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	Quarter	Site	Title	Description
05000327/FIN-2018003-01	2018Q3	Sequoyah	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Sequoyah Unit 1 Operating License Condition 2.C(16) and Sequoyah Unit 2 Operating License Condition 2.C(13) require in part that TVA shall implement and maintain in effect all provisions of the approved fire protection program. The Sequoyah fire protection report describes how the licensee complies with applicable sections of 10 CFR 50, Appendix R, including Section III.L.1 which states in part that alternative or dedicated shutdown capability provided for a specific fire area shall be able to achieve cold shutdown conditions within 72 hours and maintain cold shutdown conditions thereafter. Contrary to the above, since implementation of the Sequoyah Fire Protection Program, the licensee failed to maintain all aspects of the approved program. Specifically, in August 2018, the licensee discovered that the sites ability to achieve cold shutdown conditions within 72 hours would be challenged due to an inadequate evaluation of the RHR pumps functionality during certain Appendix R fire scenarios.
05000296/FIN-2018012-01	2018Q3	Browns Ferry	Failure to correct an inoperable 250V Shutdown Board Battery Charger	A self-revealed, Green, NCV of Technical Specifications (TS) 3.8.4 was identified when the licensee failed to correct an inoperable 250V Shutdown Board (SDBD) 3EB Battery Charger on Unit 3. Specifically, in 2014 the 250V SDBD 3EB Battery Charger was entered into the Corrective Action Program (CAP) as a Condition Adverse to Quality (CAQ), but no actions were taken to correct the condition, which led to the component being in inoperable for longer than the allowed outage time defined in TS 3.8.4.
05000327/FIN-2018001-03	2018Q1	Sequoyah	Licensee-Identified Violation	This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Sequoyah Unit 1 and Unit 2 Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (ABGTS), requires two ABGTS trains be operable in modes 1, 2, 3, and 4. Contrary to the above, from March 3-7, 2017, the licensee blocked open door A212 resulting in the inoperability of the auxiliary building secondary containment enclosure boundary and thus inoperability of both trains of the ABGTS. Significance/Severity Level: Green. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building.Corrective Action Reference: CR1269767
05000327/FIN-2018001-01	2018Q1	Sequoyah	Essential Raw Cooling Water Pumps Inoperable due to Frozen Motor Bearing Cooling Lines	A self-revealing Green NCV of Technical Specification 5.4.1, Procedures, was identified when Sequoyah/TVA did not establish, implement and maintain applicable procedures recommended in Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, the essential raw cooling water (ERCW) pump motor maintenance procedure, 0-MI-MRR-067-002.0, Removal/Disassembly/Reassembly Instruction for ERCW Pumps does not contain specific direction for the slope of the motor bearing cooling supply and return lines for the motor reassembly.
05000327/FIN-2018001-02	2018Q1	Sequoyah	Improper Calibration of Reactor Trip Instrumentation Results in a Condition Prohibited by Technical Specifications	A self-revealing Green finding and associated NCV of Sequoyah Unit 1 Technical Specification 5.4, Procedures, was identified on June 25, 2016, when the licensee did not implement procedures to calibrate Delta-T/Tavg Channel IV with the correct test equipment input impedance settings, which resulted in Delta-T/Tavg Channel IV being out of technical specifications allowed tolerances.
05000338/FIN-2018001-01	2018Q1	North Anna	Failure to Assure Service Water Pump Sheds from Emergency Bus upon LOOP or SBO	A self-revealing Greennon-cited violation (NCV) of Technical Specification (TS)5.4.1.a, was identified for the licensees failure to have adequate written procedures for assuring proper configuration control in areas affected by maintenance or plant modifications. Specifically, the licensee failed to detect and correct a disconnected lead from contact C1 on 1-SW-62-1SWEB03. This directly led to the failure of the 1B service water (SW) pump to shed from the 1J emergency bus during performance of maintenance procedure 1-PT-83.2 on March 11, 2018.
05000327/FIN-2017002-01	2017Q2	Sequoyah	Licensee-Identified Violation	Unit 1 and Unit 2 technical specifications LCO 3.7.10 required that if both trains of CREVS become inoperable than LCO 3.0.3 shall be immediately entered. Additionally, LCO 3.0.3 requires both units to be placed in Mode 3 within seven hours if the condition was not rectified. Contrary to the above, on August 10, with both trains of CREVS rendered inoperable, both units remained in Mode 1 for a period of approximately 24 hours. The finding was entered into the licensees CAP as CR 1201905. This finding was assessed using NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, and was determined to be of very low safety significance (Green) due to the finding only representing a degradation of the radiological barrier function provided for the control room.
05000327/FIN-2017002-02	2017Q2	Sequoyah	Licensee-Identified Violation	Unit 1 and Unit 2 facility technical specifications LCO 3.6.10 required two operable EGTS systems in Modes 1 through 4. Contrary to the above, on August 2, 2016,during a system review, plant engineers noted a design flaw that could have resulted in one train of EGTS being rendered inoperable since initial plant operation. This problem was entered into the licensees CAP as CR 1198440 and CR 1200028. The TVA probabilistic risk assessment model does not consider the EGTS in core damage and large early release frequencies. The EGTS system is designed to maintain the shield building at a negative pressure and filter any leakage past the steel liner during a design basis event. With the EGTS inoperable, dose would still remain below 10 CFR 100 limits. The finding was screened using IMC 0609, Appendix A At Power Operation, and was determined to be of very low safety significance (Green). According to Exhibit 3, an issue related to degradation of the radiological barrier function of the reactor building is considered to be of very low safety significance.
05000327/FIN-2017001-01	2017Q1	Sequoyah	Degraded Fire Barrier Penetration	Green . The NRC identified a non -cited violation ( NCV ) of the facilitys operating license for the failure to identify a non functional fire barrier penetration and enter it into the corrective action program (CAP) when the initial damage to the fire barrier occurred. The licensee also failed to implement required compensatory measures for a nonfunctional fire barrier penetration contrary to the approved fire protection report . The licensee entered the issues into their CAP as Condition Report (CR) 1263322 . The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that there was no assurance the fire barrier would prevent the spread of fire through the cable penetration during a design basis fire. The finding was of very low safety significance (Green) due to fully functional automatic suppression systems on either side of the fire barrier . T he inspectors identified a cross- cutting aspect in the identification component of the Problem Identification and Resolution area because the licensee failed to enter the damaged fire barrier into their CAP after it was initially damaged . (P.1)
05000327/FIN-2016004-01	2016Q4	Sequoyah	Degraded Fire Barrier Penetrations	Green. The NRC identified a non-cited violation (NCV) of the facilitys operating license for the licensees failure to ensure that all fire barrier penetrations in fire zones boundaries protecting safety related areas are functional at all times. Specifically, on eight separate fire barrier penetrations, the licensee failed to recognize that the barrier had become damaged to the point of being nonfunctional. The licensee also failed to implement required compensatory measures for a nonfunctional fire barrier penetration contrary to the approved fire protection report (FPR). The licensee entered the issues into their corrective action program (CAP) as Condition Reports (CRs) 1229468, 1229470, 1243550, 1243970, 1243552, 1243554, 1243555, and 1243557. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the fire barriers being damaged to the point of declaring the fire barrier penetrations nonfunctional, there was no assurance that the fire barrier would prevent the spread of fire through the cable penetration during a design basis fire. The inspectors performed the SDP using NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and assigned a High degradation rating, giving no credit for Barrier Protection in accordance with the Fire Barrier Degradation section. The inspectors concluded, that the finding was of very low safety significance (Green) due to fully functional automatic suppression systems on either side of the fire barrier (Question 1.4.3-C). Using Manual Chapter 0310, Aspects Within the Cross-Cutting Areas, the inspectors identified a cross-cutting aspect in the Identification component of the Problem Identification and Resolution area, because the licensee failed to enter the damaged fire barrier into their CAP after it was initially damaged (P.1)
05000327/FIN-2016003-01	2016Q3	Sequoyah	Hydrogen Mitigation System Inoperable Longer than Allowed by Technical Specifications	A self-revealing NCV of Technical Specification 3.6.8, Hydrogen Mitigation System (HMS), was identified for the licensees failure to restore an inoperable train of HMS within the 7 day completion time or place the unit in Mode 3 within the action time of 6 hours. Each train of HMS has 34 hydrogen igniters and SR 3.6.8.1 defines an operable train as one that has at least 33 igniters operable. A review of the operating history revealed the A train HMS had only 31 operable igniters for a period of 91 days due to a mispositioned circuit breaker. Upon discovery of the unexpected condition, the circuit breaker was closed to restore operability to the HMS train. The licensee entered the issue into their CAP as CR 1179126. The licensees failure to preclude an inoperable HMS train for more than 7 days without a subsequent plant shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstones objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity as hydrogen igniters have a high risk significance in ice condenser style containments. The finding was screened to Green based on the fact that the loss of igniters did not affect multiple igniters in adjacent compartments. The inspectors determined that the finding had a cross cutting aspect of Avoid Complacency within the Human Performance area because the licensee failed to implement appropriate error reduction tools while working near the HMS circuit breakers (H.12).
05000327/FIN-2016003-02	2016Q3	Sequoyah	Isolation of Fire Suppression System to a Significant Portion of the Plant Site	A self-revealing non-cited violation (NCV) of the facility operating licenses DPR-77 and DPR-79 conditions 2.C.(16) and 2.C.(13), respectively, was identified for the licensees failure to properly implement the clearance process such that the fire suppression system was rendered non-functional for approximately 41 hours. The licensee inappropriately expanded an existing clearance on March 29, 2016 in order to attempt to reduce boundary valve leakage affecting existing maintenance on the fire suppression system within a valve pit. Subsequently on March 30, 2016 during fire system testing, technicians noted a lack of system pressure and it was ultimately concluded the clearance expansion had inadvertently isolated fire suppression water to a significant portion of the site. Upon discovery of the clearance error, the system was restored to a functional status. The licensee entered the issue into their corrective action program (CAP) as CR 1155763. The licensees failure to properly assess the system impact of a clearance revision for the High Pressure Fire Protection (HPFP) suppression header and enter the required FPR Operating Requirement (FOR) Action was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inability to pressurize the HPFP system from either the electric or diesel-driven fire pumps rendered the fire suppression system inoperable. Based on the complexities of this particular event, the inspectors concluded that Appendix M, Significance Determination Process Using Qualitative Criteria, of IMC 0609 should be performed in lieu of a Phase 3 analysis. Under appendix M, the Senior Reactor Analyst (SRA) performed an initial bounding evaluation using qualitative methods. The licensee submitted a detailed analysis that estimated an upper bound for the risk of the finding which was less than 1E-6. The SRA performed a review of this screening analysis as part of this SDP evaluation. In addition to the SRA review, the resident inspectors performed an independent review of the licensees estimation of the success of actions used to recover the isolated fire header. To the extent reviewed, the methodology and results were determined to be acceptable for use in this SDP review of this Performance Deficiency. The SRA concurred with the submitted results of the licensees screening analysis, and has determined the finding to be GREEN. The inspectors determined that the finding had a cross cutting aspect of Procedural Adherence within the Human Performance area, because the licensee failed to consider the affect that changing a clearance order could have on the operability of the fire suppression system.
05000425/FIN-2016002-01	2016Q2	Vogtle	Failure to properly implement a maintenance procedure caused a Reactor Trip	A self-revealing non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly implement procedure 24750- 2, Steam Generator Level (Narrow Range) Protection Channel II 2L-519 Channel Operational Test and Channel Calibration. During testing of Unit 2 loop 1 steam generator (S/G) narrow range channel 2L-519 the channel was not removed from scan resulting in a reactor trip. The licensees immediate corrective actions were to remove the technicians performing the calibration from maintenance duties for formal remediation. The licensee documented this condition in CR 10230073. The performance deficiency (PD) was more than minor because it adversely affected the Initiating Events cornerstone objective in that the failure to properly remove channel 2L-519 from scan resulted in a reactor trip. The finding was determined to be Green because the PD did not result in a loss of mitigation equipment used to transition the reactor to a stable shutdown condition. The finding was assigned a cross cutting aspect of Avoid Complacency because maintenance technicians failed to implement appropriate error reduction tools to verify that the correct channel was removed from scan for testing.
05000327/FIN-2016002-01	2016Q2	Sequoyah	Isolation of Fire Suppression System to a Significant Portion of the Plant Site	A self-revealing apparent violation (AV) of the facility operating licenses DPR-77 and DPR-79 conditions 2.C.(16) and 2.C.(13) was identified for the licensees failure to properly implement the clearance process such that the fire suppression system was rendered non-functional for approximately 48 hours. The licensee inappropriately expanded an existing clearance on March 29 in order to attempt to reduce boundary valve leakage affecting existing maintenance on the fire suppression system within a valve pit. Subsequently, on March 30, during fire system testing, technicians noted a lack of system pressure and it was ultimately concluded the clearance expansion had inadvertently isolated fire suppression water to a significant portion of the site. Upon discovery of the clearance error, the system was restored to a functional status after being isolated for approximately 48 hours. The licensee entered the issue into their corrective action program as condition report (CR) 1155763. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inability to pressurize the high pressure fire protection (HPFP) system from either the electric or diesel-driven fire pumps rendered the fire suppression system inoperable. The finding could not be screened to Green and is pending a significance determination. The inspectors determined that the finding had a cross-cutting aspect of Procedural Adherence within the Human Performance area, because the licensee failed to consider the effect that changing a clearance order could have on the operability of the fire suppression system. (H.8).
05000327/FIN-2016001-01	2016Q1	Sequoyah	Inadequate Application of Flame Retardant on Cable Room Penetrations	The NRC identified a non-cited violation (NCV) of Unit 1 and 2 Technical Specification 5.4.1 for the licensees failure to adequately implement fire protection procedures. Specifically, the inspectors identified several cables located within a cable tray that penetrated the floor of the cable spreading room that were not adequately coating with fire retardant material as required by plant procedures. The licensee placed the issue into the corrective action program (CAP) and implemented a fire watch for the degraded condition. The inspectors determined that the failure to adequately implement all requirements of the licensees fire protection program procedures was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding was of very low safety significance (Green) because of the fire protection defense in depth concept provided other barriers to prevent the spread of fires. The cause of this finding was related to the procedural adherence component of the human performance area, because the licensee failed to properly install cable bundles through wall penetrations.
05000327/FIN-2016001-02	2016Q1	Sequoyah	Inadvertent Safety Injection Due to Inadequate Main Steam Procedure	A self-revealing NCV of Units 1 & 2 Technical Specification, 5.4.1 was documented for the licensees failure to implement an adequate procedure associate with the startup of the main steam system. Specifically, the licensee caused an inadvertent safety injection which unnecessarily challenged the operators due to an inadequate draining of the main steam header during system start up. The licensee placed the issue into the CAP. The failure of the licensee to adequately drain condensate from the main steam header resulted in an inadvertent safety injection (SI) and was a performance eficiency. The finding was determined to be greater than minor because it adversely effected the Procedure Quality attribute of the Initiating Events Cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of this finding was evaluated in accordance with the Manual Chapter 0609 Appendix A, The Significance Determination Process for Findings At-Power. Although the unit was in Mode 3 at the time, this appendix was chosen because the plant did not meet the entry conditions for residual heat removal system operation. The inspectors concluded that the finding was of very low safety significance (Green) because no significant initiating event prompted this transient. The finding was determined to have a cross-cutting aspect in the operating experience component of the problem identification and resolution area, because the licensee failed to evaluate and implement relevant internal and external operating experience.
05000327/FIN-2015004-01	2015Q4	Sequoyah	Failure to Recognize and Submit for Approval a Reduction in Effectiveness of the Emergency Plan	The inspectors identified a Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations, Part 50.54(q), for changes to the licensees radiological emergency plan, effective December 18, 2014, that reduced the effectiveness of the plan and therefore, should have received NRC approval prior to making the change. Specifically, the effectiveness of TVAs Radiological Emergency Plan (Generic Part), Revision 104, was reduced by the inadvertent removal of the offsite telephone communications description for the Health Physics Network and Emergency Notification System communication tools, as well as the monthly testing of those devices. The licensees failure to recognize that Revision 104 reduced the effectiveness of the emergency plan was a performance deficiency. The licensee entered this issue into their corrective action program (CAP) as Condition Report (CR) 1093684 This finding is more than minor because it brings into question the thoroughness of the licensees review process when making changes to the emergency plan and adversely affects the procedure quality attribute of the emergency preparedness cornerstone objective. This finding is a violation of NRC requirements and because it has the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with IMC 0612, Appendix B. This finding is determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involves the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan is reduced.
05000259/FIN-2015007-01	2015Q4	Browns Ferry	Failure to Specify Adequate Instrument Ranges for MSIV Leakage Testing	A NRC identified NCV of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XI, Test Control, was identified for the failure to specify adequate test instrumentation for performing MSIV leak rate testing. Specifically, the licensee test procedure allowed the use of high range test instruments to measure low leakage rates while performing the combined leak rate testing on the Unit 1 B Main Steam Line. This resulted in instrument uncertainties large enough to impact the validity of the test results. The licensee immediately entered this issue into their corrective action program as CR 1117381. The licensee performed an evaluation and determined that the latest test results provided reasonable assurance of operability. This performance deficiency was more than minor because if left uncorrected had the potential to lead to a more significant safety concern by masking the failure to meet test acceptance criteria. The finding was screened for significance using the Barrier Integrity cornerstone column of IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, dated 7/1/2012, and IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated 7/1/2012, and was determined to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding was assigned a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not initiate a corrective action to identify the cause of the negative leak rate results obtained during the recent performance of the test procedure (P.1).
05000259/FIN-2015007-02	2015Q4	Browns Ferry	Failure to develop a PM schedule that specified inspection of the EDG neutral grounding resistor	A NRC-identified non-cited violation (NCV) of Technical Specifications (TS) 5.4.1 was identified for the failure to develop a preventive maintenance (PM) schedule that specified inspection of the Emergency Diesel Generators (EDG) neutral grounding resistor as recommended by Regulatory Guide (RG) 1.33, 9.b. Specifically, procedures failed to provide proper guidance to maintain the grounding resistor in accordance with design basis as described in the UFSAR and electrical calculations. Upon identification of the issue, the licensee performed a visual inspection of the resistor and determined that it was functional based on no signs of physical degradation or damage. The licensee entered this issue into the corrective action program (CAP) as CR1114779 to evaluate and implement appropriate corrective actions. This performance deficiency was more than minor because if left uncorrected it could result in a more significant safety concern. Specifically, lack of inspections of the secondary grounding resistor could allow for an undetected condition which would cause transient voltages capable of damaging safety related equipment. The finding was screened for significance using the Mitigating Systems cornerstone column of IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, dated June 19, 2012, and was determined to be of very low safety significance (Green) using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At- Power, dated June 19, 2012, because the finding affected the design or qualification of a Mitigating SSC, and the SSC maintained its operability as documented in CR 1114779. No cross-cutting was assigned because it is not indicative of current licensee performance.
05000259/FIN-2015007-04	2015Q4	Browns Ferry	Licensee-Identified Violation	Technical Specification 5.4.1, required, in part, that written procedures shall be established, implemented, and maintained covering the following activities: a. the applicable procedures recommended in Regulatory Guide (RG) 1.33, Rev 2, Appendix A. Procedures recommended in Appendix A to RG 1.33, included procedures for performing maintenance, and specifically, preventive maintenance schedules should be developed to specify...inspections of equipment. 0-TI-522, Program for Implementing NRC Generic Letter 89-13 required in part, in section 7.2, that BFN will maintain an inspection and cleaning program in accordance with the BFN PM Program to verify the heat transfer capability of the safety related Heat Exchangers cooled by EECW an RHRSW, and the PMs provide for reassessing this inspection frequency based on the results of inspections, not to exceed 5 years. Contrary to this requirement, since January 22, 2013, the licensee did not implement their PM schedule for inspections of the 3C RHR HX appropriately, because they allowed the PM to extend beyond the maximum of 5 years. Consequently, when the heat exchanger was opened, it failed the acceptance criterion of no more than 77 tubes plugged. This finding was not greater than very low safety significance (Green) because it was a deficiency affecting the design of a Mitigating SSC, and the SSC maintained its operability or functionality (as demonstrated by past operability evaluation for PERs 750848 and 750858). The licensee entered this issue into their CAP as CR 674040.
05000259/FIN-2015007-03	2015Q4	Browns Ferry	Procurement of Electrical Equipment for Ungrounded Electrical Systems	The team identified an unresolved item (URI) related to the licensees procurement of electrical equipment for ungrounded electrical systems. The 480 VAC system for each unit consists of 480-V Load Center Unit Substations with each substation consisting of 4160-480-V transformers, primary terminal box, and close-coupled or bus duct connected 480-V, metal-enclosed switchgear. The 480-VAC distribution system is three-phase ungrounded. Each substation bus is normally fed from its own transformer, with an alternate source consisting either of an adjacent 480-VAC bus section or of another transformer serving as standby. Ventilated dry-type transformers are three-phase, delta-delta configuration so that the 480 VAC system is ungrounded. Ungrounded systems are susceptible to overvoltage conditions resulting from a single line to ground fault. A line to ground fault will result in a sustained higher voltage to ground on the ungrounded phases. Industry standard IEEE 242 (Buff Book) Protection and Coordination of Industrial and Commercial Power Systems, section 8.2.5 Ungrounded Systems stated if this ground fault is intermittent or allowed to continue, the system could be subjected to possible severe overvoltages to ground, which can be high...(cause line to ground voltages several times normal voltage on all three phases). Because of the potential for overvoltage conditions, specifications, purchase orders or procurement documents for equipment such as motors, cables, and switchgear should identify that the equipment is intended for use on an ungrounded system. The team requested the original specifications for the installed BFN safety-related motors BFN-2-MTR-068-0003; 2-FCV-68-3 (Recirc Pump 2A Disch VLV) fed from the 480V Reactor MOV Board 2E and, BFN-3-MTR-073-0002; 3-FCV-73-2 (HPCI Steam Line INBD Isolation VLV) fed from the 480V Reactor MOV Board 3A to determine if the intended service condition as a 480 VAC ungrounded system was appropriately identified. The team reviewed Procurement Engineering Group packages CRP205J - PO 733602 and CFK570P PO 836093 for the safety-related motors and determined that the ungrounded system requirement was not identified. Equipment intended for service on ungrounded systems is designed to withstand the sustained higher line to ground voltages than can occur on grounded systems. These insulation systems are not typically provided unless the purchaser specifies an ungrounded system. Industry standard NEMA MG 1 Motors and Generators, section 14.31 Machines Operating On An Ungrounded System stated: Alternating-current machines are intended for continuous operation with the neutral at or near ground potential. Operation on ungrounded systems with one line at ground potential should be done only for infrequent periods of short duration, for example as required for normal fault clearance. If it is intended to operate the machine continuously or for prolonged periods in such conditions, a special machine with a level of insulation suitable for such operation is required. The motor manufacturer should be consulted before selecting a motor for such an application. The NRC will review the licensees responses to follow-up questions asked during a conference call with the licensee on January 21, 2016. Based on this future review, the NRC will make a determination if the licensee properly procured electrical components for ungrounded systems. More information is needed to determine if more than a minor performance deficiency or violation exists associated with this issue, thus a URI is being opened.
05000327/FIN-2015003-01	2015Q3	Sequoyah	Inadequate Clearance Causes damage to A train SSPS	A self-revealing Green NCV of Unit 1 Technical Specification (TS) 6.8.1.a was identified for the licensees failure to adequately establish a clearance boundary during plant maintenance. Specifically, the licensee caused damage to a safety-related component during maintenance as a result of a failure to de-energize all electrical sources during maintenance troubleshooting activities. The licensee placed the issue into their corrective action program (CAP) and corrected the identified deficiencies. The inspectors determined that the failure to adequately implement clearance procedures was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding was of very low safety significance (Green) as the affected safety significant component was repaired within 24 hours. The cause of this finding was related to the cross-cutting aspect of leaders ensuring that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety.
05000327/FIN-2015003-02	2015Q3	Sequoyah	Failure to Implement Work Risk Activity and Oversight of Supplemental Personnel Procedures	A self-revealing Green NCV of TS 6.8.1.a, Administrative Controls of Procedures and Programs, was identified for the licensees failure to implement procedures related to quality during the surveillance capsule relocation activity. Specifically, procedures NPG-SPP-07.3, Work Activity Risk Management, and NPG-SPP.07.7, NPG TCM Role and Oversight of Supplemental Personnel, were not appropriately implemented. The deficiency was entered into the licensees CAP as Problem Evaluation Report (PER) 1016839. This finding was determined to be greater than minor because it was associated with the Human Performance attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the failure to properly secure reactor vessel surveillance capsules and the subsequent damage to the reactor vessel pressure boundary, reactor internals and fuel filter screens. The proper higher risk categorization would have led to enhanced contractor oversight, and the ability to detect when the contractors were performing actions outside the approved procedure. These additional oversights would reasonably be expected to prevent the events that led to the surveillance capsule ejections, and eliminate any potential to cause damage to the reactor vessel pressure boundary, reactor internals, and fuel filter screens. The inspectors identified a cross-cutting aspect in the Human Performance Consistent Process cross-cutting area. Specifically, the licensee failed to consistently incorporate risk insights, as required by procedure NPG-SPP-07.3, which resulted in less than conservative classification for an infrequently performed activity inside the reactor vessel performed by contract personnel.
05000327/FIN-2015002-02	2015Q2	Sequoyah	Spilled Specimen Capsule	The inspectors identified an unresolved item (URI) associated with the control of specimen capsules inside the reactor vessel. During this refueling outage, the licensee noted that two specimen capsules had become dislodged from their location on the core barrel. This particular outage required a 10 year ISI inspection and the core barrel was removed as part of the outage plan. The two capsules were noted to have been moved during the last refueling outage. A formal root cause evaluation was performed. The root cause team included industry experts independent of the licensee organization. The root cause team noted several procedural violations during the previous capsule move. The team concluded that these significant errors led to the improper seating of the capsules in the specimen baskets and ultimately allowed the capsules to dislodge from the core barrel. A significant foreign object retrieval evolution was completed during the outage and the core barrel, lower internals, and lower vessel head were inspected by the licensee and the NRC. The specimen parts were collected and placed in storage containers and transferred to the spent fuel pool. The unit was restarted on May 15 without a full accountability of the specimen parts. The inspectors determined that more inspection of this issue is required in order to understand all aspects of the incident. This issue will be tracked as URI 05000327/2015002-02, Spilled Specimen Capsule.
05000328/FIN-2015002-01	2015Q2	Sequoyah	Failure to Adequately Follow Foreign Material Control Procedures	A self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to follow a foreign material exclusion procedure and precluded foreign material from entering the safetyrelated Essential Raw Cooling Water (ERCW) system. This resulted in wood debris within the ERCW that ultimately migrated to the 2B2 emergency diesel heat exchanger. Immediate corrective actions included removal of the foreign material and the performance of an engineering analysis to ensure the wood debris did not affect the system operability. The licensee placed this issue into their corrective action program as CR 1033792. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency allowed a piece of wood to enter into the 2B2 emergency diesel heat exchanger blocking 11 tubes. Because the finding was a deficiency affecting the design of a mitigating structure, systems, or components (SSCs) that was confirmed not to have resulted in the loss of operability, it was determined to be of very low safety significance (Green). The inspectors determined that no cross-cutting aspect will be assigned to this performance deficiency since it occurred in 2008 and is therefore not indicative of current licensee performance.
05000327/FIN-2015002-03	2015Q2	Sequoyah	Licensee-Identified Violation	Unit 1 Technical Specifications Section 3.3.1.1 requires two channels of intermediate range nuclear instrumentation in Mode 2 to provide the input to the P-6 interlock. This interlock allows the operator to block the source range channels during a reactor startup. This is done to prevent damage to the detectors as power is elevated to levels that could damage the detector. Contrary to the above, between March 11 and March 27, the P6 interlock was inoperable due to a non-conservative bias, concurrent with Unit 1 being in Mode 2 on March 14 from 0558 to 1010. This problem was entered into the licensees corrective action program as CR 1005422. The finding was screened using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, and was determined to be of very low safety significance (Green).
05000327/FIN-2015001-01	2015Q1	Sequoyah	Failure to Follow Procedure Results in an Inadvertent Sprinkler Deluge in the Cable Spreading Room	A self-revealing Green non-cited violation (NCV) of Technical Specification (TS) 6.8.1.f, Fire Protection Program Implementation, was identified for the licensees failure to follow a fire protection procedure. Specifically, the licensee failed to isolate the fire main from the cable spreading room (CSR) header during testing as required by procedure. This resulted in pressurization of the fire header to the cable spreading room which then caused a rupture of one of the sprinkler heads in the room. The licensee entered this issue into their corrective action program (CAP) as problem evaluation report (PER) 1001695. As immediate corrective actions, the licensee replaced the failed sprinkler head and conducted a formal review of the incident. The finding was determined to be more than minor because it was associated with the human performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the excessive amount of water sprayed in the CSR increased the likelihood of a plant transient due to the potential impact on non-waterproof junction boxes located in the CSR as well as safety-related instrument racks located in the auxiliary instrument room (AIR) directly below the CSR. Using Appendix A, Exhibit 1, Initiating Events Screening Questions, the finding was determined to be of very low safety significance because the deficiency did not cause a reactor trip nor a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding was determined to have a cross-cutting aspect in the avoid complacency component of the human performance area (H.12), because the technicians failed to properly implement appropriate error reduction techniques while performing a fire protection procedure.
05000327/FIN-2014005-01	2014Q4	Sequoyah	Licensee-Identified Violation	Licensee Event Report (LER 2-2014-001-00) was submitted following the licensees discovery that Sequoyah Unit 2 operated in a condition prohibited by Technical Specifications (TS). During mode 1 operation, train B containment purge had been operated without the minimum required operable radiation monitoring channels. TS Limiting Condition for Operation (LCO) 3.3.3.1 and TS LCO 3.3.2 required with less than the minimum channels operable, plant operation may continue provided the containment purge supply and exhaust valves are maintained closed. Contrary to the above during plant operations from April 8 and April 28, 2014, train B containment purge supply and exhaust valves were opened to place purge in service three times with containment purge air exhaust radiation monitors (SQN-2-RM-090-130 and SQN-2-RM-090-131) being incorrectly aligned to train A purge. This was licensee identified and entered into the CAP as PER 878321. The finding was screened using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 3 Barrier Integrity Screening Questions and determined the finding to be of very low safety significance (Green) because it does not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components, or involve an actual reduction in function of hydrogen igniters.
05000327/FIN-2014004-03	2014Q3	Sequoyah	Licensee-Identified Violation	TS 3.9.4.c requires in part that during fuel movement, all containment penetrations shall be closed or capable of being closed by an automatic valve. In addition, TS 3.9.4.c allows exceptions to this requirement for penetrations that traverse to the auxiliary building secondary containment enclosure (ABSCE) where these penetrations may be open under administrative controls during fuel movement. Contrary to the above, on several instances between 2000 and 2014, the licensee opened penetrations (between containment and ABSCE) during fuel movement without adequate administrative controls in place. The finding was considered more than minor because it was associated with the Barrier Integrity cornerstone and affected the cornerstones ability to preserve the containment boundary. The inspectors determined that, although the finding involved a violation of the containment control, TS 3.9.4, the finding did not: 1) involve a loss of reactor coolant system (RCS) inventory; 2) degrade ability to terminate a leak path or add RCS inventory as needed; or 3) degrade the ability to recover RHR once it was lost. Therefore, according Appendix G, the finding did not require a quantitative (phase 2 or 3) analysis. Findings in the shut-down condition that do not require a quantitative analysis are considered to be of very low safety significance (Green). This issue was entered into the CAP as PER 886970. Note that this issue is also discussed under Section 4OA3 of this report as it involved a LER.
05000285/FIN-2014009-12	2014Q3	Fort Calhoun	Failure to Maintain Effectiveness of an Emergency Plan	A cited violation of 10 CFR 50.54(q)(2), Conditions of License, was identified involving the failure to maintain the effectiveness of the sites emergency plan. Specifically, the licensee established an Alert low river level emergency classification criteria that was below the raw water pumps minimum suction requirements, contrary to the standard emergency action level scheme. The licensee entered this issue into its corrective action program as Condition Report 2014-08757 which included actions to re-evaluate the capability of the raw water pumps to operate at low river levels. This finding was more than minor, and therefore a finding, because it was associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone and affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, inaccurate emergency actions levels degrade the licensees ability to implement adequate measures to protect public health and safety. The finding was evaluated using the Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance (Green) because the finding was not a lost or degraded risk significant planning function. The planning standard function was not degraded because the emergency classifications would have been declared although potentially in a delayed manner. This finding has a cross-cutting aspect in the area of human performance in that the licensee did not ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee deferred funding for a vendor analysis of the capabilities of the raw water pumps at the design low river level.
05000285/FIN-2014009-24	2014Q3	Fort Calhoun	Licensee-Identified Violation	Title 10 CFR 50.73(a)(1) requires, in part, that licensees shall submit a licensee event report for any event of the type described in this paragraph within 60 days after the discovery of the event. Contrary to the above, on February 5, 2012, November 15, 2011, and February 19, 2013, the licensee failed to submit a licensee event report for an event meeting the requirements for reporting specified in 10 CFR 50.73. Specifically, the licensee submitted Licensee Event Reports 2012-013, 2012-015 and 2013-001 greater than 60 days following discovery of a reportable event. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV, non-cited violation. The licensee entered this issue into their corrective action program as CR 2014-02792.
05000285/FIN-2014009-13	2014Q3	Fort Calhoun	Failure to Perform Evaluation for Design Change	A cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, was identified involving the failure to evaluate if a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the licensee did not evaluate a change that would permanently substitute a manual action for an automatic action to add water and nitrogen gas to the component cooling water surge tank. The licensee entered this issue into its corrective action program as Condition Report 2014-09080 and initiated action to evaluate the change to the component cooling water system. The NRC determined that the licensees failure to perform an evaluation prior to implementing a proposed change described in the Updated Safety Analysis Report was a violation of 10 CFR 50.59. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC evaluated the performance deficiency using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the team evaluated this finding using the significance determination process to assess its significance. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy this performance deficiency is being characterized as a Severity Level IV violation. The team determined that a cross-cutting aspect was not applicable to this finding because the issue was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-01	2014Q3	Fort Calhoun	Failure to Initiate Condition Reports for Gaps Identified in Resolving NRC Non-Cited Violations	A non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, was identified involving the failure to follow procedures to initiate condition reports to enter conditions adverse to quality into the corrective action program. Specifically, the licensee failed to initiate condition reports in accordance with Procedure FCSG 24-1, Condition Report Initiation, Step 4.1.1.G, when deficiencies related to the stations corrective actions implemented for NRC violations were identified. The licensee entered this issue into its corrective action program as Condition Report 2014-09063 and initiated action to write condition reports for identified gaps related to previous NRC violations. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it would have the potential to lead to a more significant safety concern. The team performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding was of very low safety significance (Green) because it did not involve a loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of human performance because the licensee elected to use an informal system to resolve these issues rather than the corrective action program.
05000285/FIN-2014009-08	2014Q3	Fort Calhoun	Failure to Report Loss of Environmental Qualification of Safety Related Limit Switches within Required Time Limits	A non-cited violation of 10 CFR 50.73(a)(1), Licensee Event Report System, was identified involving the failure to submit a required licensee event report. Specifically, the licensee failed to report within 60 days the discovery that NamcoTM Type EA 180 limit switches were not environmentally qualified as required due to inadequate maintenance procedures, a condition that resulted in operation prohibited by the plants technical specifications. The licensee restored compliance by submitting Licensee Event Report 05000285/2014-004 on June 20, 2014. The licensee entered this issue into its corrective action program as Condition Report 2014-08454. The NRC determined that the failure to submit a licensee event report within the time limits specified in regulations was a violation of 10 CFR 50.73. This violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV, non-cited violation. The NRC determined that a cross-cutting aspect was not applicable because the issue was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-04	2014Q3	Fort Calhoun	Failure to Perform an Evaluation for a New Operator Manual Action to Refill Component Cooling Water System During Post- Accident Conditions	A non-cited violation of 10 CFR 50.59, Changes, Test, and Experiments, was identified involving the failure to evaluate if a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the licensee failed to evaluate if a change implemented under Engineering Change 59252 that credited the non-safety related demineralized water system as a make-up source to the component cooling water system during post-accident conditions represented an adverse change to the Updated Safety Analysis Report described design function. The licensee entered this deficiency into its corrective action program for resolution as Condition Report 2014-09151 and established action items to update Engineering Change 59252. The NRC determined that the licensees failure to perform an evaluation prior to implementing a proposed change described in the Updated Safety Analysis Report was a violation of 10 CFR 50.59. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the NRC evaluated the violation using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the NRC evaluated this finding using the significance determination process to assess its significance. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this performance deficiency is characterized as a Severity Level IV violation. The team determined that a cross-cutting aspect was not applicable because the issue involving the failure to perform an adequate 10 CFR 50.59 evaluation was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-25	2014Q3	Fort Calhoun	Licensee-Identified Violation	Technical Specification 5.8.1.a, requires, in part, that written procedures be established, implemented, and maintained as recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Paragraph 9.a, requires that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, the licensee failed to establish procedures for maintenance that can affect the performance of safety related equipment as recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, prior to May 3, 2013, the licensees maintenance procedure for NamcoTM Type EA 180 limit switches did not specify the correct torque values for the switch top cover to maintain the components environmental qualifications. This finding was determined to be of very low safety significance because the affected limits switches only affected the radiological barrier provided for by the control room. This issue was entered into the licensees corrective action program as CR 2012-03651.
05000285/FIN-2014009-23	2014Q3	Fort Calhoun	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, on June 2, 2008, the licensee completed flow scan valve testing for the high pressure safety injection alternate header isolation valve (HCV-2987) that showed a much higher stem friction value than previously analyzed, but failed to promptly identify and correct the condition adverse to quality until CR 2012-01601 was initiated on February 29, 2012. This finding is of very low safety significance (Green) because valve HCV-2987s failure did not represent an actual loss of safety function of a single train for greater than the technical specification allowed outage time in that EOP/AOP Attachments, Revision 13, dated November 19, 2002, requires operators to also close downstream valves that would back up the closure function of valve HCV-2987. This issue was entered into the licensees corrective action program as CR 2012-01601.
05000285/FIN-2014009-22	2014Q3	Fort Calhoun	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from initial construction until January 13, 2013, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to control the design inputs to ensure that piping in the chemical and volume control system would perform acceptably during a seismic event. This finding is of very low safety significance (Green) because a chemical and volume control system piping failure event is enveloped by the small break loss of coolant accident as described in Updated Safety Analysis Report Section 14.5.5. This issue was entered into the licensees corrective action program as CR 2013-01796.
05000285/FIN-2014009-21	2014Q3	Fort Calhoun	Failure to Take Timely Corrective Actions for an Unsealed Raw Water System Control Panel	A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take corrective actions to address a design deficiency affecting the control panel for raw water strainer AC-12B. Consequently, the panel experienced a water intrusion event on August 3, 2014, resulting in an unplanned inoperability of the raw water system. Following identification of this issue, the licensee implemented corrective actions to seal conduits leading to control panel AI-348 to prevent future water intrusion. The licensee entered this issue into its corrective action program as Condition Report 2014-09572. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of problem identification and resolution in that the licensee failed to adequately review and provide timely responses to past operating experience that demonstrated that panel AI-348 was susceptible to water intrusion.
05000285/FIN-2014009-20	2014Q3	Fort Calhoun	Failure to Correct Conditions Adverse to Quality in the Diesel Generator Stating Air System	A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective actions to address service life related degradation of the emergency diesel generator starting air system. As a result, diesel generator 1 failed to roll during planned surveillance testing due to a degraded diesel starting air valve. The licensee replaced the faulty starting air valve and implemented corrective actions to develop preventative maintenance strategies for the starting air system. The licensee entered this issue into the corrective action program as Condition Report 2014-09424. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings , Exhibit 3, Mitigating Systems Screening Questions, dated May 9, 2014, the finding was of very low safety significance (Green) because the finding does not represent a loss of system safety function and the finding does not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to recognize and plan for the possibility of latent issues, and inherent risk, even while expecting successful outcomes when determining the repair schedule for starting air valve SA-148.
05000285/FIN-2014009-18	2014Q3	Fort Calhoun	Failure to Complete Corrective Actions in a Timely Manner	A non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective actions to address deficiencies in station calculations. Specifically, the licensee failed to update station calculations to incorporate actual test data for sluice gate leakage to ensure design basis flood levels do not adversely affect equipment important to safety. The licensee entered this issue into its corrective action program as Condition Report 2014-09156 and initiated actions to update station calculations. This finding was more than minor, and therefore a finding, because if left uncorrected, the finding would have the potential to lead to a more significant safety concern. Specifically, failure to complete accurate calculations that support engineering modifications for mitigating the consequences of an external flooding event could lead to unanalyzed conditions adversely affecting safety related systems or components. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed prioritize an update to Calculation FC08081 following completion of the May 2013 in-leakage test.
05000327/FIN-2014004-02	2014Q3	Sequoyah	Licensee-Identified Violation	Unit 1 SR 4.7.3.b requires in part that at least once per 18 months, each CCS pump start automatically on a SIS signal. Contrary to the above, the C-S (swing) CCS pump was never tested to automatically start using a Unit 1 SIS signal. The licensee considered this to be a never performed surveillance rather than simply a missed surveillance since no testing records since initial operation in 1980 could be located. The pump had been successfully tested every 18 months from the Unit 2 SIS as this pump is normally powered from Unit 2. As an alternate and abnormal lineup, the pump could be powered from Unit 1 and thus the requirement for start testing using the Unit 1 SIS. This finding was considered more than minor because it was associated with the mitigating system cornerstone and affected the cornerstones reliability due to the failure to fully test the CS CCS pump for over 30 years. The finding was considered of very low safety significance as it remained operable and available during the affected period due to successfully passing the initial surveillance test performed on January 13, 2014. The issue was entered into the CAP as PER 826482. Note that this issue is also discussed under Section 4OA3 of this report as it involved a LER.
05000285/FIN-2014009-02	2014Q3	Fort Calhoun	Multiple Examples of Failure to Evaluate Operability of Degraded or Non-Conforming Condition	Multiple examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to follow Procedure OP-FC-108-115, Operability Determinations, Revision 0a. In each example, the team identified that the licensee failed to make an immediate determination of operability for a degraded or non-conforming condition or failed to make an immediate determination of operability based on a detailed examination of the deficiency. The licensee took immediate corrective actions to update the incomplete or inaccurate operability determinations and entered the collective failures to follow station operability procedures into their corrective action program as Condition Report 2014-09163. This performance deficiency was more than minor, and therefore a finding, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective of ensuring the reliability of systems that respond to initiating events. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to use decision-making practices that demonstrate that a proposed action is to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee made non-conservative decisions related to the impact of degraded or non-conforming conditions.
05000285/FIN-2014009-03	2014Q3	Fort Calhoun	Failure to Adequately Perform an Operability Evaluation and a 50.59 Evaluation	A non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to evaluate and implement adequate compensatory measures for a degraded condition associated with raw water pump AC-10C. Specifically, the licensees operability determination established a compensatory measure to place pump AC-10C in pull-to-lock, contrary to the system single failure analysis design criteria described in the Updated Safety Analysis Report. The licensee entered this issue into its corrective action program as Condition Reports 2014-09104 and 2014-08515 and performed an operability evaluation and associated 10 CFR 50.59 evaluation that used an acceptable compensatory measure to pump water from affected manholes prior to affecting the degraded power feeder cable for raw water pump AC-10C. The NRC evaluated this performance deficiency as both a reactor oversight process finding and a traditional enforcement violation. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of problem identification and resolution with an aspect of evaluation because the licensee failed to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2). In addition, because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function in that the failure to obtain a license amendment for a change that could result in a malfunction of a structure, system or component with a different result than previously evaluated in the Updated Safety Analysis Report is in violation of 10 CFR 50.59(c)(2)(vi), the NRC also evaluated the violation using traditional enforcement. Since this violation is associated with a Green reactor oversight process violation, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.
05000285/FIN-2014009-05	2014Q3	Fort Calhoun	Inadequate Design Inputs into Safety Injection Piping Stress Calculation	A non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to implement appropriate design control measures associated with a safety-related pipe stress calculation. Specifically, several unverified and potentially non-conservative inputs were identified associated with Calculation FC07240 used to analyze stresses on a pipe reduction tee in the safety injection system. The licensee entered this issue into the corrective action program as Condition Report 2014-09098 and initiated action to update Calculation FC07240. This performance deficiency was more than minor, and therefore a finding, because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of components that respond to initiating events. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to apply the appropriate rigor when evaluating the overstressed pipe union tee.
05000285/FIN-2014009-06	2014Q3	Fort Calhoun	Failure to Maintain Design Control of Raw Water Strainer Control Panel	A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to maintain design control of the raw water strainer AC-12B control panel AI-348. Specifically, the licensee failed to adequately design control panel AI-348 to protect it from the effects of spraying and wetting as required by the plants licensing and design basis. The licensee entered this issue into its corrective action program as Condition Reports 2013-03301 and 2014-06974 and initiated action to encase control panel AI-348 to protect it against the effects of spraying and wetting. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, control panel AI-348 was not designed to prevent water intrusion that resulted in a loss of power to raw water strainer AC-12B. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the organization thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000328/FIN-2014004-01	2014Q3	Sequoyah	Failure to Perform Adequate Maintenance on Containment Vacuum Relief Valve	A Self-revealing Green Non-Cited Violation (NCV) of Technical Specification (TS) 6.8.1.a. was identified for the licensees failure to adequately implement a maintenance procedure associated with a vacuum relief containment isolation valve. Specifically, during a refueling outage on May 24, 2014, the licensee failed to properly install a locking wire associated with the spring tension bolts on the Unit 2 containment vacuum relief valve. This error ultimately led to a failure of the valve on June 24 at 1600 and entry into TS 3.6.3, Containment Isolation Valves. The valve was ultimately repaired and the valve was declared operable on June 26 at 0026. The inspectors determined that the licensees failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency. This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstones objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible. The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is Green. The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. (H.1).
05000285/FIN-2014009-07	2014Q3	Fort Calhoun	Failure to Accurately Model Flow Path for External Flood Mitigation	A non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to accurately model cell level control of river water during external flooding events. Specifically, the licensee failed to account for losses due to the physical obstructions of trash racks for inflowing river water, the decreased withdrawal rate of the raw water pumps due to fouling across the traveling screens, and a bounding in leakage rate for the sluice gates when the river level is at maximum level of 1014 mean sea level and the intake cell levels are at minimum level of 9769 . The licensee entered this issue into its corrective action program as Condition Report 2014-09155, performed an operability determination, and initiated action to update station calculations related to intake cell level control. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the finding would have the potential to lead to a more significant safety concern. Specifically, the failure to accurately model flow in and out of the cells could adversely affect the external flooding mitigation strategy beyond previously identified equipment capacities and operator actions. This finding was associated with the Mitigating Systems Cornerstone. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, in that the licensee failed to incorporate relevant internal operating experience related to previous NRC inspection into Calculation FC08081.
05000285/FIN-2014009-11	2014Q3	Fort Calhoun	Failure to Ensure Safe Operations at Design Basis Low River Level	A cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to ensure that the safety-related raw water pumps are available for safe plant operations down to the design basis low river level. Specifically, station analysis and abnormal operating procedures would not allow operation of the raw water pumps to the design basis low river water level. The licensee entered this issue into its corrective action program as Condition Report 2014-09159 which included actions to reevaluate the capability of the raw water pumps to operate at low river levels. This finding was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of human performance in that the licensee did not ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee deferred funding for a vendor analysis of the capabilities of the raw water pumps at the design low river level.
05000285/FIN-2014009-19	2014Q3	Fort Calhoun	Failure to Maintain B.5.b Equipment in a State of Readiness to Support Mitigation Strategies	A non-cited violation of 10 CFR 50.54(hh)(2), Conditions of License, was identified involving the failure to maintain available equipment needed to implement mitigating strategies to maintain or restore core, containment, and spent fuel pool cooling capabilities following large fires or explosions. Specifically, the licensee failed to maintain available a flexible suction hose related to the reactor coolant system heat removal mitigating strategy. The licensee initiated Condition Report 2014-08876 to address this deficiency and initiated action to procure and replace the missing flexible suction hose. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The NRC determined that this finding was of very low safety significance (Green) using NRC Manual Chapter IMC 0609, Appendix L, B.5.b Significance Determination Process, because it resulted in an unrecoverable unavailability of an individual mitigating strategy but did not result in multiple unavailable mitigating strategies such that reactor coolant system heat removal could not occur. This finding has a crosscutting aspect in the area of human performance in that the licensees inadequate B.5.b inventory procedure contributed to the lack of recognition that the degraded flexible suction hose was required to implement mitigating strategies.

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