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| Site | Quarter | Significance | Cornerstone | Violation of | Description | System | |
|---|---|---|---|---|---|---|---|
| 05000293/FIN-2018002-05 | Pilgrim | 2018Q2 | Green | No Cornerstone | 10 CFR 50.72(b)(3)(v), Loss of Safety Function | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR 50.72(b)(3)(v)(C) requires licensees to a notify the NRC within 8 hours any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above, Entergy did not make a required notification pursuant to 10 CFR 50.72(b)(3)(v)(C). Specifically, on June 20, 2017, secondary containment was declared inoperable due to simultaneous opening of both airlock doors, and Entergy did not make the required notification until June 22, 2017. Significance/Severity: This violation is being treated under the NRCs traditional enforcement process, for impeding the regulatory process, specifically Entergy did not make a required notification, as outlined in Inspection Manual Chapter 0612, Appendix B. The Reactor Oversight Processs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The severity of this violation was determined to be Severity Level IV, as outlined in Example 9 from Section 6.9.d. of the NRC Enforcement Policy. Corrective Action References: CR-PNP-2017-06380 and CR-PNP-2017-07015 The disposition of this finding closes Licensee Event Report 2017-011-00. | Secondary containment |
| 05000348/FIN-2018002-06 | Farley | 2018Q2 | Green | Mitigating Systems | 10 CFR 50 Appendix B Criterion XI, Test Control | Violation: 10 CFR 50, Appendix B, Criterion XI, Test Control, required in part, a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in all applicable design documents. Contrary to the above, the Unit 1 pressurizer power operated relief valve (PORV) PCV-445A was not set up properly for testing and the written test procedures did not incorporate the acceptance limits in all applicable design documents. Specifically, the open and closed limit switches were not set up properly which would result in shorter stroke times during testing per licensee procedure FNP-1-STP-45.11, Miscellaneous Cold Shutdown Valves Inservice Test. Additionally, licensee procedure FNP-1-STP-201.28, Pressurizer Power Operated Relief Valves Position Indication and Relay Logic Contact Verification Q1B31PCV0444B and Q1B31PCV0445A, Ver. 14, allowed a minimum stroke length of 0.5 inches while a vendor evaluation in Request for Engineering Review (RER) 941414 stated a minimum stroke length of 0.56 inches was required. | |
| 05000348/FIN-2018002-07 | Farley | 2018Q2 | Green | Mitigating Systems | Technical Specification | Violation: Technical Specifications (TS) Limiting Condition of Operability (LCO) 3.3.1, Reactor Trip System (RTS) Instrumentation, required the RTS instrumentation for each Function in Table 3.3.1-1 to be operable. The over temperature delta-T (T) function listed in Table 3.3.1-1 required 3 channels to be operable in Modes 1 and 2. With one channel inoperable, the required actions of Condition E of LCO 3.3.1 are required to be performed within the completion time. LCO 3.0.3 required in part, when an LCO is not met and the associated actions are not met, an associated action is not provided, or if directed by the associated actions, the unit shall be placed in a mode or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: Mode 3 within 7 hours; Mode 4 within 13 hours; and Mode 5 within 37 hours. Contrary to the above, since Unit 2 entered Mode 2 on Nov. 12, 2017, at 1138 with two channels of the OT delta T function inoperable until Nov. 13, 2017, at 0115 when one channel of the T function was restored, the licensee failed to place Unit 2 in Mode 3 within 7 hours and then Mode 4 within 13 hours as required by LCO 3.0.3. The time the two channels of the OTdelta T function was inoperable totaled 13 hours and 37 minutes. LCO 3.3.1 does not provide an associated action with two channels of the OT delta T function inoperable in Modes 1 and 2. The OT delta T trip function is provided to ensure that the design limit departure from nucleate boiling ratio (DNBR) is met. The inputs to the OT deltaT trip include pressure, coolant temperature, axial power distribution and reactor power as indicated by loop delta temperatures at full reactor coolant flow. Power range channel NI-42 provided the channel 2 input and pressurizer pressure instrument PT-457 provided the channel 3 input into the OT deltaT function. PT-457 was declared inoperable on Nov. 11, 2017, at 0522 and NI-42 was declared inoperable on Nov. 13, 2017, at 0136. Because NI-42 was found with a degraded center pin on high voltage cable connector, it was determined to be inoperable since Nov. 10, 2017. As a result, Unit 2 entered Mode 2 with two inoperable channels of OT delta T which is contrary to TS requirements. | |
| 05000348/FIN-2018002-08 | Farley | 2018Q2 | Green | Mitigating Systems | Technical Specification | Violation: Farley Nuclear Plant Unit 2 Technical Specifications (TS) limiting condition for operation (LCO) 3.7.5, Auxiliary Feedwater System, required all three auxiliary feedwater (AFW) trains shall be operable in modes 1, 2, and 3. For Condition A, one steam supply to turbine driven AFW pump inoperable, the required action A.1 was to restore the affected equipment to operable status within the required completion time of 7 days. If the required action and associated completion time is not met, action statement, Condition C required that the unit be in mode 3 within 6 hours and mode 4 within 12 hours. TS Surveillance Requirement (SR) 3.7.5.5 required verification that the turbine driven AFW pump steam admission valves open when air is supplied from their respective air accumulators. Contrary to the above, the licensee determined the steam admission valve (Q2N12HV3235B) was inoperable longer than the required action completion time of 7 days between May 6, 2016 and October 15, 2017, while Unit 2 was in modes 1, 2, and 3. Unit 2 was not placed in mode 3 or 4 as required by condition C of TS LCO 3.7.5. On October 31, 2017, a turbine-driven auxiliary feedwater (TDAFW) pump steam admission valve (Q2N12HV3235B) was tested with a flow scan analysis device during a refueling outage, while the plant was in Mode 6. This valve is the B-train steam admission valve that supplies steam to the TDAFW pump from the 2C steam generator. There is a redundant A-train steam admission valve that supplies steam from the 2B steam generator. During valve flow scan testing of the valve actuator it was discovered that air was leaking past the actuator piston o-ring seal inside the valve air actuator. Air leakage was measured greater than 10 psig per minute which was significant enough that the valve would not meet surveillance requirement (SR) 3.7.5.5 when instrument air was supplied solely from the valves associated air accumulator. Although the valve would stroke open with air supplied only from the accumulator, the SR 2-hour acceptance criteria to maintain the valve open could not be met. Each steam admission valve has an air accumulator associated with it. The air accumulator is designed to provide a sufficient quantity of air to ensure operation of the valve during a loss of power event or other failure of the normal instrument air supply for a period of two hours. Also, the inspectors determined that the licensee missed an opportunity to determine the cause of the o-ring failure since the o-ring was discarded during actuator rework. Procedure NMP-ES-001, Equipment Reliability Process Description, requires the preservation of physical evidence when failures occur. | Steam Generator Auxiliary Feedwater |
| 05000275/FIN-2018008-04 | Diablo Canyon | 2018Q2 | Green | Mitigating Systems | 10 CFR 50 Appendix B 10 CFR 50 Appendix B Criterion III, Design Control | This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program. This is being treated as a non-cited violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy. 7 Violation: Title 10 CFR Part 50, Appendix B, Criterion III, requires that measures shall include provisions to assure that appropriate quality standards are specified and included in design documents, and that deviations from such standards are controlled. Contrary to the above, from approximately February 2004 until August 2017, the licensee did not assure that appropriate quality standards were specified and included in design documents, and that deviations from such standards were controlled. Specifically, the licensee had classified the seat o-ring used in Crosby and Lonergan pressure relief valves (e.g., RV-354 and RV-355) servicing safety-related back-up air/nitrogen applications as non-safety related when they should have been classified as safety-related. Consequently, the o-rings were procured as commercial grade (non-safety related), not dedicated as safety-related and installed in safety-related equipment. Significance/Severity Level: This violation was more than minor because it had the potential to lead to a more significant safety concern if left uncorrected. Specifically, the use of non-qualified seat o-rings had the potential to cause excessive leakage past the seat, adversely affecting the fixed air/nitrogen volume required to operate safety-related equipment during a loss of normal air/nitrogen. Using IMC 0609, Appendix A, dated June 19, 2012, the team determined that this violation was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system or component, and operability was maintained. Corrective Action Reference(s): SAPNs 50935776 and 50970247 | |
| 05000259/FIN-2018010-01 | Browns Ferry | 2018Q2 | Green | Mitigating Systems | 10 CFR 50.48, Fire Protection License Condition | The Browns Ferry Nuclear Plant, Unit 3, Renewed Facility Operating License, DPR-68, License condition 2.C(7) required, in part, that TVA Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c)... Specifically, 10 CFR 50.48(c)incorporated by reference National Fire Protection Association Standard 805 (NFPA 805), and NFPA 805 section 2.4.2.2.2, Other Required Circuits, required in part, Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. Contrary to the above, since June 22, 2016, when the NFPA 805 requirements went into effect, the licensee did not implement and maintain in effect all provisions of the approved fire protection program, because the licensee did not correctly evaluate circuits that share common power supply for their impact on their ability to achieve nuclear safety performance criteria in accordance with NFPA 805.Significance: The team evaluated the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, for Mitigating Systems, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued May 2, 2018, and determined the finding to be of very low | |
| 05000277/FIN-2018010-03 | Peach Bottom | 2018Q2 | Green | Emergency Preparedness | License Condition - Fire Protection | This violation of very low safety significance was identified by Exelon and has been entered into Exelons corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Peach Bottom Unit 2 and Unit 3 Renewed Facility Operating License Condition 2.C.(4) requires, in part, Exelon to implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report. Fire Protection Program, Peach Bottom Atomic Power Station, Units 2 and 3, is incorporated by reference into the Updated Final Safety Analysis Report, as discussed in Section 10.12, Fire Protection Program. Fire Protection Program Chapter 5.1, Methodology, assumes that two or more circuit failures resulting in spurious operation of two or more valves in series at a high/low pressure interface may occur due to a postulated fire in any given area.Fire Protection Program Chapter 6.2, Analysis of High/Low Pressure Interfaces, requires Exelon to address the situations for which the isolation valves at a given interface point consists of two electrically controlled valves in series and where the potential may exist for a single fire to cause damage to cables associated with both valves. 8 Contrary to above, as ofMarch 14, 2018, Exelon identified they failed to evaluate two motor-operated valves in series, MO-2-06-2663 and MO-2-06-038B for Unit 2, and MO-3-06-3663 and MO-3-06-038B for Unit 3, where the potential may exist for a single fire to cause damage to cables associated with both valves. Specifically, a postulated fire scenario could cause spurious opening of the valves, which may potentially result in a fire-induced loss of coolant accident through the high/low pressure system interface. Exelons evaluation identified the affected valves cables are routed through Fire Area 6N for the Unit 2 valves, and Fire Area 13N for the Unit 3 valves, and thus, a possibility exists for a single fire to cause damage to cables associated with both valves. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Significance/Severity: The inspectors performed a Phase 2 Significance Determination Process screening for this issue, in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected the post-fire safe shutdown category because of spurious operations of safe shutdown components. Based on a walkdown of Fire Areas 6N and 13N, the inspectors did not identify any credible fire ignition source scenarios that could affect both Unit 2 motor-operated valves or both Unit 3 motor operated valves. Therefore, based upon task number 2.3.5, the inspectors determined that this finding screened to very low safety significance (Green).Corrective Action References: IR 04115309, EC 623585, EC 623586 | |
| 05000255/FIN-2018011-03 | Palisades | 2018Q2 | Green | Mitigating Systems | 10 CFR 50.48, Fire Protection License Condition - Fire Protection | License condition 2.C(3)requires the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program that complies with Title 10of the Code of Federal Regulations(CFR), Part50.48(a) and 10 CFR 50.48(c), NFPA Standard NFPA 805, as approved in the Safety Evaluation Report (SER)dated February 27, 2015. Section 2.4.3.3 of NFPA 805 states, in part, that the Probabilistic Safety Assessment (PSA)(Probabilistic RiskAssessment (PRA))approach, methods, and data shall be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.Contrary to the above, from February 27, 2015, until May 14, 2018, the licensee failed to base the PSA (PRA) approach, methods, and data on the as-built and as-operated and maintained plant.Specifically, the licensees PSA (PRA) model/analysis credited the suppression system located in the cable spreading room to suppress a type 2 fire scenarios, whereas the actual room contained numerous obstructions by the stacked cable trays located near the ceiling that interfered with the water spray pattern discharged from the sprinklers from providing adequate water density pattern to suppress a fire in areas below the cable trays which contained electrical panels.Significance/Severity Level: The performance deficiency was determined to be more-than-minor, and therefore, a finding because the performance deficiency, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to correctly model/analyze the as-built condition of the suppression system located in the cable spreading room in the PRA could potentially affect the risk associated with a fire in the room and could result in inappropriately screening out the effects of otherchanges associated with the fire area.The finding was of very-low safety significance (Green). While there may be a change to the plants baseline risk as a result of this issue, this is a fire modeling issue only; no physical plant fire protection feature was altered by the fire PRA model. Therefore, there was no increase in actual core damage risk to the physical plant. | |
| 05000390/FIN-2018050-01 | Watts Bar | 2018Q2 | Green | Initiating Events Mitigating Systems | 10 CFR 50 Appendix B Criterion III, Design Control | This violation of very low safety significance (Green)was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a Non-CitedViolation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 of the Code of Federal Regulations(10 CFR) Part 50 (10 CFR 50), Appendix B, Criterion III, Design Control, requires the licensee to effectively implement design control measures for piping analysis calculations* associated with the Unit 1 and Unit 2 emergency core cooling systems (ECCS).Contrary to the above, since initial operation of Unit 1 in 1996 and Unit 2 in 2016, Tennessee Valley Authority failed to ensure the proper hydraulic time history was utilized in TVAs TPIPE special purpose computer program used to determine static and dynamic linear elastic analyses for the ECCS including the effects of pipe voiding. This resulted in non-conservative voiding design acceptance criteria for the RHR and SI systems of both units. This performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to utilize proper hydraulic time history in the licensees TPIPE computer model resulted in developing non-conservative voiding acceptance criteria that was used during operation that could challenge ECCS functionality. The finding was determined to be of very low safety significance since additional analysis determined with reasonable assurance that the systems remained operable but non-conforming and would have performed their safety function.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 affected the design or qualification of mitigating systems; however, the mitigating systems maintained their operability. Corrective Action Reference:CR 1407257 | Emergency Core Cooling System |
| 05000272/FIN-2018403-02 | Salem | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000458/FIN-2018406-03 | River Bend | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000261/FIN-2018410-01 | Robinson | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000454/FIN-2018410-01 | Byron | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000219/FIN-2018410-04 | Oyster Creek | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000293/FIN-2018002-06 | Pilgrim | 2018Q2 | Severity level Minor | No Cornerstone | 10 CFR 50.9, Completeness and Accuracy of Information 10 CFR 50.73, Licensee Event Report System | This violation of minor significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a minor violation, consistent with the NRC Enforcement Policy. On June 22, 2015, Entergy submitted a licensee event report in accordance with 10 CFR 50.73 that contained information that was not complete or accurate in all material respects, contrary to the requirements in 10 CFR 50.9. Specifically, the licensee submitted Licensee Event Report 2015-004-00 to communicate the failure during testing of time delay Agastat relay 27A-B1X/TDDO intended to provide under-voltage protection for 480V emergency bus B6 by transferring power from bus B1 to bus B2. In the licensee event report, Entergy incorrectly documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. Upon identifying the issue, on March 8, 2016, Entergy submitted a revised licensee event report with the correct information. Enforcement: 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 22, 2015, Entergy provided information to the Commission that was not complete and accurate in all material respects. In the licensee event report, the licensee documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. However, bus B6 would actually have tripped from bus B1 and lost power completely. This information was material to the NRC because the NRC requires timely and accurate reporting of information related to events in order to evaluate the potential safety significance and required NRC response. Entergy identified the inaccuracy and entered the issue into its corrective action program (CR-PNP-2015-9762). On March 8, 2016, Entergy submitted a revision to the licensee event report (2015-004-01) that corrected the report. This failure to comply with 10 CFR 50.9 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes Licensee Event Report 05000293/2015-004-01. | |
| 05000263/FIN-2018002-01 | Monticello | 2018Q2 | Green | Emergency Preparedness Mitigating Systems | 10 CFR 50.72 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness 10 CFR 50.47, Emergency Plans 10 CFR 50 Appendix E 10 CFR 50.54 10 CFR 50.54(q) 10 CFR 50.47(b)(8) | 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 Section2.3.2 of the Enforcement Policy.Enforcement: Violation: Title 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 of 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(8) requires, in part, that a licensee must provide and maintain adequate emergency facilities and equipment to support the emergency response plan.Contrary to the above requirements, on March 23, 2018, the licensee identified the site failed to maintain the effectiveness of the emergency plan by not providing and/or maintaining equipment capable of measuring the Immediately Dangerous to Life and Health (IDLH) concentrations for several toxic chemicals as required to properly classify an Alert Emergency Action Level (EAL). Specifically, while performing an emergency equipment inventory, the licensee identified that detector tubes (Draeger tubes) available to measure chlorine gas concentrations were not capable of measuring the IDLH concentration of 10 ppm required to identify the threshold level for classifying an Alert EAL (HA 3.1) since the measurement range of the available sample tubes was 50500 ppm.The inability to properly classify the Alert EAL represented a Loss of Emergency Assessment Capability and resulted in the licensees submission of Event Notification Report # 53298 in accordance with the requirements of 10 CFR 50.72(b)(3)(xiii). An immediate extent of condition review performed by the licensee identified additional deficiencies in adequate sampling methods for determining IDLH concentrations for Butadiene, Ethylene Dichloride, and Gasoline. Additionally, the licensee identified that in April 2015 there was missed opportunity to correct this deficiency when an Emergency Preparedness (EP) Coordinator, performing a Control Room Emergency Equipment Inventory, identified the need to order and replace the existing chlorine detector tubes. The EP Coordinator added the incorrect detector tubes to the existing inventory form without validating the tubes detection range and accuracy to ensure it was capable of detecting the IDLH threshold concentration level of 10 ppm.Upon identification of the issue, the licensee implemented compensatory measures for determining the EAL classification and entered the issue into the corrective action program (CR 501000009876). On May 08, 2018, the licensee implemented the sites new EAL classification procedure that was developed using NEI 9901, Revision 6, which does not require atmospheric sampling (use of detection tubes) for classification of EAL HA 3.1.Significance/Severity Level: Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.81, the inspectors determined this finding was 10 of very low safety significance (Green) because a significant amount of equipment necessary to implement the E-plan was not available or functional to the extent that any key ERO member could not perform his/her assigned functions, in the absence of compensatory measures (Degraded Planning Standard), specifically the ability to accurately classify the Alert EAL. Determining the finding significance using IMC 0609, Appendix B, Table 5.41, results in the same finding significance (very low significance) since the performance deficiency would have rendered an EAL initiating condition ineffective such that the Alert would have been declared in a degraded manner.Corrective Action Reference: 501000009876, CR Toxic Gas Detector Tube. | |
| 05000324/FIN-2018011-03 | Brunswick | 2018Q2 | Mitigating Systems | The licensee used 0.78 eV as the limiting activation energy for Rosemount transmitters. The activation energy was based upon an academic paper documenting experimental work performed for the early space program and first published in 1965. The paper cautioned the reader that the methods used were experimental and were not validated. A 0.5 eV activation energy for electronics was documented by the Electric Power Research Institute (EPRI) report NP-1558, which attributed it to electron migration of aluminum. Reports published by the Institute of Electrical and Electronics Engineers (IEEE) indicated that activation energies for various electronics and their failure modes could range from 0.5-0.66 eV. The licensee did not document an independent failure modes and effects analysis to justify the activation energy that they used. In addition, the licensee chose to use less limiting activation energies that were not proven to be justified. Finally, the licensee was unable to demonstrate acceptable margins for extrapolation confidence. The IEEE standard 323-1974, section 6.5.2, Mathematical Modeling, stated, the first step in the qualification by analysis is generally the construction of a valid mathematical model of the electric equipment to be qualified. The mathematical model shall be based upon established principles, verifiable test data, or operating experience data. The mathematical model shall be such that the performance of the electric equipment is a function of time and the pertinent environmental parameters. All environmental parameters listed in the equipment specification must be accounted for in the construction of the mathematical model unless it can be shown that the effects of the parameter of interest are dependent on the effects of the remaining environmental parameters. Planned Closure Actions: The team must determine whether the activation energy used for the transmitters was appropriate and, if not, whether the licensee had the responsibility to verify the information provided by their vendors and contractors. The region is discussing this issue with NRC headquarters to find a resolution to this issue. | |||
| 05000331/FIN-2018411-01 | Duane Arnold | 2018Q1 | Green | Physical Protection | 10 CFR 73 | ||
| 05000387/FIN-2018010-01 | Susquehanna | 2018Q1 | Green | Barrier Integrity | 10 CFR 50.49 | 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 Section2.3.2 of the Enforcement Policy.Violation: 10 CFR 50.49(e)(5) requires, in part, that the electrical equipment qualification program must replace or refurbish the equipment at the end of its designated life.Contrary to the above, on November 16, 2017, the licensee identified that thirteen Unit 1, NAMCO limit switches in environmentally qualified (EQ) applications inside primary containment were not installed in their fully qualified configuration. Specifically, contrary to vendor instructions and EQAR-004 requirements, the limit switches for several containment isolation valves (CIV) have had their covers removed and reinstalled without replacing the gasket and cover screw O-rings. For this application, opening and/or removing the limit switch gasket /and cover screws O-ring constituted the end of the gasket/O-ring designated life. Significance/Severity Level: The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance (Green), because the limit switches provide only an open or closed signal indication in the main control room, so that operators are aware of the valve position, and can make appropriate assessment of plant conditions. The safety function of the containment isolation valves was not affected. | Primary containment |
| 05000255/FIN-2018010-01 | Palisades | 2018Q1 | Green | Initiating Events Mitigating Systems | 10 CFR 50.55a, Codes and Standards | Violation: Title 10 of theCode of Federal Regulations (CFR) Part 50.55a(g)(4), Inservice Inspection Standards Requirement for Operating Plants, requires that, throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the 2006 edition through 2008 addenda of the ASME Boiler and Pressure Vessel Code. This edition of the AMSE Code requires that a VT3 visual examination of supports other than piping supports be performed once every 10year inservice inspection (ISI) interval. Contrary to the above, since the beginning of plant operation, the safety-related CCW and SW pump lateral supports (classified as ASME Code Section XI Class 3) had never been included in the ISI program and therefore had never had the required VT3 examination performed during each 10year ISI interval. Corrective actions included incorporating the supports into the ISI program, scheduling the inspections as required, and validating that the supports were still capable of performing their safety function and that the CCW and SW systems remained operable.Significance/Severity Level: The inspectors determined that the failure to perform ASME Code Section XI required inspections of the CCW and SW pump lateral supports was a performance deficiency. The inspectors determined the performance deficiency was more than minor because it adversely affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to periodically inspect the pump lateral supports could result in the failure to identify a nonfunctional support that could increase the risk of a pump failure.The inspectors assessed the significance of the finding using Appendix A of the SDP. The finding was determined to be of very low safety significance (Green) because although it was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), the SSC remained operable. Corrective Action Reference: CRPLP201705784, OE Review Identified Palisades Failure to Inspect ASME Class 3 Pump Supports for SW and CCW Pumps, 1/26/2018 Safety Conscious Work Environment Observations Based on interviews with plant staff and reviews ofthe latest safety culture survey results to assess the safety conscious work environment on site, the team determined that, in general, plant personnel appeared willing to raise nuclear safety concerns through at least one of the several means available. Most of those interviewed had an adequate knowledge of the CAP process and would initiate a CR, or work with someone who would do so on their behalf, if they knew of a safety concern. A weakness was identified in plant personnel knowledge ofhow to use the electronic CR system. Specifically, there were some personnel who were not familiar with how to generate a CR or how to track the resolution of a CR. Personnel also expressed an overall frustration with feedback provided on a CR; either with difficulties in being able to see how something was resolved or with not being able to understand the decision-making process for the resolution of issues.Most individuals expressed a willingness to raise safety concerns without fear of retaliation and all employees knew the importance of having a strong safety conscious work environment. There were some instances where the free flow of information or a willingness to raise concerns through an individuals direct line of supervision were hampered due to the perception that supervision was not receptive to receiving the concern or addressing the issue. In some cases, this presented an uncomfortable work environment for the affected individuals. However, when presented with this situation, all individuals knew of other supervisors that they could bring their concerns to or other avenues to use to address anissue. All plant personnel were aware of the Employee Concerns Program (ECP), knew who the ECP coordinator was, and most were willing to use it as an avenue to raise concerns, if desired. However, some individuals believed that the ECP lacked the appropriate level of confidentiality to effectively address concerns. | |
| 05000272/FIN-2018007-02 | Salem | 2018Q1 | Green | Mitigating Systems | 10 CFR 50 Appendix R License Condition - Fire Protection Section III.G Section III.L | 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 an NCV, consistent with Section2.3.2 of the Enforcement Policy.Violation:Salem Unit 2Facility Operating License Condition 2.(C).(10), in part, requires PSEG to implement and maintain in effect all provisions of the fire protection program as described in the UFSAR, as approved by the NRC. SC.ER-PS.FP-0001-A3, Salem Fire Protection Report-Safe Shutdown Analysis, Revision 0, establishes the basis for demonstrating a capability to achieve and maintain post-fire safe shutdown as described in the UFSAR, Section 9.5.1.1, Fire Protection Program. SC.ER-PS.FP-0001-A3 states that 10 CFR Part 50, Appendix R, Section III.L describes the safe shutdown requirements when an alternate or dedicated shutdown capability is provided as required in Appendix R, Section III.G.3. Appendix R, Section III.L.3, states, in part, that alternative shutdown capability shall be independent of the specific fire area. SC.ER-PS.FP-0001-A3 designates Fire Areas 2FA-AB-84A and 2FA-AB-64A as alternative shutdown areas.Contrary to the above, as of January 10, 2018, PSEG identified that the power cable that supplies power to both the Nuclear Instrumentation System Wide Range Amplifier Panel 962 and Signal Processor Panel 964 is routed through Fire Areas 2FA-AB-84A and 2FA-AB-64A without a required fire barrier and, therefore, was not independent of the specific fire area. As a result, for a fire event in either of these alternative shutdown areas, the power supply to the Unit 2 Hot Shutdown Panel Source Range and Power Range monitor could be lost and result in the loss of the neutron monitoring function. PSEG promptly implemented compensatory measures for this deficiency that included establishing a fire watch for the affected area. Because this violation was of very low safety significance (Green) and was entered into PSEGs Corrective Action Program (NOTF 20785256), this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.Significance/Severity Level: The team performed a Phase 1 Significance Determination Process (SDP) screening, in accordance with Inspection Manual Chapter 0609, Appendix F, Fire Protection SDP, Task 1.4.5: Post-fire Safe-shutdown. This issue screened to very low safety significance (Green) because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. Specifically, the Hot Shutdown Panel Source Range and Power Range monitor only provides a process monitoring function for reactivity control and safe shutdown actions would be determined using reactor coolant system chemistry sampling for boron concentration. Corrective Action Reference: NOTF20785256 | Reactor Coolant System |
| 05000266/FIN-2018001-05 | Point Beach | 2018Q1 | Green | Mitigating Systems | Technical Specification | Violation: Technical Specification (TS) 3.0.4 states in part that entry into a MODE or other specified condition in the Applicability of a limiting condition for operation (LCO) shall only be made when the LCOs Surveillances have been met... TS 3.7.5 Auxiliary Feedwater (AFW) Limiting Condition SR 3.7.5.1 required in part Verify each AFW manual, power operated, and automatic valve in each water path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. Contrary to the above, at 1500 on October 29, 2017, Unit 1 entered MODE 3 and the licensee failed to verify that AFW (System required for MODE 3) turbine driven (TD) AFW steam supply valves 1MS235 and 1MS237 were in the correct (open) position. These valves were in fact shut rendering the TDAFW pump inoperable until the licensee identified this error and opened these valves at 1610 on October 29, 2017(reference; Licensee Event Report 05000266/201700200, Operation or Condition Prohibited by Technical Specifications). Significance/Severity: This licensee identified finding, affected the Mitigating Systems Cornerstone and was screened in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, issued June 19, 2012. Because of the short duration (~1 hour) that the TDAFW pump was not operable, the inspectors determined that this finding is of very low safety significance (Green) because: the performance deficiency was not a design or qualification issue; it did not represent a loss of the system function; the train was neither inoperable for greater than its allowed outage time nor was it inoperable for greater than 24 hours; and was not part of an external event mitigating system. Corrective Action Reference: AR 02233500 Made Mode Change With Inoperable TDAFW | Auxiliary Feedwater |
| 05000416/FIN-2018001-05 | Grand Gulf | 2018Q1 | Green | Mitigating Systems | 10 CFR 50.72(b)(3)(v), Loss of Safety Function | 10 CFR 50.72(b)(3)(v)(D) requires the licensee to report an event or condition that could have prevented fulfillment of a safety function (accident mitigation).Contrary to the above, from February 18, 2018, until February 23, 2018, Grand Gulf Nuclear Station failed to make a timely event report for an event or condition that could have prevented fulfillment of a safety function (accident mitigation). Specifically, Grand Gulf Nuclear Station experienced the concurrent inoperability of the division 2 diesel generator and the high pressure core spray diesel generator. Per Technical Specification Bases 3.8.1.E.1, there are insufficient standby ac sources available in this condition to power the minimum required engineered safety feature functions.Significance/Severity Level: In accordance with NRC Enforcement Policy, Section 6.9.d.9, the failure to make a report required by 10 CFR 50.72 is a Severity Level IV violation.Corrective Action Reference(s): The licensee entered the failure to make a timely report into the corrective action program as CR-GGN-2018-1595. | High Pressure Core Spray |
| 05000498/FIN-2018001-04 | South Texas | 2018Q1 | Severity level IV | Mitigating Systems | License Condition - Fire Protection | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 CFR 50.9, Completeness and accuracy of information, requires, in part, that information required by the Commissions regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. STP Nuclear Operating Company, Unit 2 Renewed Facility Operating License Condition 2.E. Fire Protection states, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 62. Updated Final Safety Analysis Report Subsection 9.5.1.6.1 Administrative Controls states, in part that the operability/functional capability of the fire protection systems required to protect safe shutdown capability is assured through the implementation of an administrative program. This program includes compensatory actions for systems out-of-service.Procedure 0PGP03-ZF-0001, Fire Protection Program, Revision 31, Step 7.3 requires, in part, that completed fire watch logs Form 4 or their equivalent shall be retained for 3 years.Contrary to the above, the licensee failed to maintained information required by the Commissions regulations, orders, or license conditions that was complete and accurate in all material respects as evidenced by the following two examples:1. On May 25, 2016, the written fire watch log, documented on Form 4, for Unit 2 Fire Watch 10118, for Room 105, for the hours of 1928 and 2015, indicated that the hourly fire watches were conducted by passing through the areas covered by the fire watch. However, the fire watch never entered Room 105 for these 2 hours. The hourly fire watch patrol data is material to the NRC in that it provides sufficient evidence of compliance with regulatory requirements.2. On May 25-26, 2016, the electronic fire watch scanned logs for Unit 2 Fire Watch 10118, for Room 105 between the hours of 2105 on May 25, 2016, to 0504 on May 26, 2016, show that the 9 hourly fire watches were conducted by passing through the areas covered by the fire watch. However, a temporary scan point was placed at the base of the ladder in Room 002 to scan for Room 105. The hourly fire watch individuals never entered Room 105. The hourly fire watch patrol data is material to the NRC in that it provides sufficient evidence of compliance with regulatory requirements.Significance/Severity Level: Although this violation is willful, it was brought to the NRCs attention by the licensee, it involved isolated acts of low-level individuals, and it was addressed by appropriate remedial action. Therefore, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy. Corrective Action Reference: Condition Report 18-0948 | |
| 05000289/FIN-2018001-03 | Three Mile Island | 2018Q1 | Green | Mitigating Systems | 10 CFR 50.63 | This violation of very low safety significancewas 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 Section2.3.2 of the Enforcement Policy.Violation: 10 CFR 50.63(c)(2) states, in part, that the alternate ac power source will constitute acceptable capability to withstand station blackout provided an analysis is performed which demonstrates that the plant has this capability from onset of the station blackout until the alternate ac source and required shutdown equipment are started and lined up to operate. The time required for startup and alignment of the alternate ac power source and this equipment shall be demonstrated by test. If the alternate ac source can be demonstrated by test to be available to power the shutdown buses within 10 minutes of the onset of station blackout, then no coping analysis is required. The Three Mile Island Unit 1 Station Blackout Evaluation Report 990-1879 identifies the station blackout (SBO) diesel generator as the alternate ac power source for the unit. Contrary to the above, from January 11, 2018, to January 12, 2018, the Three Mile Island Unit 1 alternate ac power source did not constitute acceptable capability to withstand station blackout. Specifically, during this timeframe, the SBO diesel generator was rendered unavailable due to fire service valve FS-V-225 being closed with no dedicated operator to reopen the valve. The time required for startup and alignment of the SBO diesel generator in this configuration had not been demonstrated by test to be available to power the shutdown buses within 10 minutes of the onset of station blackout.Significance/Severity Level: The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding required a detailed risk evaluation due it representing an actual loss of function of one non-Technical Specification train of equipment designated as high safety-significance for more than 24 hours. A Region I senior reactor analyst completed the detailed risk evaluation and estimated the increase in core damage frequency (CDF) associated with this performance deficiency to be 7E-8/yr or of very low safety significance (Green). The senior reactor analyst used the Systems Analysis Programs for Hands-On Evaluation (SAPHIRE) Revision 8.1.6, Standardized Plant Analysis Risk (SPAR) Model, Version 8.54, for evaluating the increase in risk. The analyst performed the assessment by failing the station blackout diesel generator for an exposure period of 30 hours due to its assumed unavailability. The dominant core damage sequence involved a steam line break in the turbine building (SLBTB) with a failure to isolate the steam line break, a loss of reactor coolant pump (RCP) seal cooling, failure of rapid secondary depressurization, failure of the RCP seal stage 2 integrity and failure of the High Pressure Injection mitigating function. In accordance with IMC 0609, Appendix H, Containment Integrity Significance Determination Process, Figure 5.1, the increase in core damage frequency per year was below 1E-7/yr and therefore the Large Early Release Frequency (LERF) contribution was determined not to have an effect on the very low safety significance determination.Corrective Action Reference(s): CR 04093302 | |
| 05000327/FIN-2018001-03 | Sequoyah | 2018Q1 | Green | Mitigating Systems | Technical Specification | 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 | Secondary containment Auxiliary Building Gas Treatment System |
| 05000255/FIN-2018001-03 | Palisades | 2018Q1 | Green | Mitigating Systems | Technical Specification | A violation of very low safety significance (Green) was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a Non-Cited Violation consistent with Section 2.3.2 of the Enforcement Policy. Enforcement:Violation: Technical Specification 3.7.6 requires that the combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T81) shall be greater or equal than 100,000 gallons. LCO 3.7.6, Condition A states that if the useable volume is not within this limit then A.1 Verify OPERABILITY of backup water supplies in 4 hours andA.2 Restore condensate volume to within limit in 7 days. Condition B states that if the Required Action and associated Completion Time is not met then B.1 Be in MODE 3 in 6 hours and B.2 Be in MODE 4 without reliance on steam generators for heat removal in 30 hours. Contrary to the above, on December 7, 2017 and March 3, 2016, the licensee failed to enter and comply with the actions required by LCO 3.7.6 Condition A and Condition B when Primary Makeup Tank Makeup Control Valve CV2008 could not be fully opened, resulting in a combined useable volume of the CST and T81 of less than 100,000 gallons.Significance/Severity Level: The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the CST and T81 volume were considered inoperable by the TS requirements, there was not a loss of safety function because credited backup water sources were available and operable.Therefore, the finding screened as Green.Corrective Action References: The licensee entered these issues into their CAP as CRPLP20175589, CRPLP20175554, CRPLP20175551, and CRPLP20161116 | Steam Generator |
| 05000498/FIN-2018001-03 | South Texas | 2018Q1 | Severity level IV | Mitigating Systems | License Condition - Fire Protection | This violation of very low safety significancewas identified by the licensee and has beenentered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 CFR50.48, Fire Protection, requires, in part, that licensees have a fire protection plan that outlines the plans for fire protection, fire detection, suppression capability, and limitation of damage.STP Nuclear Operating Company, Unit 2 Renewed Facility Operating License Condition 2.E. Fire Protection states, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 62. Updated Final Safety Analysis Report Subsection 9.5.1.6.1 Administrative Controls states, in part, that the operability/functional capability of the fire protection systems required to protect safe shutdown capability is assured through the implementation of an administrative program. This program includes compensatory actions for systems out-of-service.Procedure 0PGP03-ZF-0001, Fire Protection Program, Revision 31, Step 4.7.2.14, requires, in part, hourly fire watch personnel must pass through the areas covered by the fire watch and then sign and enter the time on the fire watch log, using the bar code reader, at least once every clock hour.Contrary to the above, on May 25 and 26, 2016, the licensees fire watch personnel failed to pass through the areas covered by the fire watch and then sign and enter the time on the fire watch log, using the bar code reader, at least once every clock hour. Specifically, two fire watch individuals documented conducting Unit 2 Fire Watch 10118 for Room 105 starting on May 25, 2016, at 1928 and finishing on May 26, 2016, at 0504 when in fact the individuals did not enter Room 105.Significance/Severity Level: Although this violation is willful, it was brought to the NRCs attention by the licensee, it involved isolated acts of low-level individuals, and it was addressed by appropriate remedial action. Therefore, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy. Corrective Action References: Condition Reports 16-7305, 16-7089, 16-7344, and 16-9077 | |
| 05000313/FIN-2018001-03 | Arkansas Nuclear | 2018Q1 | Green | Occupational Radiation Safety | 10 CFR 20 | Title10CFR20.1501(a) requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20, and that are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels, concentrations, or quantities of radioactive materials, and the potential radiological hazards that could be present.Contrary to the above, on August 7, 2017, the licensee failed to make necessary surveys of the Unit 2, 2T-15 tank room, that were reasonable to evaluate the magnitude and extent of radiation levels that could be present. Consequently, workers were allowed access to an area with dose rates up to 1000 millirem per hour at 30 cm without a proper briefing or oversight 17 Significance: Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because: (1) it was not associated with as low as is reasonably achievable (ALARA) 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.Corrective Action Reference(s): CR-ANO-2-2017-04634 and CR-ANO-2-2017-0533 | |
| 05000334/FIN-2018001-02 | Beaver Valley | 2018Q1 | Green | Public Radiation Safety Pr Safety | Technical Specification | Technical Specification 5.5.2 (c), Radioactive Effluent Controls Program, requires monitoring, sampling, and analysis of gaseous effluents. Contrary to the above, from 1989 to the present, the sample pump flow rates through several isokinetic nozzles was too high to allow for accurate monitoring and representative sampling. In 1989, automatic flow control features of some effluent monitoring instruments were disabled and in 2016, several new monitors were installed on the same isokinetic nozzle sample lines. Both of these actions prevents accurate monitoring and representative sampling.Significance/Severity: The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process. The inspectors determined that finding was of very low safety significance (Green).Corrective Action Reference(s): CR-2017-04211 and CR-2018-00283 | |
| 05000321/FIN-2018001-02 | Hatch | 2018Q1 | Green | Mitigating Systems | Technical Specification | 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: Hatch Nuclear Plant Technical Specification (TS) 5.7.2 states in part, areas with radiation levels greater than 1000 mRem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry.Contrary to the above, February 6, 2018, the licensee identified dose rates of 72 Rem/hr on contact, and 3.9 Rem/hr at 30 cm on the U-1 bottom head drain valve located in the 127 foot elevation of the Subpile room, in the Unit 1 Drywell. For approximately 4 hours, the entrance to the room was not locked or continuously guarded to prevent unauthorized entry as required by TS 5.7.2. Significance/Severity: The finding was of very low safety significance (Green) because it was not an as low as reasonably achievable (ALARA) planning issue, there was no overexposure nor potential for an overexposure, and the licensees ability to assess dose was not compromised.Corrective Action Reference(s):The licensee identified and documented the failure to control access to the Lock High Radiation Area (LHRA) in Condition Report 10458608. | |
| 05000263/FIN-2018001-02 | Monticello | 2018Q1 | Green | Mitigating Systems | 10 CFR 50.59, Changes, Tests and Experiments 10 CFR 50.90 Technical Specification | Violation: Title 10 CFR 50.59(d)(1) requires, in part, that the licensee maintain records of changes to the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c).These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to Paragraph (c)(2) of this section.Title 10 CFR 50.59(c)(2)(ii) requires that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the Final Safety Analysis Report (FSAR) (as updated).Technical Specification (TS) 3.3.1.1, Reactor Protection System (RPS) Instrumentation, states the RPS instrumentation for each function in Table 3.3.1.11 shall be operable. As specified in Table 3.3.1.11, Function 5, Main Steam Isolation Valve (MSIV) - Closure (8 channels) and Function 8, Turbine Stop Valve (TSV) Closure (4 channels) are required to be operable in Mode 1. TS 3.3.1.1, Condition C.1 states with one or more functions with RPS trip capability not maintained, to restore RPS trip capability in 1 hour and was applicable to both the MSIV and TSV RPS logic functional testing.Contrary to the above, on March 7, 2009 and July 11, 2009, the licensee failed to perform and maintain a written evaluation as required by 10 CFR 50.59(d)(1) to demonstrate a change to its facility did not require a license amendment. Specifically, the licensee incorrectly concluded in its 10 CFR 50.59 evaluation SCR080319, dated September 29, 2008, that no license amendment was required prior to implementing two surveillance test procedures; 0009 Turbine Stop Valve Closure Scram Test Procedure, Revision 16 on March 7, 2009 and; 0008 Main Steam Line Isolation Valve Closure Scram Test Procedure, Revision 20 on July 11, 2009. The test fixture was applied during quarterly surveillance testing through September 16, 2017.Implementation of procedures 0008 and 0009, respectively, resulted in the loss of RPS trip Function 5 (MSIV) and Function 8 (TSV) by bypassing more than the TS minimum allowed inputs per channel to maintain functionality, thereby violating the requirements of TS 3.3.1.1. Loss of these functions resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated) as specified by 10 CFR 50.59(c)(2)(ii).On November 14, 2017, the licensee generated CAP 501000005391 after conducting an operating experience evaluation of a similar event at another station concluding the event was applicable to the Monticello Plant. The surveillance procedures were immediately quarantined and subsequently revised on December 8, 2017 and December 11, 2017, to remove the use of the RPS test fixture.Significance/Severity Level:Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined this finding was of very low safety significance (Green) because it did not affect a single RPS trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown.The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV The disposition of this violation closes LER 05000263/201700600.Corrective Action Reference: 501000005391 | Reactor Protection System Main Steam Isolation Valve Main Steam Line |
| 05000443/FIN-2018001-01 | Seabrook | 2018Q1 | Green | Mitigating Systems | 10 CFR 50 Appendix B 10 CFR 50 Appendix B Criterion III, Design Control | 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 Section2.3.2 of the Enforcement Policy.Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control,requires, in part, that measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.Contrary to the above, from an unknown date until January 10, 2018, NextEra did not have a measure for verifying the adequacy of design of seven safety-related electrical manholes. Specifically, there were no original calculations to support the design of the manholes; and when NextEra staff reconstituted the structural design calculation, the results concluded that four of the seven manholes would not meet the design specification unless the loading demands were reduced from 500 to 200 pounds per square foot.Significance/Severity Level: Green because all four structures remained capable of performing their safety function.Corrective Action References:AR 02243800 and AR 02255652 | |
| 05000387/FIN-2018001-01 | Susquehanna | 2018Q1 | Green | Mitigating Systems | Technical Specification | 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: Susquehanna Unit 1 TS section 5.4.1 requires that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Susquehannas implementing instruction NDAP-QA-0503, General Housekeeping, Transient Material and Internal Cleanliness, Revision 45 implements aspects of the Regulatory Guide administrative procedures requirements. NDAP-QA-0503 section 6.1.5.h requires, in part, that transient equipment shall be located such that it will not impact safety related equipment during a seismic event. Locate all items at a distance greater than the height of the item from safety related equipment. Additionally, TS 3.5.1 Action Statement I directs immediate entry into Limiting Condition for Operation (LCO)3.0.3 if one core spray subsystem is inoperable with one low pressure coolant injection (LPCI) subsystem inoperable. LCO 3.0.3 requires action to be taken within 1 hour to place the unit in MODE 2 within 7 hours and MODE 3 within 13 hours.Contrary to the above, from December 1, 2017 to December 3, 2017, Susquehanna staged a 540 pound, ten foot long replacement pipe on 34 inch high stands within 34 inches of the safety related Unit 1, B Core Spray room cooler. Susquehanna concluded that the room cooler was inoperable because the pipe could have reasonably contacted and damaged the flexible conduit for the power cable to the room cooler during a seismic event. Additionally, from 7:48 a.m. on December 2, 2017 to 1:35 p.m. on December 3, 2017, maintenance was performed on the Unit 1, division 2 LPCI swing bus motor generator which rendered the division 2 LPCI system inoperable. During this time, Susquehanna did not perform the required actions of LCO 3.0.3 and remained in MODE 1.Significance/Severity Level: This violation is of very low safety significance (Green), since this finding did not represent a loss of system, a loss of function of at least a single train for greater than its TS allowed outage time, or a loss of a non-TS train. Corrective Action Reference(s): CR-2017-20227; CR-2018-01717; CR-2018-02250 | Core Spray Low Pressure Coolant Injection |
| 05000298/FIN-2018001-01 | Cooper | 2018Q1 | Green | Mitigating Systems | 10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR 50 Appendix B Criterion III, Design Control | Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that for those systems, structures, and components to which this appendix applies, Design control measures shall provide for verifying or checking the adequacy of design.Contrary to the above, between September 2003, and December 19, 2017, the licensee failed to verify or check the adequacy of design of quality-related components associated with the Division 1 and 2 emergency diesel generator 125 Vdc control power circuits. Specifically, in 2003, the licensee modified the design of the control power circuit through Part Evaluation (PE) 4222806 and replaced 24 original light bulb lamp assemblies with a different style of light bulb and a carbon film dropping resistor (vs. the original wire-wound design). This change created an unrecognized vulnerability that left the affected portions of the circuit with dropping resistors that provided insufficient protection from shorting due to indication light bulb failures. As a result, on December 19, 2017, the licensee declared both emergency diesel generators inoperable due to the design vulnerability.Significance/Severity Level: The finding created a design vulnerability in the emergency diesel generator control power circuits, and resulted in the Division 1 and 2 emergency diesel generators being declared inoperable at the time of discovery. Although the emergency diesel generators were declared inoperable, subsequent licensee analysis determined that the system retained its function, and maintained a reasonable expectation of operability while the design deficiencies existed. Accordingly, the inspectors assessed the significance of this finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, and determined this finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), but the SSC maintained its operability. Corrective Action Reference(s):Immediate corrective actions included compensatory measures to remove light bulbs from the vulnerable lamp assemblies in order to eliminate the shorting hazard. This issue was entered into the licensees corrective action program as Condition Report CR-CNS-2017-07513, and the licensee initiated a root cause evaluation | Emergency Diesel Generator |
| 05000333/FIN-2017004-03 | FitzPatrick | 2017Q4 | Green | Mitigating Systems | 10 CFR 50.65 10 CFR 50.65(a)(4) | The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV. 10 CFR 50.65(a)(4) states, in part, that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety. Contrary to the above, on March 28, 2011, and April 16, 2015, before performing maintenance activities on the electric bay unit coolers, as discussed in Section 4OA2.4, Exelon did not assess and manage the increase in risk that resulted from the maintenance activities. This issue was documented in Condition Report JAF-2016-0838. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609 Appendix K, Flowchart 2, Assessment of RMAs. The inspectors determined that the violation was of very low safety significance (Green) because the incremental core damage probability was less than 1E-5, with three risk management actions taken during the maintenance activities. | |
| 05000321/FIN-2017502-01 | Hatch | 2017Q4 | Green | Mitigating Systems | 10 CFR 50.54 10 CFR 50.54(q) | The following licensee-identified violation of NRC requirements was determined to be of very low safety significance or Severity Level IV and meet the NRC Enforcement Policy criteria for being dispositioned as a Non-Cited Violation. Because it had the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with Inspection Manual Chapter 0612, Appendix B. This finding was also determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involved the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan was reduced. Title 10 of the Code of Federal Regulations, Part 50.54(q) states, in part, that a licensee may make changes to emergency plans without prior NRC approval only if the changes do not reduce the effectiveness of the plans and the plans, as changed, continue to meet the standards of 50.47(b) and the requirements of Appendix E. Proposed changes that reduce the effectiveness of the approved emergency plans may not be implemented without application to and approval by the NRC. Contrary to the above, on multiple occasions between 2008 and 2014, the licensee implemented changes to their Radiological Emergency Plan and Emergency Action Levels (EALs) which reduced the effectiveness of the Plan. Specifically, the licensee deleted and/or changed EAL threshold values, all of which would have resulted in a change that reduced the effectiveness of the approved Emergency Plan and was implemented without application to and approval by the NRC. Because the violation was entered into the licensees corrective action program as Condition Report 10421212, it is being treated as a Green non-cited licensee-identified SL IV violation consistent with Section 2.3.2 of the Enforcement Policy. | |
| 05000263/FIN-2017409-02 | Monticello | 2017Q4 | Green | Physical Protection | 10 CFR 73 | ||
| 05000219/FIN-2017405-01 | Oyster Creek | 2017Q4 | Physical Protection | 10 CFR 73 | |||
| 05000382/FIN-2017403-02 | Waterford | 2017Q4 | Green | Physical Protection | 10 CFR 73 | ||
| 05000454/FIN-2017403-01 | Byron | 2017Q4 | Green | Physical Protection | 10 CFR 73 | ||
| 05000416/FIN-2017007-08 | Grand Gulf | 2017Q4 | Green | Mitigating Systems | 10 CFR 50 Appendix B Criterion III, Design Control | 10 CFR 50 Appendix B, Criterion III, requires in part, That measures shall be established to assure that the design bases are correctly translated into specifications, drawings procedures, and instructions. Contrary to the above, from original plant construction until June 22, 2016, GGNS failed to ensure the design basis tornado and differential pressures associated with it, would not cause a spurious trip of the Division I and II standby diesel generators. Specifically, a design basis tornado, includes a differential pressure of 3.0 psig, whereas an active diesel generator trip on high crankcase pressure actuates at 1.5 psig. The licensee identified this issue using an effective operating experience program and entered it in the corrective action program as Condition Report CR -GGN -2016- 04919. The violation is of very low safety significance (Green), for the same reason as NCV -05000416/2017007 -05, discussed in Section 1R21.4.5 of this report. | |
| 05000416/FIN-2017007-07 | Grand Gulf | 2017Q4 | Green | Mitigating Systems | Technical Specification - Procedures | The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as non- cited violations. Technical Specification 5.4.1(a) requires written procedures to be established, implemented, and maintained as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 4.e recommends, in part, instructions for startup of shutdown cooling and reactor vessel head spray system be prepared. Contrary to the above, from about 2004 until September 1, 2017, the 04-1-01-E12-2 instruction failed to provide instruction for placing the alternate decay heat removal system in service. Specifically, Step 4.9.2a.7(d) instructs an operator to, Manually control component cooling water temperature by throttling P44-F010A(B)(C), PSW inlet to CCW HXs. However, the purpose of that step is to throttle plant service water flow through the alternate decay heat removal system and component cooling water system to ensure both systems have plant service water flow, which is not accomplished by the instruction step. The licensee identified this procedural violation before the system was credited for availability during an inservice demonstration on September 1, 2017, and entered it in the corrective action program as Condition Report CR-GGN-2017-08643. The violation is of very low safety significance (Green) because, although the procedure did delay placing the system in service due to the procedure error, the system was capable of performing its design function, consistent with Inspection Manual Chapter 0609, Appendix G, Attachment 1, Exhibit 3 screening. | Service water Shutdown Cooling Decay Heat Removal |
| 05000395/FIN-2017007-04 | Summer | 2017Q4 | Green | Mitigating Systems | 10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants Technical Specification 10 CFR 50 Appendix B Criterion III, Design Control | Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, since 2010, the licensee failed to evaluate the loading of the emergency diesel generators at the maximum voltage and frequency allowed by TS 3/4.8.1 in Calculation DC08360-006, Diesel Generator 1A and 1B Loading, Rev. 12, and to evaluate battery terminal voltage at the maximum battery cell-to-cell resistance allowed by TS 3/4.8.2 in Calculation DC08320-010, Class 1E 125 Volt DC System Voltages and Voltage Drop, Rev. 18. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating SSC, and the SSC maintained its operability. The licensee entered these issues into their CAP as CRs 10-02395 and 10-02033. ATTACHMENT: SUPPLEMENTAL INFORMATION | Emergency Diesel Generator |
| 05000247/FIN-2017007-03 | Indian Point | 2017Q4 | Green | Mitigating Systems | Technical Specification - Procedures | The following violation of very low safety significance (Green) was identified by Entergy and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Indian Point Unit 2 Technical Specification 5.4.1.k and Indian Point Unit 3 Technical Specification 5.4.1.d require written procedures shall be established, implemented, and maintained covering FPP implementation. Procedure 0-PT-M004, Fire Extinguisher Inspection, Revision 9 implements monthly inspections of portable fire extinguishers to verify hydrostatic testing and periodic maintenance was performed within the periodicity specified by NFPA 10-1990, Standard for Portable Fire 17 Extinguishers. Procedure 0-PT-M004 requires portable fire extinguishers to be removed from service and replaced if their periodic maintenance or hydrostatic testing are not current within the specified periodicity. Contrary to the above, from July 1, 2015, to August 25, 2017, approximately 200 portable fire extinguishers were not removed from service and replaced when they exceeded their specified periodicity for maintenance and/or hydrostatic testing. Specifically, plant staff did not properly implement procedure 0-PT-M004 to verify portable fire extinguishers were periodically maintained and hydrostatically tested at intervals specified by NFPA 10-1990. The application software used in conjunction with 0-PT-M004 to perform monthly fire extinguisher inspections was revised in June 2015 and plant staff did not set up the software to require verification that fire extinguisher maintenance and hydro tests were current. Fire protection engineers identified the issue in August 2017, evaluated the subject fire extinguishers, and determined they were acceptable for continued use until December 31, 2017, based on the relatively short period of untimely maintenance/testing, satisfactory monthly verifications of physical condition, and the availability of additional portable fire extinguishers in the affected fire areas. Additional corrective actions included initiation of an accelerated maintenance and hydro test program to ensure all portable fire extinguishers met NFPA-10 maintenance and test requirements by December 21, 2017, and revision of 0-PT-M004 and the associated application software. Entergy entered the issue into the CAP (CRs IP3-2017-04084 and IP3- 2017-02945). The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the frequency of a fire and did not involve a loss or degradation of equipment or function specifically designed to mitigate an external event. | |
| 05000220/FIN-2017004-05 | Nine Mile Point | 2017Q4 | Green | Initiating Events Mitigating Systems | 10 CFR 50.65 10 CFR 50.65(a)(4) | The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV. Title 10 CFR 50.65(a)(4) requires, in part, ...the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Exelon procedure WC-AA-101-1006, On-Line Risk Management and Assessment, Revision 001, Section 4.1.3, states to consider work activities that cause equipment to be unavailable (e.g., trains of systems) for assessment of risk under the requirements of 10 CFR 50.65(a)(4). Contrary to the above, on October 17, 2017, Exelon identified a discrepancy in PARAGON (risk software) that resulted in an improper risk assessment for the days planned work. Review and correction of the error resulted in an elevated risk condition of Yellow during Nine Mile Point Unit 1, 11 feedwater pump (FW) maintenance. This performance deficiency was determined to be more than minor because it adversely affected the human performance attribute of the Mitigating Systems cornerstone and affected cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, on October 17, 2017, Exelon identified a planned activity that resulted in an unplanned Yellow risk activity during planned maintenance of the 11 FW pump. In addition, IMC 0612, Appendix E, Examples of Minor Issues, under Section 7, Maintenance Rule, Example E for inadequate risk assessment states in part that a more-than-minor issue would put the plant into a higher licensee-established risk category. The finding was evaluated using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding was determined to affect the overall plant risk with the 11 FW Pump being out of service for maintenance with PARAGON not elevating the overall plant risk from green to yellow. The risk deficit was elevated and determined to not be greater than 1E-6 event per year for Incremental Core Damage Probability Differential and not greater than 1E-7 events per year for Incremental Large Early Release Probability Differential. Therefore, the finding was determined to be of very low safety significance (Green). Exelon entered this issue into its CAP as IR 04064241. | Feedwater |
| 05000282/FIN-2017004-03 | Prairie Island | 2017Q4 | Green | Barrier Integrity | Technical Specification | Prairie Island TS LCO 3.0.3 requires, in part, that when an LCO is not met and an associated ACTION is not provided, action shall be initiated within 1 hour to place the unit in MODE 3 within 7 hours.Contrary to the above, at 1556 hours on May 4, 2016, the licensee failed to place Unit 2 in MODE 3 within 7 hours due to no associated ACTION provided within TS 3.6.5, Containment Spray and Cooling Systems for two containment cooling trains not OPERABLE. Specifically, between May 4 and May 5, 2016, operators failed to recognize that with the ongoing unplanned inoperability of the 122 control room chiller, and the subsequent unplanned inoperability of the A train #23 CFCU, the 122 control room chiller was a required support system for the B train #22 and #24 CFCUs. Therefore, with both of the Unit 2 CFCU trains inoperable, LCO 3.0.3 was required to be entered to place Unit 2 in Mode 3 within 7 hours. Because the supported system TS applicability was not recognized, LCO 3.0.3 was not entered as required and both trains of Unit 2 CFCUs were inoperable for approximately 35 hours.Because the inspectors answered No to questions B.1 and B.2 under Exhibit 3, Barrier Integrity Screening Questions of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the finding screened as very low safety significance (Green). The issue was entered into the licensees CAP as CAP 501000002726. Corrective actions included re-assessing shared system LCOs between Units 1 and 2, revising the LCO tracking database, implementing new standards for LCO 3.0.6 applications, and revisions to the Safety Function Determination Program. | Containment Spray |
| 05000282/FIN-2017004-01 | Prairie Island | 2017Q4 | Green | Occupational Radiation Safety | Technical Specification | Technical Specification 5.7.1 states, High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates. Technical Specification 5.7.1, further requires in part, that each entryway to such an areashall be barricaded and conspicuously posted as a high radiation area.Contrary to the above, on October 19, 2017, a licensee system engineer identified during the performance of a maintenance and engineering inspection that a chain that functioned as the barricade for the 22 reactor coolant pump vault, a posted high radiation area, was not installed. The licensee documented this issue in CAP 501000004026. The inspectors determined that this issue was of very-low safety significance (Green) after reviewing IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that this finding was not an ALARA Planning or Work Control issue; was not an overexposure; was not a substantial potential for overexposure; and the ability to assess dose was not compromised. | |
| 05000324/FIN-2017004-02 | Brunswick | 2017Q4 | Green | No Cornerstone | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy, for being dispositioned as a NCV. Unit 1 and Unit 2 facility operating license DPR-71 and DPR-62 condition 2.B.(6)requires, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program. Procedure AD-EG-ALL-1522, Duties of a Fire Watch, requires periodic fire watches to be performed within their designated time periods including any allowed grace periods. Contrary to the above, during the spring 2017 Unit 2 refueling outage, between March 1 and March 19, selected periodic fire watches were missed or not performed within the required grace periods. The finding was screened using IMC 0609, Appendix F Fire Protection Significance Determination Process, and was determined to be of very low safety significance (Green), because the reactor was able to reach and maintain safe shutdown. This issue was documented in the licensees CAP as NCR 2115035. | ||
| 05000282/FIN-2017004-02 | Prairie Island | 2017Q4 | Green | Emergency Preparedness | 10 CFR 50.47, Emergency Plans 10 CFR 50 Appendix E 10 CFR 50.54 Technical Specification 10 CFR 50.47(b)(4) 10 CFR 50.54(q) | Title 10 CFR 50.54(q)(2) requires, in part, that a holder of a nuclear power reactor operating license shall follow and maintain the effectiveness of an emergency plan that meets the requirements in Title 10 CFR Part 50, Appendix E and the planning standards of Title 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4) requires, in part, that the onsite emergency response plans for nuclear power reactors must meet the following standard: 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 November 22, 2000 and September 22, 2017, the licensee failed to maintain the effectiveness of an emergency plan that met the requirements of the planning standards of 10 CFR 50.47(b). Specifically, on September 22, 2017, the licensee identified that prior assessments of NRC Information Notice 9745, Supplement 1, Environmental Qualification Deficiency for Cables and Containment Penetration Pigtails, and a subsequent industry-initiated study to determine signal errors for Prairie Islands Unit 1 & 2 containment high range radiation monitors 1R48, 1R49, 2R48 & 2R49 (used in the licensees emergency classification and action level scheme) that impacted operability of the monitors, failed to restore capability to classify EALs during certain design basis accidents.The violation was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring capability of implementing adequate measures to protect the health and safety of the public in 33 the event of a radiological emergency. The inspectors referenced IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.41 and Figure 5.41. The finding was determined to be of very low safety significance (Green) because timely and accurate EAL classification capability for an event at the General Emergency level was unaffected due to redundant and diverse indications.In response, the licensee entered the issue into the CAP as CAP 501000001861, declared the containment high range radiation monitors inoperable per TS 3.3.3, Event Monitoring Instrumentation, implemented Emergency Plan interim measures to make the emergency response organization aware of the issue, performed an extent-of-condition review, and submitted a letter to the U.S. NRC within 14 days as required by TS. Final corrective actions included the addition of a note to the Prairie Island EAL matrix to acknowledge the potential for TIC errors for the containment high range radiation monitors during the first 5 minutes for post-loss of coolant accident (LOCA) or main steam line break events inside containment. | Main Steam Line |