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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000249/FIN-2018003-01 | 2018Q3 | Green | H.7 | Self-revealing | Failure to Follow Maintenance Procedures for Assembling Unit 3 HPCI Room Cooler Fan | A self-revealing, Green non-cited violation (NCV) of Technical Specification (TS) 5.4, Procedures, was identified for the licensees failure to follow maintenance procedures DMP 570004, LPCI and HPCI Room Cooler Maintenance, and DEP 570004, HPCI Room Cooler Fan Preventive Maintenance, when assembling the Unit 3 HPCI room fan. Specifically, on one occasion when maintenance was performed on the fan, technicians installed the cam locking collar in the opposite direction of the fan shaft rotation, and on the other occasion, technicians tensioned the fan belt to the wrong value and misadjusted the alignment of the shaft sheave. Over time, this improper maintenance caused the inboard and outboard fan bearings to wear on the shaft, causing increased vibrations, and eventually leading to HPCI being declared inoperable to emergently work on the fan |
| 05000237/FIN-2018003-02 | 2018Q3 | Severity level IV | Licensee-identified | Licensee-Identified Violation | Violation: Dresden Technical Requirements Manual (TRM) Control Program (Appendix G of TRM), Section 1.5, Program Implementation, requires that proposed changes to the TRM are screened and reviewed under the 10 CFR 50.59 process in accordance with plant specific procedures. Contrary to the above, in October 2017 Dresden station approved and implemented an extension to the surveillance frequency of DIS 150020, Division I & II Low Pressure Coolant Injection (LPCI) Pumps Suction and Injection Valves Circuitry Logic System Functional Test, on Unit 2 per the Surveillance Frequency Control Program (SFCP) without the required 50.59 review. | |
| 05000237/FIN-2018001-01 | 2018Q1 | Severity level Enforcement Discretion | NRC identified | Enforcement Action: EA18016: Unanalyzed Condition for Tornado Missiles | On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The discretion applied to Technical Specification (TS) limiting condition for operations (LCOs) that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time due to structures, system, and components (SSCs) declared inoperable because of tornado generated missile issues. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Dresden Station, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed the licensee to re-establish operability when the licensee implemented, prior to the expiration of the time mandated by the affected LCOs, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened followed by more comprehensive compensatory measures within 60 days of issue discovery. The enforcement discretion was also conditional to the comprehensive measures remaining in place until permanent repairs are completed or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Section 3.5 of the Dresden Power Station Updated Final Safety Analysis Report (UFSAR) states in part that SSCs important to safety shall be adequately protected against missiles generated by various causes, including natural phenomena. On February 12, 2018, the licensee initiated IR 04103159, identifying a nonconforming condition of Section 3.5. Specifically, the vent lines for the U2, U2/3, and U3 emergency diesel generator (EDG) fuel oil tanks were not adequately protected from tornado-generated missiles. The licensee declared fuel oil tanks and their associated EDGs inoperable, and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53204 as an unanalyzed condition and potential loss of safety function. Corrective Action(s): The licensee documented the inoperability of the SSCs in the Corrective Action Program (CAP) and in the control room operating log. In addition, the affected TS LCO conditions applicable in the mode of operation at the time of discovery were documented in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: Verifying that procedures were in place and training was current for performing actions in response to a tornado event. Verifying that procedures were in place and training was current to respond to a tornado watch, such as: (1) actions to be taken relating to tornado missile hazards; (2) potential restoration of equipment important to maintaining safe shutdown conditions that is unavailable at the time of the tornado watch; (3) warning and protection strategies for personnel; and (4) damage assessment and restorative actions for equipment that may be damaged during a tornado. Establishing a heightened level of station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify DOA001002, Tornado Warning Severe Winds procedure to include actions for damage assessment and restorative actions for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: IR 04103159 | |
| 05000249/FIN-2018012-01 | 2018Q1 | Green | NRC identified | Failure to Ensure that Thermal Overload Relays are Sized Properly for Throttling Motor Operated Valves | The team identified a finding having very-low significance and an associated Non-Cited Violation of Title 10 of the Code of Federal Regulations,Part 50, Appendix B, Criterion III, Design Control.Specifically, Dresden had not verified that thermal overload relays on Unit 3 safety-related motor operated valves 3-1301-3, 3-1501-21A & 21B, 3-1501-18A & 18B, 3-1501-38A & 38B, 3-3-2301-10, 3-1501-3A & 3B, were properly sized to support the design function of repetitive jogging and throttling the valves in response to design basis transients or accidents. | |
| 05000237/FIN-2017004-01 | 2017Q4 | Green | H.9 | NRC identified | Failure to Follow Procedure,Results in Non-Functional Fire Door | The inspectors identified a finding of very-low safety significance and associated NCV of Technical Specification 5.4.1.c for the licensees failure to implement the established Fire Protection Program procedures which ensure Fire Barrier Integrity. Specifically, the licensee ran an electrical cable through the doorway of an automatically closing fire door. This was contrary to Procedure DFPP 417501, which requires in part that fire doors must not be blocked open by props or any other material in its closing path. The licensee took immediate actions to restore the fire door, by removing the obstruction and entered the issue into their Corrective Action Program (CAP). The inspectors determined that the performance deficiency was more-than-minor because it affected the Mitigating Systems cornerstone objective since the electrical cable could have prevented the fire door from performing its function. The finding was of very-low safety significance per Task 1.4.3A of IMC 0609, Appendix F. Specifically, the total combustible loading on both sides of the affected fire door was representative of a fire duration less than 1.5 hours. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, associated with the Training component, because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee believed the performance deficiency was caused by the one of the new temporary contractors brought onto the site to work in support of the D2R25 refueling outage. (H.9) |
| 05000249/FIN-2017003-01 | 2017Q3 | Severity level Enforcement Discretion | NRC identified | Granted Notice of Enforcement Discretion 173001: LCO 3.1.7 Required Action B.1 per TS 3.1.7, Standby Liquid Control System | Inspection Scope The inspectors reviewed the licensees response to and assessment of a through- wall leak that developed on the Unit 3 SLC A pump discharge piping . Specifically, on September 12, 2017, during a system operational pressure test, licensee personnel observed a through- wall leak from the forged body of a 1.5 stainless steel pipe T in the Unit 3 SLC system. The affected component is a part of the ASME Code Class 2 boundary. Due to the piping being ASME Code Class 2, it was required to be immediately isolated in accordance with Technical Requirements Manual 3.4.a, Structural Integrity. Isolating this piping resulted in both trains of the Unit 3 SLC system becoming inoperable as the leak was unisolable from both pumps. With both trains inoperable, the licensee entered Limiting Condition for Operation ( LCO ) 3.1.7, Required Action B.1 which requires the restoration of at least one train of SLC within 8 hours. 15 The inspectors examined the sites actions to uncover the issue with the Unit 3 SLC system , their actions to address the issue once it was identified, and their compensatory actions associated with the receipt of the Notice of Enforcement Discretion ( NOED ). The inspectors also reviewed licensee documents to verify that information contained in the NOED request was accurate. Inspection activities included gathering additional information regarding the licensees bases for requesting the NOED; examining the sites decision -making process for the issue; reviewing the licensees condition evaluation; observing the licensees compensatory actions; and evaluating the licensees operability determination. To correct this issue and exit the NOED, the licensee completed replacement of the affected Unit 3 pipi ng and connections, satisfactorily tested the replaced components, and declared the Unit 3 SLC system operable. Documents reviewed are listed in the Attachment. This event follow up review constituted one sample as defined in IP 71153 05. b. Findings Introduction : The inspectors opened an unresolved item associated with a potential noncompliance with TS 3.1.7 Required Action B.1 that occurred on September 12, 2017. NOED 17 3001 was granted by the NRC staff agreeing not to enforce compliance with the TS completion time for an additional 35 hours. Description : On September 10, 2017, with the Unit 3 SLC system in standby operation, an equipment operator performing rounds noted sodium pentaborate crystallization build -up under piping insulation. The licensee removed the insulation from the potential leak location, and noted a dry sodium pentaborate stain on the back of a forged piping T on the 1.5 stainless steel discharge line of the A SLC pump. The licensee Shift Manager made an immediate operability determination of operable based on the dry nature of the stain and its location being on a forged body , and not at a connection or weld location. The licensees initial evaluation surmised the stain was historical in nature and was from an adjacent valve packing leak. In the event that further investigation of the stain indicated a through -wall leak, the licensee investigated American Society of Mechanical Engineers ( ASME ) code compliant permanent and temporary repair options, to include the construction of an Engineered Clamp. This method was eventually dismissed as supports required for the clamp would have been impractical based on system configuration. On September 12, 2017, the licensee cleaned the stain off of the piping T and performed a visual inspection for leakage with the system at full operating pressure. During this test, a leak was observed emanating from the body of the piping T. Due to the leak occur ring within the ASME Code Class 2 boundary, the licensee was required to isolate it in accordance with Technical Requirements Manual 3.4.a, Structural Integrity. Isolating this piping resulted in both trains of the Unit 3 SLC system becoming inoperable, and therefore the licensee entered LCO 3.1.7, Required Action B.1, with an 8 hour required action. With a through wall leak discovered and the plant in a short duration shutdown LCO, the licensee implemented a repair plan for a permanent piping replacement and requested a NOED from the NRC to complete repairs prior to entering Required Action C.1 and C.2, which require placing the Unit in Mode 3 (hot shutdown) and Mode 4 (cold shutdown) within 12 and 36 hours , respectively. The NRC granted a NOED for an additional 35 hours at 5:46 p.m. on September 12, 2017. Consistent with NRC policy, the NRC agreed not to enforce 16 compliance with the specific TSs in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine whether there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (Unresolved Item 05000249/2017003 01, Granted Notice of Enforcement Discretion 17 3001: LCO 3.1.7 Required Action B.1 per TS 3.1.7, Standby Liquid Control System ) | |
| 05000237/FIN-2017002-01 | 2017Q2 | Green | P.3 | Self-revealing | Failure to Maintain Configuration Control in the Unit 2 Containment Pressure Suppression System | Green . A finding of very low safety significance and associated non- cited violation of Technical Specification ( TS ) 5.4.1, Procedures, was self -revealed on May 26, 2017, for the licensees failure to maintain configuration control in the Unit 2 containment pressure suppression system. Specifically, the licensee failed to maintain the instrument air stop valve to the actuator for the Unit 2 torus vent , air operated valve (AOV) 21601 60, open with the react or mode switch in Run (Mode 1) and reactor power approximately 100 percent rated thermal power (RTP). The inspectors determined that the licensees failure to maintain configuration control of the Unit 2 containment pressure suppression system was contrary to procedures for the emergency depressurization of containment with the reactor in Mode 1 and was a performance deficiency. The inspectors determined that the performance deficiency was more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the mitigating systems cornerstone attribute of configuration control with regards to the plants operating equipment alignment while affecting the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) . The inspectors determined that a Detailed Risk Evaluation was required to be performed based on answering Yes to the Mitigating Systems screening question A.4 in IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2. The result of the detailed risk evaluation was a finding of very low safety significance (Green) . This finding has a cross -cutting aspect of Resolution in the area of Problem Identification and Resolution because the licensee did not implement appropriate robust barriers to prevent bumping of the 2 1601 60SV in response to previous corrective actions 511878 02 and 2414608 16. Specifically, an identical maintenance induced bumping event resulted in the instrument air stop valve to t he Unit 3 torus main vent AOV 3 1601 60 being unintentionally repositioned closed in November 2014. Licensee corrective actions from that event addressed restraining potentially vulnerable valves prior to maintenance activities as well reassessing which ball valves required permanent robust barrier installation. (P.3) |
| 05000237/FIN-2017007-01 | 2017Q1 | Green | NRC identified | Inadequate Pre - Fire Plans | Green . The inspectors identified a finding of very-low safety significance (Green) and associated NCV of license conditions 2.E and 3.G for Units 2 and 3, respectively for the licensees failure to include the correct information in pre-fire plans. Specifically, the licensee failed to provide the location of compressed flammable gas cylinders and include d them in the Hazards in Area section of the pre-fire plans for two fire areas as required by Procedure OP - AA - 201 - 008, Pre-Fire Plan Manual. The licensee entered the issue into their Corrective Action Program and updated the pre-fire plans to contain the correct information. The inspectors determined that the performance deficiency was more-than-minor because the lack of information in the pre-fire plans regarding the hazards in the area could complicate firefighting activities by the fire brigade and could either increase the likelihood of a larger fire event or the severity of the fire . The finding was of very-low safety significance because it was associated with pre-fire plans and because the fire brigade members receive extensive training to deal with unexpected contingencies. The finding did not have a cross-cutting aspect associated with it because it was not representative of current performance as the licensee last updated the pre-fire plans in 2010. | |
| 05000237/FIN-2017007-02 | 2017Q1 | Green | NRC identified | Inadequate Procedure Steps to Ensure Proper Valve Rotation for Cold Shutdown Repair | Green . The inspectors identified a finding of very - low safety significance (Green) and associated NCV of Technical Specification 5.4.1.c for the licensees failure to have appropriate written procedures covering the Fire Protection Program for cold shutdown repairs. Specifically , Procedure DSSP 0200-T8 included inadequate repair instructions for three motor operated valves ( MOVs ) that if implemented as written could result in the valve rotating in undesired safe shutdown position , caused damage to MOVs and prevented manipulating the valve to the desire position and caused a delay in reaching cold shutdown condition. The licensee entered the issue into their Corrective Action Program, revised DSSP 0200-T8 and corrected the cable designations at the Motor Control Center for these MOVs for proper connection and phase rotation. The inspectors determined that the performance deficiency was more-than- minor because the inadequate instruction in the repair procedure could have delayed reaching cold shutdown in the event of a fire and added unnecessary burden for operations personnel during an already challenging fire event. The finding was of very-low safety significance per Task 1.3.1 of IMC 0609, Appendix F , because it only affected the ability to reach and maintain cold shutdown conditions. The finding did not have a cross-cutting aspect associated with it because it was not representative of current performance . | |
| 05000237/FIN-2017001-01 | 2017Q1 | Green | H.7 | NRC identified | Failure to Correct a Condition Adverse to Quality Associated with EDG Single Largest Load Rejection Surveillance Testing | The NRC identified a finding of very low safety significance and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality, originally identified in Issue Report (IR) 2501498, associated with instructions and acceptance criteria in the emergency diesel generator (EDG) surveillance procedures to ensure that the single largest load rejection test bounded the power demand of the largest load in accordance with Technical Specification Surveillance Requirement (TSSR) 3.8.1.10. Specifically, the failure to correct a condition adverse to quality associated with the inadequate performance of TSSR 3.8.1.10 required an operability determination and engineering assessment to ensure continued operability of the sites three EDGs. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events t prevent undesirable consequences (i.e. core damage). The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. The inspectors answered No to all questions in Exhibit 2, Mitigating Systems Screening Questions, Section A: Mitigating SSCs and Functionality. Therefore, the finding was screened as very low safety significance. The inspectors concluded that the cause of the finding involved a cross-cutting component in the area of Human Performance, Documentation, in that the licensee did not create and maintain complete, accurate and up-to-date documentation. Specifically, the licensee utilized surveillance procedures (DOS 660003, 04 and 05) which did not ensure that design post-accident conditions were met during testing. In addition, the licensee created Corrective Action Program (CAP) actions, to make procedure changes to operations surveillance DOS 660012 to establish bounding conditions for TSSR 3.8.1.10, that were never incorporated. |
| 05000237/FIN-2017001-02 | 2017Q1 | Green | P.1 | Self-revealing | Secondary Containment Inoperability Due to Lapse in Procedure Use and Adherence | A self-revealed finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1, Procedures, occurred on November 8, 2016, due to the licensees failure to follow procedures designed to ensure secondary containment integrity, when reactor building (RB) pressure relative to the outside environment was less than 0.25 inches water column (in WC) vacuum as required by TS 3.6.4.1, Secondary Containment. Specifically, work group personnel did not communicate to operations regarding degraded sealing surfaces on the RB Equipment Access outer door as required by procedure DAP 1303, Unit 2 Reactor Building Trackway Interlock Door Access Control, therefore when standby gas treatment (SBGT) started as a part of a planned surveillance test, vacuum lowered, rendering secondary containment inoperable. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Barrier Integrity Cornerstone Attribute of Human Performance and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the drop in secondary containment differential pressure to less than 0.25 in WC vacuum, resulted in a loss of secondary containment and failure of its safety function as specified by TS 3.6.4.1 and Updated Final Safety Analysis Report (UFSAR) section 6.2.3. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. The inspectors reviewed the Barrier Integrity Screening Questions in Appendix A, Exhibit 3 and answered Yes to question C.1. As a result, the finding was determined to have very low safety significance (Green). This finding has a cross cutting aspect in the area of Problem Identification and Resolution, Identification, because individuals failed to identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee did not report a condition adverse to quality with regards to degraded seals on the RB equipment access outer door to operations as required by procedure DAP 1303, therefore not ensuring secondary containment integrity. |
| 05000237/FIN-2016004-01 | 2016Q4 | Green | H.11 | Self-revealing | Failure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate Alarms | A finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was self-revealed when workers violated a radiation work permit (RWP) by entering an area that was outside of the scope of the original RWP brief without obtaining a required appropriate brief, resulting in these workers receiving unplanned electronic dosimeter dose rate alarms. These workers immediately exited the area and reported the event to the radiation protection staff. The licensee entered these issues as two separate events into their CAP as Issue Reports (IR) 02735594 and IR 02735651. The inspectors determined that the performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, because the finding impacted the program and process attribute 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. Specifically, worker entry into areas beyond the RWP briefing could lead to unintended dose. The finding was determined to be of very-low safety significance (Green) in accordance with Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls, (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 inspectors concluded that the cause of the finding involved a cross-cutting component in the human performance area of challenging the unknown because the individual did not stop when faced with an uncertain condition. Risks were not evaluated and managed before proceeding (H.11). |
| 05000237/FIN-2016004-02 | 2016Q4 | Green | Licensee-identified | Licensee-Identified Violation | Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(4) states, A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, between April 2013, and February 2016, the licensee failed to maintain the effectiveness of the emergency plan by failing to maintain the effluent parameters contained in the standard emergency classification and action level scheme. Specifically, the standard emergency classification and action level scheme associated with the radiological effluents at Dresden Nuclear Power Station was not updated to reflect the changes in the X/Q dispersion factor that were made during the April 2013, Offsite Dose Calculation Manual revision. Consequently, the effluent monitor emergency classification and action level thresholds were non-conservative by a factor of 3.8 until this condition was identified and corrected by Dresden Nuclear Power Station in February 2016. The inspectors determined that the finding was of very low significance (Green) in accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Figure 5.41, because the emergency action level classification of an Unusual Event, RU1, would be declared in a degraded manner, not within the required 15 minutes. The emergency action level classification for the Alert, Site Area Emergency, and General Emergency (RA1, RS1, and RG1) would still be capable of being declared in timely manner, within 15 minutes, using alternate conditions within the emergency action level. Because this finding is of very low safety significance, and has been entered into Exelons CAP under IR 02652711, this violation is being treated as a Green NCV consistent with Section 2.3.2 of the NRCs Enforcement Policy. | |
| 05000249/FIN-2016010-01 | 2016Q4 | White | H.6 | Self-revealing | Failure to Verify the Adequacy of Design for the Unit 3 HPCI AOP Motor Shunt Resistor Setting | A self-revealing finding preliminarily determined to be of low to moderate safety significance, and an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was associated with the licensees failure to ensure that the applicable design basis for applicable structures, systems, and components was maintained by the performance of design reviews, through the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, the licensee failed to verify the adequacy of design for the Unit 3 high pressure coolant injection (HPCI) auxiliary oil pump (AOP) motor shunt resistor setting during motor replacement in March of 2002, and then again in March of 2015, eventually resulting in pump failure in June of 2016, and inoperability of the HPCI system. The licensee documented this issue in its corrective action program (CAP) as IR 2686163. The inspectors determined that the licensees failure to verify the adequacy of design for the Unit 3 HPCI AOP motor shunt resistor setting was a performance deficiency, the cause was reasonably within the licensees ability to foresee and correct due to previous events and licensee generated causal determinations regarding the significance of adjusting the shunt field resistors on motor and pump operations, and should have been prevented. The inspectors determined the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Design Control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the failure to control the design of the Unit 3 HPCI AOP motor resulted in the degradation and ultimate failure of the pump motor windings, which is a required component for HPCI operation. The inspectors applied IMC 0609, Attachment 4, and IMC 0609, Appendix A, Exhibit 2, Section A, for Mitigating Systems to screen this finding and determined that a detailed risk evaluation was required because the finding represented a loss of system and/or function. Therefore, a coordinated effort between inspection staff and regional Senior Reactor Analyst (SRA) was required to perform an appropriate risk evaluation for the degraded condition that resulted from the finding. The SRA used the Dresden Standardized Plant Analysis Risk (SPAR) model, version 8.24 for the detailed risk evaluation. This evaluation concluded that the exposure time for the HPCI system was 1 year. The total delta core damage frequency (CDF) for the 1 year exposure period was 6.9E6/year, which is a finding of low to moderate safety significance (White). HPCI is an important high pressure injection system that is used to mitigate internal events, internal flooding, and internal fire events at Dresden. The inspectors determined the contributing cause that provided the most insight into the performance deficiency was associated with the crosscutting area of Human Performance, Design Margins because the licensee failed to operate and maintain equipment within design margins, in that margins are carefully guarded and changed only through a systematic and rigorous process with special attention placed on maintaining fission product barriers, defense-in-depth, and safety-related equipment (H.6). Specifically, the licensee failed to verify the adequacy of design for the Unit 3 HPCI AOP motor shunt resistor setting during motor replacement in March of 2002 and then again in March of 2015. |
| 05000237/FIN-2016009-01 | 2016Q3 | Green | NRC identified | Main Steam Acoustic Safety/Relief Valve Monitoring Channel Calibration Not Performed | The inspectors identified a finding of very-low safety significance for the failure to perform a 24-month channel calibration of the Regulatory Guide 1.97 safety/relief valve acoustic monitoring system in accordance with the Technical Requirements Manual. Specifically, the licensee failed to perform a channel calibration, where the channel calibration shall encompass all devices in the channel required for channel operability and the channel functional test. The performance deficiency was determined to be more-than-minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to maintain the acoustic safety/relief valve position indicators instrumentation in accordance with the Technical Requirements Manual. The performance deficiency affected the design or qualification of a mitigating system, structure or component; however, the system, structure or component maintained its functionality based on successful completion of channel functionality checks. Since the system, structure or component remained functional, the inspectors screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of the licensees current performance. | |
| 05000237/FIN-2016003-02 | 2016Q3 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Appendix R, Section III.G.3 requires, in part, that an alternative dedicated shutdown capability and its associated circuits, independent of cables, systems, or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section. Compliance with 10 CFR Part 50, Appendix R, Section III.L is considered necessary to satisfy the requirements of 10 CFR Part 50, Appendix R, Section III.G. Section III.L of 10 CFR Part 50, Appendix R, requires implementation of an alternative dedicated shutdown capability as required by Section III.G.3 of 10 CFR Part 50, Appendix R. Section III.L.3 of 10 CFR Part 50, Appendix R, states, in part, that alternative shutdown capability shall be independent of the specific fire area and that procedures shall be in effect to implement this capability. Contrary to the above, from October 15, 2003, until present, the licensee failed to maintain in effect all provisions of 10 CFR Part 50, Appendix R, Section III.G.3 and Section III.L. Specifically, the licensee failed to ensure that systems that were required for alternative shutdown capability were not free of fire effects, therefore, were not independent of the specific fire area. The licensee credits the HPCI system as the alternative to the isolation condenser for hot shutdown. Licensee procedures DSSP 0100C, Hot Shutdown Procedure Path C Revision 27 and DSSP 0100D, Hot Shutdown Procedure Path D Revision 26, inappropriately direct operators to lift leads and install electrical jumpers in order to defeat HPCI suction transfer from the condensate storage tank (CST) to the torus on low CST level or high torus level. Installation of jumpers and lifting leads is considered a repair and is not permissible for systems required to achieve safe hot shutdown. In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection SDP. The inspectors determined that the finding impacted the ability to achieve safe shutdown and, assigned the finding to the category of 1.4.5 Post-fire Safe Shutdown using Table 1 in IMC 0609, Appendix F, Attachment 1, Part 1: Fire Protection SDP Phase 1 Worksheet, dated September 20, 2013. The inspectors answered no to Question 1.4.5B, Does the fire finding affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event? in Task 1.4.5 of IMC 0609, Appendix F. The repair actions already in place in procedures DSSP 0100C and DSSP 0100D, while not allowed by Appendix R, were determined to be a viable compensatory measure that would allow the plant to reach and maintain a stable hot shutdown condition. Therefore, the inspectors determined that the finding screened as having very low safety significance (Green). This issue was entered into the licensees CAP as IR 2651479. | |
| 05000237/FIN-2016003-01 | 2016Q3 | Green | P.2 | Self-revealing | Failure to Assess Scope Changes to Corrective Maintenance Activities Affecting Safety-Related Structures, Systems, and Components | A finding of very low safety significance and associated NCV of TS 5.4.1.a, Procedures, was self-revealed for the licensees failure to maintain maintenance procedures appropriate for the circumstances that could affect performance of safety related equipment. Specifically, procedures MAAA716010, Maintenance Planning, Revision 20 and DAP 1518, Work Order Supplemental Information and Lessons Learned, Revision 17 did not ensure that scope revisions in support of corrective maintenance activities performed on high pressure coolant injection (HPCI) piping in 2013 were properly reviewed and evaluated for technical adequacy directly resulting in a through-wall steam leak on the Unit 2 HPCI inlet drain pot drain piping and safety system inoperability in May 2016. Immediate corrective actions included the replacement of the failed piping section, a determination of the extent of condition of susceptible piping to include the scheduling of a replacement work window, and changes to the maintenance planning procedures requiring engineering scope determination and oversight of scope changes for safety related corrective maintenance. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone Attribute of Procedure Quality 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). Specifically, the failure to ensure work planning procedures controlled the process of major revisions to corrective maintenance activities ensuring adequate engineering reviewing and assessment resulted in continued degradation and ultimate failure of the Unit 2 HPCI inlet drain pot drain piping. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012. The inspectors answered No to all questions in Exhibit 2, Mitigating Systems Screening Questions. Therefore, the finding was screened as very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee failed to thoroughly evaluate corrective maintenance scope changes to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee incorrectly removed scope without engineering evaluation for adequacy from the Unit 2 HPCI inlet drain pot drain line corrective maintenance following a through wall leak in 2012. Piping that was identified as part of the extent of condition of the failure in 2012, was removed from the scope of corrective maintenance activities due to maintenance personnel short falls. This specific piping failed in May of 2016 resulting in the loss of the HPCI system safety function. (P.2) |
| 05000237/FIN-2016002-01 | 2016Q2 | Green | H.1 | NRC identified | Failure to Implement and Maintain Written Procedures Regarding Breathing Air Quality Testing | A finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1703, was an NRC-identified finding for failure to implement and maintain written procedures regarding breathing air quality that resulted in the failure to perform a continuous in-line breathing air quality test during filling of self-contained breathing apparatus (SCBA) cylinders since 2009. Specifically, on May 4, 2016, during an inspection of the licensees air compressor, the inspectors identified that the in-line carbon monoxide (CO) detector located at the compressor highpressure filling station was inoperable since 2009, the procedure does not specify an alternative method of CO monitoring during the filling of the SCBA cylinders. Without specifying an alternative method of monitoring and only relying on the high-temperature safety shut-off, hazardous CO gas could be introduced into the SCBA cylinders, thus degrading the Grade-D air quality, during a compressor malfunction. The licensees corrective actions included but were not limited to revising the applicable procedures, servicing or replacing the CO monitor by the manufacturer, and installing a new air compressor at the facility. The inspectors determined that that the finding was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, in that the finding impacted the program and process attribute 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 through the use of SCBAs during an emergency response use by maintaining certified air quality. Specifically, the licensee failed to implement and maintain written procedures regarding an alternative method of monitoring air quality testing to maintain the Grade-D air quality during filling of SCBA cylinders. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as-low-as-reasonably-achievable planning issue, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting component in the area of human performance, resources, in that, the license did not ensure the adequacy of the procedure describing the alternate methods of CO monitoring during filling of Grade D air into the SCBA cylinders. (H.1) |
| 05000237/FIN-2016201-01 | 2016Q2 | Green | NRC identified | Security | ||
| 05000237/FIN-2016001-01 | 2016Q1 | Green | H.9 | NRC identified | Failure to Maintain Design Control of the 2/3 Emergency Diesel Generator | A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was self-revealed associated with the licensees failure to assure that the applicable design basis for applicable structures, systems, and components were correctly translated into specifications, procedures, and instructions. Specifically, since initial plant construction the licensee failed to correctly identify the effect a loss of non-safety 2/3 emergency diesel generator (EDG) room ventilation could have on maintaining operability of the 2/3 EDG. On November 6, 2015, during a planned maintenance outage of the normal non-safety related instrument air pneumatic supply and a failure resulting in the depressurization of the back-up non-safety related nitrogen system, the 2/3 EDG ventilation intake and exhaust dampers failed closed making the 2/3 EDG inoperable for approximately 20 minutes on two occasions from the time of discovery of the condition. The licensee incorrectly believed that a loss of the non-safety related instrument air system and its non-safety related back-up nitrogen system would cause the dampers to fail in the conservative open position. This feature was never tested; and therefore the licensee incorrectly believed the non-safety related control systems for the room ventilation system would not adversely affect the safety-related EDGs operability. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and if left uncorrected could lead to a more significant safety concern. The finding screened as very low safety significance (Green) because the inspectors answered no to questions A.1. through A.4. of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, dated June 19, 2012. This finding has a cross-cutting aspect in the area of Human Performance, Training, because the licensee did not ensure licensed operations and engineering personnel properly understood the operation and configuration of the 2/3 diesel generator ventilation system under accident conditions and its impact on the safety-related 2/3 EDGs ability to accomplish its design function. Specifically, the licensee incorrectly believed that the 2/3 EDG room ventilation system failed in a conservative manner with a loss of its non-safety related pneumatic supply systems. Corrective Action Program documents and other engineering products up until September 2015 incorrectly state that the 2/3 EDGs operability was not adversely affected by a loss of damper control pneumatics as the dampers were expected to fail open. |
| 05000237/FIN-2016405-01 | 2016Q1 | Licensee-identified | Licensee-Identified Violation | |||
| 05000237/FIN-2015004-01 | 2015Q4 | Green | H.14 | Self-revealing | Failure to Maintain Design Control of Secondary Containment Interlock Doors | A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed on September 4, 2015, when the integrity of the Secondary Containment for Units 2 and 3 was not maintained for 39 minutes when interlock features designed to prevent both doors of a Secondary Containment interlock from being simultaneously open prevented the closure of Reactor Building to Turbine Building doors 47 and 48 following simultaneous operation during routine access of the interlock by plant personnel. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of design control, and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance (Green) because the inspectors answered yes to the Barrier Integrity Screening Question C.1, Exhibit 3 of IMC 0609, Appendix A. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee failed to implement a modification which addressed a known design deficiency in the 570 foot elevation Secondary Containment interlock in 2013. The licensee reasoned that the interlock was a low traffic area and that it would be unlikely that the doors would be open simultaneously. (H.14) |
| 05000249/FIN-2015003-04 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation | A violation of TS 3.3.6.1(A.1), Primary Containment Isolation Instrumentation was identified by the licensee during a review of recent operations logs and CAP documents on September 7, 2015 by on-shift operators. On September 5, 2015, operators performing shiftly TS rounds identified that the Unit 3, A main steam line (MSL) steam flow detector, 3-0261-2A, was reading more than 25 pounds per square inch differential (psid) lower than the other three steam flow detectors on the A MSL contrary to Surveillance Requirement 3.3.6.1.1. This issue was entered into the licensees CAP at this time as IR 2551890, where the on-shift Senior Reactor Operator (SRO) recommended performing DIS 0250-01, Dresden Unit 3 Quarterly Main Steam Line High Flow Switch Calibration and incorrectly assessed TS 3.3.6.1 function 1.d Main Steam Line Flow High as being inoperable but not requiring the failed channel to be placed in trip for the failed steam flow switch. The SRO incorrectly applied the operable logic of two required channels on each MSL per trip system and did not recognize that one failed channel would require entry into TS 3.3.6.1(A.1). TS 3.3.6.1(A.1) requires that operators place the failed channel in trip within 24 hours of discovery. Operations on-shift reviews of recent logs and issue reports identified the missed entry into TS 3.3.6.1(A.1) on September 7, 2015 more than 24 hours later. The missed TS entry was entered into the licensees CAP as IR 2552152. Subsequent performance of DIS 0250-01 identified two additional channels outside of TS required trip values for MSL high steam flow. The failure to enter a required TS action statement when declaring A MSL flow detector 3-0261-2A inoperable was considered a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, a separate single failure could have prevented the safety function of isolating the A MSL on a steam line rupture casualty. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to the Appendix A, Exhibit 3 Barrier Integrity Screening Questions, therefore, the finding was determined to be of very low safety significance (Green). Licensee corrective actions included disqualifying the SRO from shift duties pending remedial training, performing an apparent cause evaluation concerning the missed TS action statement entry, replacing the failed A MSL flow switch, and calibrating all other channels which were outside the required operating band of DIS 0250-01. | |
| 05000249/FIN-2015003-03 | 2015Q3 | Green | H.5 | NRC identified | Failure to Post Protected Pathway Signs | The inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to implement all necessary prescribed risk management actions as required by licensee procedure OP-AA-108-117, Protected Equipment Program, during a Unit 3 low pressure coolant injection (LPCI) logic system functional test (LSFT) maintenance window. Specifically, the licensee failed to post protected equipment signs for the Unit 3 2203-28 instrument rack whose unavailability impacts the high pressure coolant injection (HPCI) system and would have taken the unit into a licensee-defined Orange risk condition and Technical Specification (TS) Limiting Condition for Operations (LCO) 3.0.3. The performance deficiency was determined to be more than minor because the licensee failed to implement prescribed risk management actions which if left uncorrected could have become a more significant safety concern. Specifically, operators inadvertently removed the protected equipment postings on instrument rack 2203-28 which could have significantly degraded the key safety function of reactor coolant inventory control if both the HPCI and LPCI systems became unavailable simultaneously and were required to mitigate the consequences of an accident that could result in potential offsite exposure comparable to the 10 CFR Part 100 guidelines. The finding screened as very low safety significance (Green) because the inspectors answered No to all of the Mitigating Systems Screening Questions. The inspectors determined this finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the licensee failed to implement risk management actions in accordance with procedure OP-AA-108-117. Specifically, the licensee inappropriately removed protected equipment postings prior to commencing maintenance activities which required the risk management activity to be in place. |
| 05000237/FIN-2015003-02 | 2015Q3 | Green | H.6 | NRC identified | Failure to Perform Ultimate Heat Sink Surveys in Accordance With Quality Assurance Program | The inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to perform surveys of the UHS in accordance with the Quality Assurance Program. Specifically, the licensee failed to ensure that: (1) the requirements and acceptance limits in the Updated Final Safety Analysis Report (UFSAR) and other applicable design documents for the high points of the intake and discharge canals were incorporated into the UHS test; (2) the evaluation of the bathymetric survey results accounted for the instrument uncertainty of the test equipment; and (3) the UHS bathymetric survey results were evaluated in accordance with the Quality Assurance Program to assure that the UFSAR required UHS volume was satisfied. The licensee entered this finding into their CAP and, after a review of past bathymetric surveys and other information, determined that: (1) the discharge canal high point had not degraded and was not expected to significantly degrade; and (2) the UHS remained operable because the available UHS water volume still remained above the UFSAR required volume. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality, and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because although it affected the design or qualification of the UHS, it did not result in the loss of operability or functionality of the UHS. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance, Design Margins, because the licensee did not pay special attention to maintaining the safety-related UHS. Specifically, special attention was not placed on guarding the design margin of the UHS, and a test program that did not meet all the quality assurance requirements was accepted to demonstrate the adequacy of the UHS. |
| 05000237/FIN-2015003-01 | 2015Q3 | Green | NRC identified | Failure to Evaluate the Available Net Positive Suction Head for the Diesel Generator Cooling Water Pumps Following a Dam Failure | The inspectors identified a finding of very low safety significance, and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the diesel generator cooling water (DGCW) pumps design. Specifically, the licensee failed to evaluate the net positive suction head (NPSH) available to the DGCW pumps at the most limiting ultimate heat sink (UHS) level after a postulated dam failure where they are expected to perform a safety function. The licensee entered this finding into their Corrective Action Program (CAP) and, after a review of their DGCW pump NPSH calculation and an evaluation, concluded that the DGCW pumps had adequate NPSH available to them at the most limiting UHS level after a postulated dam failure, and therefore remained operable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because although it affected the design or qualification of the DGCW pumps, it did not result in the loss of operability or functionality of the pumps. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. | |
| 05000237/FIN-2015002-01 | 2015Q2 | Green | H.7 | Self-revealing | Failure to Meet Technical Specification Surveillance Requirements Due to Foreign Material Left in the Unit 2 EDG Starting Circuit | A finding of very-low safety significance (Green) was self-revealed on April 21, 2015 while performing TS Surveillance DOS 6600-12, Diesel Generator Tests: Endurance and Margin/Full Load Rejection/ECCS (Emergency Core Cooling System)/Hot Restart, in support of Surveillance Requirement 3.8.1.16 which requires the EDG to achieve rated frequency and voltage conditions within 13 seconds when started less than or equal to five minutes from a previously loaded run, the Unit 2 Emergency Diesel Generator (EDG) failed to complete a hot restart. Licensee troubleshooting identified a degraded pressure switch associated with main bearing lube oil pressure in the start circuit which was taking several minutes to return to a low-pressure condition upon shutting down the EDG. This resulted in a failure of the start circuit relay to be energized upon initiating a start of the EDG, until the pressure switch returned to its appropriate low-pressure state. An internal investigation of the pressure switch identified strips of Teflon tape in the bellows of the pressure switch, which resulted in the pressure switchs sluggish response to lowering lube oil pressure, and a failure to meet the TS hot restart criteria. The inspectors determined that the failure to implement Procedure MAAA716-008, Foreign Material Exclusion Program, and therefore the inability to perform TS Surveillance Requirement 3.8.1.16 was a performance deficiency, and was considered more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone, and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors utilized Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very-low safety significance because each question provided in Inspection Manual Chapter (IMC) 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Documentation area, because licensee procedure MA-AA716-008, Foreign Material Exclusion Program, work instructions associated with Work Order 01410972-01, and previous calibrations of pressure switch 2-6641-526 did not include specific instructions and warnings regarding the proper use of Teflon tape with regards to preventing it from becoming foreign material. Other Dresden maintenance procedures, specifically MA-DR-0300-001, Preventive Maintenance of Hydraulic Control Unit, and DEP 0300-16, Rebuilding the Unit 2 (3) ASCO Scram Solenoid Pilot Valves, have specific warnings regarding the proper use and potential for Teflon tape to become foreign material. (H.7) |
| 05000237/FIN-2015002-05 | 2015Q2 | Green | Licensee-identified | Licensee-Identified Violation | The TS 5.5.1 required that the Offsite Dose Calculation Manual (ODCM), and its Radiological Environmental Monitoring Program (REMP) be established, implemented, and maintained. ODCM Radiological Environmental Control No. 12.6.1 defined the surveillance requirements for the REMP. Step E of this section provided requirements for Milk Station D-25 (Control) be sampled within 10 km to 30 km semimonthly as indicated in Table 12.6-1.4.a. Contrary to the requirements, the licensee did not sample the control Milk Station D-25. The missed samples were not identified by the licensee until March 2015. This issue was entered into the licensees CAP as AR-02469852. | |
| 05000237/FIN-2015002-02 | 2015Q2 | Green | H.2 | Self-revealing | Inadvertent Manipulation of a Test Switch at ESF Bus 23-1 During Surveillance Testing Results in the Inoperability of the 2/3 EDG to Unit 2 | A finding of very low safety significance (Green), and an associated NCV of TS 5.4.1, Procedures, was self-revealed on May 19, 2015, when the 2/3 EDG was made inoperable to Unit 2 due to the incorrect manipulation of a test switch by operations personnel during a TS required surveillance test. Specifically, while the licensee performed procedure DIS 1500-05, Division I and II Low-Pressure Coolant Injection ECCS Initiation Circuitry Logic System Functional Test, Step 106 ofChecklist B, operations personnel incorrectly opened test switch TS-159SD2/3 at motor control center 23-1 removing the under-voltage trip associated with the feed breaker for the Division I safety-related 4.16 kV engineered safeguards bus, causing the 2/3 EDG to be inoperable to Unit 2. The licensees failure to properly implement steps in the procedure was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone Attribute of Configuration Control, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very- low safety significance (Green), because each of the questions provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, were answered No. The finding has a cross-cutting aspect in the area of Human Performance, Field Presence, for failing to ensure senior managers applied the appropriate oversight of infrequently performed and first time work activities. Specifically, the licensee field supervisor or another senior operations manager was not present for the switching activities, which led to the configuration control error. In this instance, the surveillance test is infrequently performed (every 24 months), and the activity, which included using a maintenance procedure vice an operating procedure, was a first time evolution for both equipment perators involved. (H.2) |
| 05000237/FIN-2015007-02 | 2015Q2 | Green | P.1 | NRC identified | EDG Usable Fuel Calculations Failed to Consider Appropriate EDG Frequency Variations | The inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to account for increased fuel oil consumption during the development of the Emergency Diesel Generator (EDG) Calculation 10553-CALC-07, Dresden Station Emergency Diesel Generators Endurance Calculations, Revision 2, which resulted in non-conservative Technical Specifications (TS). Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hertz (Hz), and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour. The licensee entered the issue into their CAP as Action Request 02506869, NRC MOD/5059 Inspection: Emergency Diesel Generator Fuel Consumption, dated May 28, 2015. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to account for the increased fuel oil consumption resulting from operation at a higher EDG frequency. Therefore, the licensee did not ensure that the minimum fuel oil level in the day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution; Identification, because the licensee did not did not thoroughly evaluate the EDG fuel oil consumption when considering EDG frequency variation. Specifically, the licensee failed to translate applicable design bases into specifications which resulted in non-conservative TS. (P.1) |
| 05000237/FIN-2015007-01 | 2015Q2 | Green | H.4 | NRC identified | Procedure Revisions Resulted in Isolation Condenser Unable to Meet Design Basis | The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that applicable regulatory requirements and the isolation condensers (ICs) design bases were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the ICs design bases. The licensee entered the issue into their Corrective Action Program (CAP) as Action Request 02506445, NRC MOD/5059 Inspection: ISCO (Isolation Condenser) Operating Procedures, dated May 28, 2015. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality, and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedures would drive the operators to stop the IC during a design bases event and prevent the IC from performing its design function of removing decay heat from the reactor. The finding has a cross-cutting aspect in the area of Human Performance; Teamwork, because the licensee did not communicate and coordinate activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Operations Department failed to communicate and coordinate with the Engineering Department when developing the procedural changes. (H.4) |
| 05000237/FIN-2015002-03 | 2015Q2 | Green | P.2 | Self-revealing | Reactor Scram Due to Feedwater Level Control System Failure with a Reactor Recirculation Pump Runback | A finding of very-low safety significance (Green) was self-revealed on January 13, 2015, and again on February 6, 2015, when a loss of power to the Unit 2 feedwater level control (FWLC) system resulted in a reactor scram. The loss in power to the Unit 2 FWLC system was determined to be the result of a human performance error during the original installation of the system under Work Order (WO) 97102835, in that two spade-lug connections associated with the systems +5 Vdc power supply were not properly landed resulting in the intermittent losses in power, and reset of the FWLC system. In addition, a dual in-line package switch on a FWLC Input/Output card was improperly positioned which led to an improper anti-cavitation reactor recirculation pump runback during both events. The inspectors determined that the failure to properly land the leads associated with the Unit 2 FWLC system +5 Vdc power supply in accordance with the work instructions in WO 97102835 was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the configuration control attribute of the Initiating Events cornerstone, and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very-low safety significance (Green), because the inspectors answered "No" to the screening question, Did the finding 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 (e.g., loss off condenser, loss of feedwater)? This finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee did not thoroughly evaluate repetitive alarms and a failure of the FWLC system to ensure that resolutions addressed causes and extent of condition prior to restart following the January 13, 2015, FWLC failure and reactor scram. Specifically, licensee analysis of alarms received prior to the January 13, 2015, scram and troubleshooting of the FLWC system failure on January 13, 2015, was overly focused on multi-functional processor cards which happened to be approaching their end of expected life. Activities to investigate loose wiring connections following the January 13, 2015, scram failed to identify the incorrectly landed spade-lug connections for the +5 Vdc power supply. (P.2) |
| 05000237/FIN-2015002-04 | 2015Q2 | White | P.3 | Self-revealing | Failure to Ensure Continued Operability of Unit 2 ERV 2-02033C (2C) Following Implementation of Extended Power Uprate Plant Conditions | A finding preliminarily determined to be of lowto-moderate safety significance, and an associated Apparent Violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control; TS 3.4.3, Safety and Relief Valves; and TS 3.5.1, ECCS Operating, was self-revealed on February 7, 2015, following the discovery that one of the Unit 2 electromatic relief valves (ERVs) would not have performed its intended safety function. Vibration induced wear experienced while operating at extended power uprate (EPU) power levels resulted in the degradation of multiple ERV actuator subcomponents, which rendered the valve inoperable. This finding does not represent an immediate safety concern in that the licensee has replaced all Unit 2 and 3 ERV actuators with a hardened design successfully utilized at the Quad Cities Nuclear Power Station, which has also experienced significant steam line vibrations post EPU. The inspectors determined that the licensees apparent failure to ensure measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of SSCs, in particular ERV 2-0203-3C (2C), was a performance deficiency warranting a significance evaluation. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance. The inspectors determined that this finding has a cross-cutting aspect of Resolution in the area of Problem Identification and Resolution, since it involves the failure to implement effective corrective actions to address issues in a timely manner commensurate with their safety significance. This cross-cutting issue is conditional depending on the outcome of the preliminary White finding. (P.3) |
| 05000237/FIN-2015001-01 | 2015Q1 | Green | H.5 | Self-revealing | 10 CFR 20.1701; Failure to Implement Effective Radiological Engineering Controls | A finding of very-low safety significance, and an associated NCV of 10 CFR 20.1701 was self-revealed during work activities associated with the failure to effectively implement planned radiological engineering controls during reactor head reassembly that resulted in personal contaminations and unintended radiological intakes to workers. On November 14, 2014, during the cleaning of the reactor head studs, several workers on the refuel floor were contaminated, and received unplanned and unintended intakes of radioactive material. Corrective actions included revising applicable procedures to improve the engineering and contamination controls during reactor head reassembly. The inspectors determined that that the finding was more than minor in accordance with IMC 0612, in that the finding impacted the program and process attribute 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. Specifically, the failure to implement effective radiological engineering and contamination controls during the cleaning of the contaminated reactor head studs resulted in personal contaminations and intakes to several workers. The inspectors concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as low as reasonably-achievable planning issue, there was neither overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting component in the human performance in that the licensees management did not ensures that effective radiological engineering controls was either managed or coordinated commensurate to the work activities. |
| 05000237/FIN-2015403-01 | 2015Q1 | Green | NRC identified | Security | ||
| 05000237/FIN-2014408-02 | 2014Q4 | Severity level Enforcement Discretion | NRC identified | Licensee failed to ensure data linked to potentially disqualifying information about an individual was retained in the shared database | The inspector reviewed AR 01505017, which described a Petition for Rulemaking filed by the Nuclear Energy Institute (NEI) on January 5, 2013. The petition requested that the NRC amend its regulations to limit the scope of third-party review of licensee decisions denying or revoking an employees unescorted access at their facility. The petition stated that a person who has been determined not to be trustworthy and reliable by a licensee and denied unescorted access to a nuclear power plant could have that determination overturned by a third party. On May 8, 2014, pursuant to an arbitrators ruling, the licensee removed data linked to potentially disqualifying information regarding an individual who the licensee had previously denied unescorted access from the shared database. This issue constituted a violation of NRC requirements, in that the licensee was required to ensure that data linked to potentially disqualifying information about an individual who applied for unescorted access authorization was retained in the shared database. In addition, on July 18, 2014, the NEI requested that the NRC endorse Revisio 4 to NEI 03-01, Personnel Access Requirements for Nuclear Power Plants. Revision4 to NEI 03-01 contained a process for reviewing denials of Unescorted Access, which would allow a third party to review the circumstances surrounding the denial but ensure that NRC access autho ization requirements were being met. Although NEI has requested to withdraw the Petition for Rulemaking, the NRC and the industry are still attempting to resolve the issue. The NRC concluded that the licensee made a good faith effort to resolve the issue prior to the arbitration and that it was not reasonable for the licensee to foresee and prevent the arbitrators ruling. Therefore, no performance deficiency associated with the violation was identified. The NRC performed a risk evaluation of the issue and determined it to be of very low security significance. Based on these facts, I have been authorized, after consultation with the Director, Office of Enforcement, and the Regional Administrator, to exercise enforcement discretion and refrain from issuing enforcement for this violation | |
| 05000249/FIN-2014005-03 | 2014Q4 | Green | H.14 | Self-revealing | Failure to Maintain Configuration Control in the Unit 3 Containment Pressure Suppression System | A finding of very low safety significance and associated non-cited violation of Technical Specification (TS) 5.4.1, Procedures, was self-revealed on November 19, 2014, for the licensees failure to maintain configuration control in the Unit 3 containment pressure suppression system. Specifically, the licensee failed to maintain the instrument air stop valve to the actuator for Unit 3 torus vent 3160160 open with the reactor in the Start-up and Run Mode following refueling outage D3R23. The inspectors determined that the licensees failure to maintain configuration control of the Unit 3 containment pressure suppression system was contrary to procedures for the emergency depressurization of containment as well maintaining TS required atmospheric conditions inside the primary containment with the reactor in Mode 1 and was a performance deficiency. The inspectors determined that the finding was more than minor because it was associated with the barrier integrity cornerstone attribute of configuration control in how containment design parameters are maintained while affecting the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the finding was of very low safety significance based on answering No to all of the Barrier Integrity screening questions in IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3. The finding has a cross-cutting aspect of conservative bias in the area of Human Performance because the licensee did not implement appropriate robust barriers to prevent bumping of the 3160160SV in response to corrective action 51187802. Specifically, the licensee previously evaluated 3160160SV and non-conservatively determined that this particular valve did not require a seal to prevent inadvertent operation. |
| 05000249/FIN-2014005-01 | 2014Q4 | Green | H.9 | NRC identified | Unit 3A LPCI Heat Exchanger Supports Returned to Service with Unacceptable Indications | A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR 50.55a(g)(4) was identified by the inspectors for the licensees failure to maintain American Society of Mechanical Engineers (ASME) Code Class 2 components in accordance with ASME Code Section XI requirements. Specifically, the licensee failed to repair or replace the Unit 3A Low Pressure Coolant Injection (LPCI) heat exchanger support welds identified to have unacceptable linear flaws prior to return to service. The inspectors determined that the licensees acceptance of linear flaws in the Unit 3A LPCI heat exchanger supports that are determined to be unacceptable for continued service IAW with the ASME Code Section XI, Article IWC3000 requirements was a performance deficiency (PD). The inspectors determined that the PD was more-thanminor, and a finding, because if the PD remained uncorrected it could lead to a more significant safety concern. Absent NRC identification, the LPCI support welds with unacceptable linear flaws would have remained in service without repair or replacement. This condition could potentially lead to the failure of the Unit 3A LPCI heat exchanger supports, which in turn, could lead to a potential failure of the Unit 3A LPCI heat exchanger. The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of Findings, Table 3Significance Determination Process (SDP) Appendix Router. The inspectors answered No to the question in Section A of Table 3; and therefore, evaluated the finding using the SDP in accordance with IMC 0609, The Significance Determination Process for At-Power Operations, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors answered No to the questions in Exhibit 2 and determined that this finding did not result in a deficiency affecting the structures, systems, and components (LPCI heat exchanger) to maintain its operability or functionality. Therefore, the finding was determined to have very low safety significance. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, training, for the licensees failure to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee staff dispositioned unacceptable flaws in the LPCI heat exchanger supports for continued service using an engineering evaluation because the licensee staff lacked the specific ASME Code knowledge concerning disposition of the unacceptable indications. Therefore, the licensee failed to return the LPCI heat exchanger supports to within ASME Code acceptable flaw limits via repair or replacement prior to return to service. |
| 05000237/FIN-2014408-01 | 2014Q4 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000249/FIN-2014005-02 | 2014Q4 | White | P.5 | Self-revealing | Failure to Ensure Continued Operability of Unit 3 Electromatic Relief Valve 302033E Following Implementation of Extended Power Uprate Plant Conditions | An apparent violation (AV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having a preliminary low to moderate safety significance, was self-revealed on November 6, 2014, following the discovery that one of the Unit 3 electromatic relief valves (ERVs) would not have performed its intended safety function. Increased vibrations experienced while operating at extended power uprate (EPU) power levels resulted in the degradation of multiple ERV actuator components which rendered the valve inoperable. The inspectors determined that the licensee fully implemented the Unit 3 EPU following a main generator rewind in November 2010, but failed to verify that the ERV actuator design was suitable for operation at the continuously increased vibration levels experienced at EPU power levels. This finding does not represent an immediate safety concern in that the licensee has replaced all four Unit 3 ERV actuators with a hardened design successfully utilized at the Quad Cities Generating Station, which also experienced significant steam line vibrations post EPU. The inspectors determined that the licensees failure to ensure the continued operability of the Unit 3 ERVs following the establishment of EPU plant operating conditions was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel (SERP), using IMC 0609, Appendix A, Significance Determination Process For Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The inspectors determined that this finding has a cross-cutting aspect of operating experience in the area of Problem Identification and Resolution, since it involves the failure to implement relevant internal and external operating experience in a timely manner. |
| 05000237/FIN-2014004-02 | 2014Q3 | Green | H.7 | NRC identified | Inadequate Evacuation Time Estimate Submittals | The NRC identified a NCV of 10 CFR 50.54(q)(2) associated with 10 CFR 50.47(b)(10) and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Dresden Nuclear Power Station Emergency Plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Exelon submitted the Dresden Nuclear Power Station ETE to the NRC on December 12, 2012, prior to the required due date of December 22, 2012. The NRC completeness review found the ETEs to be incomplete due to Exelon fleet common and site-specific deficiencies, thereby preventing Exelon from providing the ETEs to responsible OROs and from updating site-specific protective action strategies as necessary. The NRC discussed its concerns regarding the completeness of the ETE, in a teleconference with Exelon on June 10, 2013, and on September 5, 2013, Exelon resubmitted the ETEs for its sites. The NRC again found the ETEs to be incomplete. The issue is a performance deficiency because it involves a failure to comply with a regulation that was under Exelons control to identify and prevent. The finding is more than minor because it is associated with the emergency preparedness cornerstone attribute of procedure quality and because it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding is of very low safety significance because it was a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The licensee had entered this issue into their corrective action program (CAP) and re-submitted a new revision of the Dresden Nuclear Power Station ETE to the NRC on May 2, 2014, which was found to be complete by the NRC. The cause of the finding is related to the cross-cutting element of Human Performance, Documentation. |
| 05000237/FIN-2014004-03 | 2014Q3 | Green | Licensee-identified | Licensee-Identified Violation | A violation of 10 CFR 50.65(b)(2)(i) was identified by the licensee during a review of systems and components utilized in the emergency operating procedures (EOP) as compared to functions scoped into the sites Maintenance Rule Program. While reviewing emergency operating procedure DEOP 3001, Secondary Containment Control the Site Maintenance Rule Coordinator (SMRC) noted that one of the entry criterion for the procedure included receiving a Reactor Building Floor Drain Sump (RBFDS) Hi-Hi level alarm. The SMRC noted that the RBFDS was scoped into the Maintenance Rule Program, but the Hi-Hi level alarm function was not. This event was entered into the licensees CAP as IR 1698084. The failure to scope into the Maintenance Rule Program non-safety related structures, systems, and components that are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures was considered a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, no maintenance performance criteria would have been established to ensure the reliability of a function serving as an entry criterion for an EOP associated with maintaining containment integrity. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, and Appendix A, The Significance Determination Process (SDP) for Findings At Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to the Appendix A, Exhibit 3 barrier integrity screening questions, therefore, the finding was determined to be of very low safety significance (Green). Licensee corrective actions included adding the RBFDS alarm function into the Maintenance Rule Program and performing an extent of condition review of all EOPs for SSC not scoped into the Maintenance Rule Program. | |
| 05000237/FIN-2014004-01 | 2014Q3 | Severity level IV | H.6 | NRC identified | Failure to Perform an Adequate 10 CFR 50.59 Evaluation for Procedure DOP 130002 | The inspectors identified a NCV of 10 CFR 50.59, Changes, Tests and Experiments, when, on February 10, 2011, the licensee failed to complete a 10 CFR 50.59 evaluation when they revised procedure DOP 130002 to change the position of Motor Operated Valve (MOV) 213013, Reactor Inlet Isolation, such that the Isolation Condenser (IC) system would not meet its design requirement of removing 84.2E+06 BTUs in 20 minutes when initiated from its minimum Technical Specification(TS) level and maximum TS temperature. The inspectors determined that the licensees failure to identify that the valve position adjustment required a 10 CFR 50.59 evaluation was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. This finding was more than minor because there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, by establishing a new position setting of MOV 213013, the licensee failed to determine that the proposed change would cause isolation condenser tubes to become exposed in the design basis accident such that it adversely affected a Final Safety Analysis Report described design function, which required an evaluation to be performed. In accordance with IMC 0612, Appendix B, Issue Screening, traditional enforcement does apply as the violation impacted the regulatory process. Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of the system and/or function, did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, and did not result in the actual loss of one or more trains of non-technical specification equipment. Inspectors assessed the violation in accordance with the Enforcement Policy, and determined it to be a Severity Level IV violation because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). This finding has a cross-cutting aspect of Design Margins (IMC 0310, H.6) in the area of human performance, for failing to carefully guard and maintain the IC design requirement of removing 84.2E+06 BTU in 20 minutes. |
| 05000237/FIN-2014008-01 | 2014Q2 | Green | H.12 | NRC identified | Inadequate Applicability Reviews of Configuration Changes for De-Energizing Safety-Related Valves | The inspectors identified a finding of very low safety significance (Green) related to inadequate applicability reviews of operational configuration changes that were implemented as a result of the licensee's Multiple Spurious Operation (MSO) evaluations. Specifically, the licensee failed to follow procedural requirements for determining the applicability for performing 10 CFR 50.59 screening and evaluations for changes made to the facility which de-energized several safety-related motor operated valves (MOVs). The procedural action required that the configuration changes be screened for applicability for a specific 10 CFR Part 50.59 evaluation since aspects of the changes were not completely controlled under the licensee's Fire Protection Program. The licensee entered this issue into their Corrective Action Program to perform a 10 CFR 50.59 screening of changes for each affected system to ensure that all aspects of component design were evaluated. The performance deficiency was determined to be more than minor because the issue, if left uncorrected, would have become a more significant safety concern. Specifically, by not individually evaluating the scope and applicability of plant configuration changes, the licensee lost the ability to ensure that all aspects of component design were appropriately evaluated against the plant's design and licensing basis. Such changes have the potential to adversely affect design or operation of systems. Failure to evaluate such aspects allows the potential for adverse changes to go undetected. This finding has a cross-cutting aspect in the area of Human Performance because the licensee became complacent during the conduct of performing applicability reviews that were related to the facility's Fire Protection Program, and failed to recognize changes that included elements outside of the scope of fire protection. (H.12). |
| 05000237/FIN-2014008-02 | 2014Q2 | Green | H.5 | NRC identified | Failure to Seismically Secure Nitrogen Bottles | The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specifications (TS) Section 5.4.1.a, for the licensees failure to seismically restrain nitrogen bottles located near safety-related motor control centers (MCCs). Specifically, the licensee failed to seismically restrain a cart with two nitrogen bottles located near safety-related MCCs per their procedures for the handling and storage of compressed gas cylinders and restraint of portable equipment. The licensee entered this issue into their corrective action program, moved the cart with the nitrogen bottles away from the MCCs, and secured it to a column nearby. The inspectors determined that the finding was more than minor because during a seismic event the bottles could have tipped over and impacted the MCCs, thereby causing a loss of safety-related equipment, such as the Unit 2/3 emergency diesel generator. The finding was determined to be of very low safety significance based on a detailed risk-evaluation. The finding has a cross-cutting aspect in the area of Human performance because maintenance and operations personnel did not coordinate during a change out of nitrogen bottles which resulted in the bottles being left unsecured. (H.5) |
| 05000237/FIN-2014003-01 | 2014Q2 | Green | H.8 | NRC identified | Failure to Take Appropriate Corrective Action When a Maintenance Rule Performance Goal for the Standby Coolant System was Not Met | The inspectors identified a finding of very low safety significance and non-cited violation of 10 CFR 50.65(a)(1), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to take corrective actions by performing an (a)(1) determination when the standby coolant supply system preventative maintenance (a)(2) demonstration was failed. Specifically, in November 2013, the standby coolant supply system exceeded its maintenance rule performance criteria when it experienced an additional maintenance preventable functional failure. The licensee failed to appropriately account for this failure in their Maintenance Rule Program and, as a result, the site failed to perform appropriate corrective action, by failing to perform an (a)(1) determination in accordance with Procedures ERAA310, Implementation of the Maintenance Rule, and ERAA-3101005, Maintenance RuleDispositioning Between (a)(1) and (a)(2), Revision 6. Corrective actions taken by the licensee to address this issue included performing a maintenance rule (a)(1) determination and placing the system into (a)(1) status. The issue was entered into the licensees corrective action program as issue report (IR) 1644740, NRC Questions D2R23 Performance of DOS 390001, and IR 1650033, MRule A1 Determination Needed for Missed MRFF Z391. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstones attribute of Equipment Performance and affected the cornerstones objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to identify a functional failure during a periodic (a)(2) demonstration purposed to provide reasonable assurance that the structures, systems, and components (SSCs), the standby coolant injection valve MO 23902, was capable of performing its intended function as specified in licensee emergency operating procedure DEOP 050003, Alternate Water Injection Systems, Revision 22. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems cornerstone. The inspectors answered Yes to the question Does the finding represent a loss of system and/or function and determined that a Detailed Risk Evaluation was required. The Senior Reactor Analysts (SRAs) evaluated the finding using the Dresden Standardized Plant Analysis Risk (SPAR) model version 8.18 and Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.0.9.0 software. The exposure time for the unavailability of the Standby Coolant Supply Valve 23902 was assumed to be the maximum value of one year. The result was a delta core damage frequency (CDF) of 6.6E8/yr. The dominant sequence was a medium loss of coolant accident initiating event with a failure of suppression pool cooling, a failure of power conversion system recovery, and a failure of late injection. Based on the Detailed Risk Evaluation, the SRAs determined that the finding was of very low safety significance (Green). This finding had a crosscutting aspect in Human Performance, Procedure Adherence, because the licensee failed to appropriately document the failure of a standby coolant supply valve in accordance with periodic test procedure DOS 390001, Standby Coolant Supply Functional Test. (H.8) |
| 05000237/FIN-2014003-02 | 2014Q2 | Green | Licensee-identified | Licensee-Identified Violation | TS 5.4.1 requires that written procedures shall be established, implemented, and maintained for procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Section 7.e.2 includes implementation of the Radiation Survey Program. Station Procedure RPAA503, unconditional release survey method requires, in part, that materials have no detectable radioactivity for unconditional release from the site. Contrary to the above, on June 9, 2014, a contracted sewage truck transporting contaminated sewage from Unit 1 ejector pit to the licensees sewage treatment plant was unconditionally released after the truck was emptied. Specifically, the sewage truck was unconditionally released to the contractors facility without the proper authorization from RP Management. On the following day, the truck was returned to the licensees facility for survey and decontamination by the RP staff. The empty sewage truck contained traced amount of radioactivity of Co60 and tritium above minimal detectable activities. The licensee investigation determined that the empty sewage truck did not leak or cause contamination during transit on the public road. This event was entered into the licensees CAP as CR 01673475. The Radiation Protection Department immediately stopped work. Future transport of sewage between the licensed facility and the licensees sewage treatment plant will be escorted by radiation protection personnel to ensure that drivers follow licensee direction. The significance of the finding was determined by using Inspection Manual Chapter 0609, Appendix D, "Public Radiation Safety SDP." The issue is of very low safety significance (Green) because it involved radioactive material control, was not a finding involving transportation, and did not result in public exposure greater than 0.005 rem. | |
| 05000237/FIN-2014403-01 | 2014Q1 | Green | NRC identified | Security | ||
| 05000237/FIN-2014403-02 | 2014Q1 | Severity level IV | Licensee-identified | Licensee-Identified Violation | ||
| 05000237/FIN-2014007-01 | 2014Q1 | Green | H.12 | NRC identified | Failure to Adequately Incorporate GE Operating Experience into Vendor Manual | The inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure that operating experience provided via a vendor Service Information Letter (SIL) was properly evaluated and incorporated into the vendor manual contrary to the requirements of procedure RSAA115, Operating Experience. The failure to properly assess operating experience for alternating current (AC) Motors resulted in a condition where specific deficiencies could go unrealized under the licensees conditioned based monitoring program. The licensee initiated action request (AR) 1633528 and 1635766 to document and develop corrective actions for the issue. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control 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 failure to adequately evaluate and document the basis for the use or rejection of 9 out of 10 experiences presented in General Electric (GE) SIL 484, Supplement 6, could cause a reduction in reliability for safety related systems that use AC motors. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding was screened against the Mitigating Systems Cornerstone, Exhibit 2 of Appendix A, and determined to be of very low safety significance because the answer was no to all of the screening questions. This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency (H.12), because individuals failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. |