A first point feedwater heater normal and high level control valve (LCV) controllers to their specified setpoints. As a result, the A heater and separator drain pumps tripped and this led to an unplanned power reduction from 100 percent to 60 percent reactor power on April 12, 2015. FENOCs corrective action included adjusting the setpoints of the LCV controllers to their specified setpoints and entering the issue into their corrective action program as condition report 2015-05088. The performance deficiency was more-than-minor because it was associated with the Configuration Control attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, the performance deficiency was similar to example 4.b in IMC 0612 Appendix E, in that failing to follow procedure caused a reactor transient. This finding was determined to be of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Training, because FENOC failed to ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, FENOC did not ensure that knowledge was adequate to perform maintenance on the A first point feedwater heater LCVs (H.9).
05000289/FIN-2015002-01	2015Q2	Three Mile Island	Failure to Maintain Turbine Bypass Valve Simulator Modeling	A self-revealing NCV of 10 CFR Part 55.46(c), Plant-Referenced Simulators, was identified for Exelons failure to ensure that the plant-referenced simulator demonstrated expected plant response to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, Exelon failed to ensure simulator modeling of once through steam generator (OTSG) turbine bypass valve (TBV) operation was consistent with the actual plant which introduced negative operator training and challenged orderly unit shutdown on May 7, 2015. The licensee documented their corrective actions for this issue in TMI issue reports (IR) 02496279 and 2497542, which included software changes to the simulator to reflect actual system design, crew remediation, and procedure changes. The performance deficiency is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the simulator difference introduced negative operator training and, as a result, challenged orderly shutdown of the unit on May 7, 2015. The inspectors evaluated the finding in accordance with NRC Manual Chapter 0609, Significance Determination Process, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process. The finding was determined to have very low safety significance (Green) because the impact on operator performance was not during a reportable event. This finding has no cross-cutting aspect assigned because the cause was not representative of current licensee performance. Specifically, the difference in TBV modeling existed since initial simulator certification on June 28, 1990.
05000219/FIN-2015002-01	2015Q2	Oyster Creek	Inadequate Assessment of 4k Emergency Switchgear Roll-Up Door Degraded Floor Gasket	The inspectors identified a finding associated with Exelon procedure, OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess a degraded floor gasket for the D emergency 4 kilovolt (kV) switchgear roll-up door. Specifically, Exelon did not adequately assess the flood and fire functionality of the degraded gasket, which is credited to provide protection to safety-related D emergency 4kV switchgear during a postulated internal flood event and to contain the carbon dioxide (CO2) gaseous suppression system during a postulated fire within the D switchgear room. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing the operability determination procedure and enhancing operator training in fire and flood functionality of gaskets. Additional corrective actions included repairing the gasket and performing a detailed analysis of the ability of degraded gasket to meet its flooding and fire function. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded floor gasket could have resulted in increased water level in the D emergency 4kV switchgear room during a postulated internal flood due to a fire water pipe rupture, therefore affecting the reliability of the D emergency 4k switchgear to perform its safety function. In addition, the degraded floor gasket could have resulted in CO2 leakage out of the D emergency 4k switchgear room during a postulated fire in that room, therefore affecting the reliability of the D emergency 4k switchgear gaseous suppression system to perform its safety function. The inspectors determined that this finding is of very low safety significance (Green) because it is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. This 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 the causes and extent of conditions commensurate with their safety significance. Specifically, Exelon staff did not thoroughly evaluate the issue associated with the degraded floor gasket for fire and flood functionality.
05000219/FIN-2015002-02	2015Q2	Oyster Creek	Failure Rates Exceed Twenty Percent for Annual Requalification Exam	A self-revealing finding was identified associated with inadequate licensed operator performance during licensed operator requalification exams in accordance with TQ-AA-150, Operator Training Program. Specifically, two of seven crews failed the simulator scenario portion of the requalification examinations. As an immediate corrective action, the crews that failed were restricted from licensed duties. Exelon entered this issue into the corrective action program, and facility training staff remediated the crews (the crews were retrained and successfully retested), and those crews were returned to licensed duties. This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, two of seven crews failed to demonstrate a satisfactory understanding of the knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors determined the finding to be of very low safety significance (Green) because it is related to requalification exam results, did not result in a failure rate of greater than forty percent, and the two crews were remediated (i.e., the crews were retrained and successfully retested) prior to returning to shift. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Exelon staff did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce.
05000219/FIN-2015002-03	2015Q2	Oyster Creek	Reactor Water Cleanup Procedure Not Followed Resulting in a Level Transient	A self-revealing NCV of Technical Specification 6.8.1(a), Procedures and Programs, was identified because Exelon did not follow procedure 303, Reactor Cleanup Demineralizer System, during the system restoration on March 26, 2015. Specifically, during startup from a forced outage (1F36), Exelon did not follow procedure 303, which required correct valve lineups for system restoration of reactor water cleanup (RWCU) after system isolation. This resulted in decreasing reactor water level, which was automatically terminated by a second RWCU isolation. Exelon entered this issue into the corrective action program. Planned corrective actions include enhancing operator training in system knowledge and procedure compliance and revising startup procedures. This finding is determined to be more than minor because it is associated with the human performance 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, Exelon did not properly lineup the RWCU system after isolation, which resulted in a water level transient and challenging the critical safety function of inventory control. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, the operators did not stop and fully communicate plant condition after the initial RWCU isolation. Consequently, operators opened the RWCU system inlet valve due to the increasing water level without following procedure guidance.
05000219/FIN-2015002-04	2015Q2	Oyster Creek	Reset of the Automatic Voltage Regulator Controller Led to an Automatic Reactor Scram	A self-revealing finding was identified because Exelon did not properly screen work in accordance with MA-AA-716-010, Maintenance Planning. Specifically, on September 12, 2014, Exelon did not screen the automatic voltage regulators (AVR) human machine interface (HMI) post-maintenance test per the maintenance planning procedure. As a result, on October 12, 2014, Exelon personnel performing the post-maintenance test did not have a work order, which would have included plant configurations and limitations associated with the test. This led to an automatic reactor scram. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing with work planners that a work order is required for similar work activities. This finding was determined to be more than minor because it is associated with the human performance 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 plant operation. Specifically, resetting the three AVR controllers caused an automatic plant scram. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, on October 12, 2014, Exelon personnel did not stop when faced with the uncertain situation of the HMI screen that did not respond as expected.
05000289/FIN-2015001-01	2015Q1	Three Mile Island	Licensee-Identified Violation	LER 05000289/2014-001-00 describes an unanalyzed condition in which Exelon identified DC motor control circuits were unfused. Specifically, Exelon did not provide overcurrent protection for wiring associated with 250VDC full-voltage control circuits for four non-safety emergency bearing oil pumps in the turbine building to prevent wires from overheating due to fire-induced faults and excessive currents flowing through the cable. With enough current flowing through the cable, the potential exists that the overloaded motor control wiring could damage adjacent control circuit wiring for both instrument air compressors (IA-P-1A/B), which are needed to achieve and maintain post-fire safe shutdown for a fire in the cable spreading room. This condition could result in a loss of the associated safe shutdown components or a secondary fire in another fire area. The failure to protect safe shutdown cables from the effect of postulated fires was a performance deficiency. This performance deficiency was a violation of TMI Operating License Condition 2.C.(4), which requires, in part, post-fire safe shutdown cables remain free of the effects of fire-induced cable faults during postulated fires. Contrary to the above, Exelon identified they failed to meet this requirement and the condition existed since initial construction. The issue was more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase 2 screening criteria. The finding screened to Green based upon task number 2.3.5, and because no credible fire ignition source was determined to adversely affect the motor control circuits of concern as determined. Additionally, a fire area of concern (cable spreading area) is an alternate shutdown fire area protected by detection and an automatic suppression system. The cables in the other fire area of concern (turbine building) are Institute of Electrical and Electronics Engineers 383 (thermoset) construction with steel armor and tied to station ground which decreases the likelihood of inter-cable and intra-cable interactions. Because this finding is of very low safety significance and had been entered into Exelons corrective action program (IRs 1651702, 1658837, 1658842), this violation is being treated as a Green, licensee-identified NCV consistent with the NRCs Enforcement Policy.
05000219/FIN-2015001-01	2015Q1	Oyster Creek	Post Maintenance Test Results Were Not Evaluated to Assure that Technical Specifications Requirements Were Satisfied	The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Exelon did not document and adequately evaluate test results to assure that test requirements had been satisfied. Specifically, Exelon did not perform the proper post maintenance test procedure to assure that the requirements of Technical Specification 4.5.G.3 were satisfied following installation of a temporary modification to secondary containment. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 2440643. Corrective actions include revising the process to perform the correct post maintenance test to ensure Technical Specification 4.5.G.3 is met. This finding is more than minor because it is associated with the configuration control (Standby Gas Trains) attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process: Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014. Because the finding degraded the ability to close or isolate secondary containment, the inspectors were required to further assess the finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, issued May 6, 2004. The inspectors determined that this finding is of very low safety significance (Green) because the decay heat values were low, given that the unit had been shut down for approximately three days, and reactor water level was greater than that required for movement of irradiated fuel assemblies within the reactor pressure vessel. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not perform the post maintenance test specified by the work order.
05000219/FIN-2015001-02	2015Q1	Oyster Creek	Inadequate Post Maintenance Testing for Emergency Service Water Pump Breaker	The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for Exelons failure to develop an adequate post maintenance test to determine operability of the A emergency service water pump breaker. Specifically, the corrective maintenance work performed on April 16, 2013, did not correct the cause of the failure and Exelon did not perform an adequate post maintenance test to verify conditions had been corrected. As a result, the emergency service water system was returned to service even though it did not meet all the requirements for operability. The issue was not identified and resolved until a subsequent surveillance test on April 17, 2013, which identified a failed breaker. Exelon entered this issue into their corrective action program (IR 2471069). Planned corrective actions include revising work order activities to specify the correct post maintenance test. This performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected its objective to ensure the availability and reliability of the systems that respond to initiating events. Specifically, the inadequate post maintenance test for A emergency service water pump breaker on April 16, 2013, led to the A emergency service water pump failing to perform its function during the subsequent surveillance testing on April 17, 2013. The inspectors assessed this finding in accordance with the IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors concluded that this finding did not represent an actual loss of function of the emergency service water system for greater than its technical specification allowed outage time (15 days). Therefore, the inspectors determined that this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Work Management, in that Exelons work planning and executing of work activities did not include documented instructions for performing an adequate post maintenance test.
05000219/FIN-2015001-04	2015Q1	Oyster Creek	Licensee-Identified Violation	Technical Specification 3.5.B, Secondary Containment, requires in part, that secondary containment integrity be maintained at all times when the reactor vessel head and the drywell head are in not in place. Technical Specification 1.14, Secondary Containment Integrity, requires in part, that the standby gas treatment system is operable. Technical Specification 4.5.G.3 specifies that with the trunnion room door open and the trunnion room is isolated from secondary containment in support of outage activities, testing of the standby gas treatment system to be performed to demonstrate the capability to maintain 14 inch of water vacuum under calm wind conditions and a standby gas treatment system filter train flow rate of not more than 4000 cfm. Contrary to Technical Specification 3.5.B, on September 20, 2014, with the reactor vessel head and drywell head removed for the refueling outage, Exelon determined that they did not have secondary containment integrity when performing testing to demonstrate standby gas treatment system capability in accordance with Technical Specification 4.5.G.3 and subsequently found that the outer railroad air lock personnel access hatch had not been closed properly, which prevented a proper vacuum from being achieved. Exelon entered this issue into the corrective action program as IR 2383852. Using guidance in IMC 0609, Appendices G and H, the inspectors determined that this finding was of very low safety significance (Green) because the decay heat values were low and the reactor water level inventory was above that required to move irradiated fuel.
05000219/FIN-2015001-05	2015Q1	Oyster Creek	Licensee-Identified Violation	Technical Specification 6.8.1b states that, Written procedures shall be established, implemented, and maintained covering surveillance and test activities of equipment that affects nuclear safety and radioactive waste management equipment. Contrary to the above, from August 2012 through September 2013, Exelon took no action following receipt of ten lubricating oil analysis report results taken from two emergency diesel generator No. 2 sample locations which indicated silver content at 1.0 ppm, which exceeded procedural action levels. Specifically, Exelon maintenance procedure MA-AA-716-230-1001, Oil Analysis Interpretation Guideline, Section 3 governs safety system oil analyses and describes actions to be taken when equipment wear metals exceed specific thresholds, as obtained through monthly oil analysis. Section 3 of procedure MA-AA-716-230-1001 lists potential actions to be taken when oil analysis results indicate silver content above 0.3 and 0.7 ppm respectively. These actions include resampling immediately to verify abnormal results, performing confirmatory testing using more accurate methods if required, reviewing all vibration and thermography data immediately for adverse trends, and contacting the equipment manufacturer for additional assistance. Exelon identified this issue on October 21, 2013, during the performance of a 24-month lubricating oil system inspection on the emergency diesel generator No. 2 when silver metal shavings were found in the main lubricating oil filter housing and in the sump below cylinder #15. The inspectors determined that the failure to identify an out-of-specification lubricating oil sample result on numerous occasions was a performance deficiency that was within Exelons ability to foresee and correct. The inspectors determined that the issue adversely impacted the reliability of the safety-related emergency diesel generator in that the wrist pin bearing was degraded and had partially failed. The inspectors determined that the issue was of very low safety significance (Green) because it did not: affect design or qualification; represent a loss of system or function; exceed technical speciation allowed outage times; and involve external events. Exelon entered this issue into the corrective action program as IR 1575045.
05000219/FIN-2015001-03	2015Q1	Oyster Creek	Incomplete 50.72 and 50.73 Reports Associated with Secondary Containment Integrity	The inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a) in that Exelon did not provide complete information in reports submitted per 10 CFR 50.72 and 10 CFR 50.73. Specifically, a licensee event report (LER) submitted on November 18, 2014, did not discuss a separate, partially opened secondary containment door that was discovered during the same time frame, which could have prevented the fulfillment of the safety function of secondary containment, and therefore was required to be discussed in the original LER. Exelon entered this issue into their corrective action program as IR 2440641. Planned corrective actions include revising the original LER to add a discussion of the partially opened secondary containment door. The inspectors determined that not providing a complete report in accordance with 10 CFR 50.9(a) is a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented. Because the issue had the potential to affect the NRCs ability to perform its regulatory oversight function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. In accordance with Section 2.2.2.d of the NRC Enforcement Policy, the inspectors determined that the performance deficiency identified with the reporting aspect of the event is a Severity Level IV violation because it is of more than minor concern with relatively inappreciable potential safety significance and is related to findings that were determined to be more than minor issues. In accordance with IMC 0612, Appendix B, this issue was not assigned a cross-cutting aspect.
05000219/FIN-2014005-02	2014Q4	Oyster Creek	Inadequate Review of Change in Maintenance Process Results in Inoperable Emergency Diesel Generator	The inspectors identified a preliminary White finding and an associated apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control, because Exelon staff did not review the suitability of the application of a different maintenance process at Oyster Creek that was essential to a safety-related function of the emergency diesel generators (EDG). Specifically, in May 2005, Exelon staff changed the method for tensioning the cooling fan belt on the EDG from measuring belt deflection to belt frequency and did not verify the adequacy of the acceptance criteria stated for the new method. As a result, Exelon staff did not identify that the specified belt frequency imposed a stress above the fatigue endurance limit of the shaft material, making the EDG cooling fan shaft susceptible to fatigue and subsequent failure on July 28, 2014. As a consequence, Exelon also violated Technical Specification 3.7.C, because the EDG No. 2 was determined to be inoperable for greater than the technical specification allowed outage time. Exelons immediate corrective actions included entering the issue into their corrective action program as issue report (IR) 1686101, replacing the EDG No. 2 fan shaft, examining the EDG No.1 fan shaft for extent of condition, and performing a failure analysis to determine the causes of the broken shaft. This finding is more than minor because it is associated with the equipment performance 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). In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors screened the finding for safety significance and determined that a detailed risk evaluation was required because the finding represented an actual loss of function of a single train for greater than its technical specification allowed outage time. The detailed risk evaluation concluded that the increase in core damage frequency was 5.1E-6, or White (low to moderate safety significance). This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance.

Semantic search

Query log

Navigation menu