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05000255/FIN-2018003-012018Q3PalisadesWire Not Landed on Safety Injection Initiation Relay CircuitThe inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018
05000282/FIN-2018002-012018Q2Prairie IslandResults of ISFSICask Array Dose Calculation Not Incorporated into FSARPrairie Island ISFSI FSAR, as updated, Revision 18, Section A7A.7 evaluates off-site dose rates for an array of ISFSI casks. In this dose rate calculation, explicit modeling credit is given to the earthen berm that surrounds the Prairie Island ISFSI as discussed in Section A7A.7.1. The earthen berm provides radiation shielding for the ISFSI. This calculation allows the licensee to demonstrate, in part, compliance with Title 10 of the Code of Federal Regulations (CFR) 72.104(a) which requires, in part, that, During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ. Calculation TN40HT0502, TN40HT Far Field Shielding Calculations, Revision 0, was performed by the licensee in support of a License Amendment Request (LAR) to modify the Prairie Island ISFSI TN40 cask design (designated as TN40HT casks). The TN40HT LAR was submitted to the NRC by the licensee on March 28, 2008. This dose rate calculation does not credit the earthen berm and, in part, also allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a). The licensee also provided this calculation directly to the NRC in a February 29, 2012, letter in response to a Request for Supplemental Information (RSI) from the NRC associated with the license renewal application for the Prairie Island ISFSI. Although the results from calculation TN40HT0502 for a single cask was incorporated into the Prairie Island ISFSI FSAR, Revision 18, in Tables A7A.22 and A7A.61, the results from TN40HT0502 for an array of casks which, in part, allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a), has not been incorporated into the ISFSI FSAR, Revision 18.Title 10 CFR 72.70, Safety analysis report updating requires, in part, that (a) Each specific licensee for an ISFSI shall update periodically, as provided in paragraphs (b) and (c) of this section, the FSAR to assure that the information included in the report contains the latest information developed (b) Each update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last update to the FSAR under this section. The update shall include the effects of: (2) All safety analyses and evaluations performed by the licensee in support of approved license amendments.This Unresolved Item is being opened to determine whether or not the licensee is required to update the ISFSI FSAR with the results of calculation TN40HT0502 for an array of casks in accordance with 10 CFR 72.70.Planned Closure Action: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Material Safety and Safeguards to determine whether or not calculation TN40HT0502 is subject to the FSAR updating requirements of 10 CFR 72.70 for the Prairie Island ISFSI.
05000282/FIN-2018001-022018Q1Prairie IslandQuestions Regarding the Corrective Action and Aging Management Programs Following the Discovery of 122 DDCLP FOST Vent Piping DegradationOn November 28, 2017, the inspectors identified a small hole in the vent piping for the below-ground 122 DDCLP FOST (located outside and adjacent to the plant screenhouse). The station generated AR 501000005894 and the shift manager declared the supported 22 DDCL pump operable-but-degraded with a temporary procedure change to AB4, Flood as a compensatory measure and wrapping of the pipe to preclude foreign material intrusion. The site backed up the immediate operability determination with a POD, evaluated past operability (no issues identified) and, subsequently replaced the affected portion of the pipe to restore full qualification. The inspectors concluded that these short term actions were acceptable to address the issue, but identified several concerns regarding prior actions to address the vent pipe corrosion. On March 1, 2018, the inspectors were provided the final evaluations for AR 501000005894. After review, the inspectors were concerned that the evaluations did not perform a sufficient review of: whether the corrective action program properly dispositioned corrosion of the pipe when first identified in July of 2015;whether the corrective action program and aging management program (AMP) performed as required to correctly classify and correct and/or manage the corrosion aging mechanism; and whether the extent of cause/condition for the adjacent 121 DDCLP FOST vent pipe was properly addressed.The inspectors passed these concerns to individuals in the engineering and regulatory affairs departments, but the licensee then stated that the evaluations provided on March 1, were, in actuality, still in a revision/review phase. The licensee stated that the final evaluations would likely address the inspectors concerns. On March 22, the inspectors were provided the final evaluations, but it appeared that only minor changes were made and the inspectors concerns were not addressed. On March 28, the inspectors again voiced their concerns with the licensee and two new ARs (501000010169 and 501000010178) were created documenting the following:the AR written identifying corrosion of the piping in July of 2015 was not evaluated under the AMP, the condition was determined to be operable and fully qualified, and it was closed to a work request to re-coat the piping but was never performed.the AR written in April of 2016 again noted the corrosion, but was closed to a non-conservative evaluation, the issue was not evaluated under the AMP, operability was again assessed as operable and fully qualified, and a work request was issued to apply a coating (not completed until May of 2017) 12 the AMP engineer was not consulted in 2015 or 2016 to determine if/how the issue fit into the AMP requirements for increased monitoring, development of acceptance criteria, and final corrective actions.Planned Closure Actions: To resolve this item, the inspectors will review planned actions regarding the degraded 121 DDCLP FOST vent pipe, further extent of condition reviews, and review planned licensee condition and causal evaluations regarding programmatic and/or human performance aspects of the issue.Licensee Actions: At the end of the inspection period, the licensee began excavation activities to replace the 121 DDCLP FOST vent pipe and had apparent cause and extent of condition evaluations in progress.Corrective Action Program References: ARs 501000005894, 501000010169 and 501000010178.
05000306/FIN-2018001-012018Q1Prairie IslandQuestions Regarding Corrective Action Program, Use of Operating Experience, and Qualification of the 21 125 VDC Battery due to Cell Lid CrackingThe inspectors identified an unresolved item regarding manual override of the auto-closure function of component cooling water system valves. Specifically, the inspectors noted that the system was not protected from tornado generated missiles when valves CV39153 & CV39154 are opened per procedure to support system alignments. The inspectors initially determined that further review was needed to determine if Technical Specifications are met if/when CV39153 & CV39154 are maintained open.Corrective Action Reference: AR 501000001642; 2017 50.59 Potential PD Evaluation 1133; 08/15/2017 Closure Basis: The inspectors reviewed the license basis documentation, procedures, and interviewed licensee personnel, and did not identify any licensee failure to meet a requirement or standard.
05000282/FIN-2017004-012017Q4Prairie IslandLicensee-Identified ViolationTechnical 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.
05000282/FIN-2017004-022017Q4Prairie IslandLicensee-Identified ViolationTitle 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.
05000282/FIN-2017004-032017Q4Prairie IslandLicensee-Identified ViolationPrairie 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.
05000282/FIN-2017004-042017Q4Prairie IslandLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to the above, on October 17, 2017, with Unit 2 in Mode 5, Cold Shutdown, the licensee failed to accomplish procedure 2C12.2, Purification and Chemical Addition Unit 2; Revision 34. Specifically, control room operators signed off steps as completed without validating that the procedure actions were performed in the field. These procedure steps that intended to close letdown valves and open purification valves, resulted in unintended transfer of primary coolant from the RCS to the chemical and volume control system hold-up tank instead of back to the RCS. In turn, this resulted in a reduction in RCS inventorywith reactor vessel level at approximately 1 foot below the flange (reduced inventory operations). Due to operators quickly recognizing a lack of letdown flow as discussed during a pre-job brief, the purification evolution was halted and actions were taken to restore reactor vessel level.Because the inspectors answered No to questions B.2 and B.3 under Exhibit 2, Initiating Events Screening Questions of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). Specifically, the loss of inventory event was self-limiting such that the leakage would have stopped before impacting the operating method of decay heat removal (shutdown cooling via RHR in this case). The issue was entered into the licensees CAP as CAP 501000003923. Corrective actions included an operations department human performance clock reset to share the lessons learned from the event.
05000282/FIN-2017003-042017Q3Prairie IslandLicensee-Identified ViolationPrairie Island Technical Specification 3.0.6 requires, in part, that an evaluation shall be performed in accordance with Technical Specification 5.5.13, Safety Function Determination Program, when a supported system LCO is not met solely due to a support system LCO not being met. Specifically, if a loss of safety function is determined to exist by the Safety Function Determination Program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.Contrary to this TS requirement, between August 18 and 22, 2017, control room operators did not evaluate Unit 2 A Component Cooling, Auxiliary Feedwater, and Cooling Water supported system LCOs while the 121 Safeguards Chilled Water support system LCO was not met. As a result, the appropriate Conditions and Required Actions were not entered during Unit 2 B Component Cooling and Auxiliary Feedwater supported system maintenance and testing activities for which a loss of safety function existed. Because the inspectors answered No to all questions under Exhibit 2.A of IMC 0609, Appendix A, The Significance Determination Process for Findings at-Power, the finding screened as very low safety significance (Green). Specifically, the finding did not represent (result in) an actual loss of function of two separate safety systems out-of-service for greater than their TS-allowed outage times. The above issues were documented in the licensees CAP as CAP 501000001929. Corrective actions included revisions to applicable station procedures for implementing TS 3.0.6 and the Safety Function Determination Program.
05000282/FIN-2017003-032017Q3Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program shall be established to assure that all testing required to demonstrate that SSCs will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, the licensee established procedure 5 AWI 3.12.4, 26 Post-Maintenance Testing, Revision 24, as the program for selecting and documenting post maintenance tests (PMTs) and return to service tests to ensure that SSCs would perform their intended function when returned to service. Contrary to the above, on September 20, 2017, the licensee failed to assure that testing required the demonstrate that three safety injection system actuation relays would perform satisfactorily in service was identified and performed in accordance with written test procedures, which incorporated the requirements and acceptance limits contained in applicable design documents. The three safety injection system actuation relays had not been tested following replacement during planned maintenance. Specifically, while reviewing PMT activities performed on the D5 EDG on September 19, 2017, the licensee identified three safety injection system actuation relays that had not been tested following replacement during planned maintenance. As a result, the D5 EDG was declared inoperable at the time of discovery on September 20, 2017. In response, the licensee performed an in-depth review of all recent D5 EDG maintenance activities to ensure that all PMT requirements were met and performed SP 2150, D5 Diesel Generator Function Test, on September 21, 2017, to adequately test all three safety injection system actuation relays and an additional D5 EDG slow start test to fully demonstrate operability of D5. Because the inspectors answered No to all questions under Exhibit 2.A of IMC 0609, Appendix A, The Significance Determination Process for Findings at-Power, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CAP 501000002920. Corrective actions included performing an apparent cause evaluation, department clock reset, and planned changes to 5 AWI 3.12.4 to ensure all required PMT activities are performed satisfactorily prior to returning SSCs to service.
05000282/FIN-2017003-022017Q3Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.48(b)(2) requires, in part, that all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O. Appendix R, Section III.G.3 of 10 CFR Part 50, requires, in part, that alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section. In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration. Contrary to the above, on December 21, 2015, the licensee failed to provide an alternative or dedicated shutdown capability for 17 MOVs credited in the licensees Appendix R Safe Shutdown Analysis that did not satisfy the requirements of 10 CFR Part 50, Appendix R, Section G.2. Specifically the MOVs could have been rendered unavailable for manual operator action following a postulated fire in the control or relay rooms. These manual actions were required to achieve and maintain safe shut down in the event of a fire that resulted in functional loss and/or evacuation of the control and/or relay rooms. Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non compliances identified as a result of a licensees transition to the new risk informed, performance based fire protection approach included in 10 CFR 50.48(c), and for 25 certain existing non compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk informed, performance based approach is referred to as National Fire Protection Association (NFPA) 805, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. At the time of discovery, the licensee was in transition to NFPA 805 and therefore the licensee-identified violation was evaluated in accordance with the criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation: (1) the licensee identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. The finding also met additional criteria established in section 12.01.b of IMC 0305, Operating Assessment Program. In addition, in order for the NRC to consider granting enforcement discretion the violation must not be associated with a finding of high safety significance (i.e., Red). The licensee performed risk evaluation V.SPA.16.001, Revision 0, dated March 27, 2017, and determined that this issue was not associated with a finding of high safety significance. A Region III Senior Reactor Analyst (SRA) reviewed the evaluation and concluded that the result was reasonable and that the finding was less than Red and eligible for enforcement discretion. The dominant core damage sequence from the licensees evaluation was a fire in the Control Room or Cable Spreading Room which could cause spurious operation of several MOVs necessary for safe shutdown. The SRA used IMC 0609, Appendix F, Fire Protection Significance Determination Process, to review the results of the licensees evaluation. The SRA validated the licensees calculations through a series of walkdowns, reviews of the calculation and verification of the values used were consistent with NUREG-6850 and IMC 0609, Appendix F. The licensees results were approximately 1E6 deltaCDF and 2E8 deltaLERF for this finding and hence were significantly lower than the 1E4 deltaCDF threshold for a finding of high safety significance. In addition, the licensee entered this issue into their corrective action program as CAP 1506561. As a result, the inspectors concluded that the violation met all four criteria established by Section 9.1(a) and that the NRC was exercising enforcement discretion to not cite this violation in accordance with the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues.
05000282/FIN-2017003-012017Q3Prairie IslandFailure to Ensure Correct Operation of Meteorological TowerA finding of very-low safety significance, and an associated NCV of Technical Specification (TS) 5.4.1 was identified by the NRC inspectors for the failure to implement and maintain procedures to ensure adequate operation of a meteorological tower. The licensee entered this issue into their Corrective Action Program (CAP) as CAP 501000001091, dated July 27, 2017. The licensee had initiated efforts to assess and remove unnecessary vegetation growth. The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding impacted the Plant Facilities/Equipment and Instrumentation Attribute of the Public Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, existing meteorological tower procedures did not include the assessment and subsequent removal of trees that could impair the correct operation of sensors located at the 10 meter elevation of the tower. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008. The violation was of very-low safety significance (Green) because: it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or Title 10 of the Code of Federal Regulations (CFR), Part 20.1301(e) criteria. The inspectors concluded that the most significant contributing cause of the performance deficiency involved the Resolution cross cutting component in the area of problem identification and resolution because this issue was previously entered into the licensees CAP in 2015 and closed with no action taken. (P.3)
05000282/FIN-2017002-032017Q2Prairie IslandFailure to Make an 8Hour Report Required by05000306/201700203 10 CFR 50.72(b)(3)(ii)(B)The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.72(b)(3)(ii)(B) due to the licensees failure on March 20, 2017, to report an unanalyzed condition within eight hours of discovery. Specifically, removing the lower latch assembly of a transom above Door 225, a steam exclusion barrier, during maintenance resulted in the inoperability of the Units 1 and 2 safeguards batteries and Auxiliary Feed Water (AFW) systems, and Unit 1 safeguards bus as determined by CAP 1549724.The inspectors determined that the failure to submit a report required by 10 CFR 50.72 for the unanalyzed condition described above was a performance deficiency. The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the information that 10 CFR 50.72 reporting serves. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement process. The inspectors determined that this issue was a SL IV violation based on Example 6.9.d.9 in the NRC Enforcement Policy. Example 6.9.d.9 specifically states, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. Because the issue has been evaluated under the Traditional Enforcement process, there was no cross-cutting aspect associated with this violation.
05000282/FIN-2017002-012017Q2Prairie IslandFailure to Properly Implement the Minor Maintenance Process During Door 225 Transom MaintenanceThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of TS 5.4.1.a, Procedures, associated with the licensees failure to properly implement Procedure FPWMMMP01, Minor Maintenance Process, Revision 5, while planning and performing maintenance on a steam exclusion barriertransom latch assembly. Specifically, on February 3, 2017, maintenance workers in coordination with the Fix-It-Now (FIN) Senior Reactor Operator (SRO) removed the lower latch assembly from a transom above Door 225 that rendered the steam exclusion barrier non-functional. Consequently, for an approximately five minute window during maintenance on the latch assembly, the 11 safeguards battery system was rendered inoperable with respect to a postulated turbine building High Energy Line Break (HELB) event. The licensee entered the issues into the Corrective Action Program (CAP) as CAPs 1548470 and 1549724.The inspectors determined that the licensees failure to properly implement procedure FPWMMMP01 as required by Technical Specification (TS) 5.4.1.a. was aperformance deficiency. The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems Cornerstone attribute of Human Performance 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 inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. Since the inspectors answered No to all questions within IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of Teamwork in the Human Performance cross-cutting area, and involved individuals and work groups not properly communicating and coordinating their activities within and across organizational boundaries to ensure nuclear safety was maintained. (H.4)
05000282/FIN-2017002-022017Q2Prairie IslandFailure to Implement the Emergency PlanA self-revealed finding, and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54 (q)(2), and 10 CFR 50.47 (b)(5) was identified on August 13, 2016, after a Notice of Unusual Event (NOUE) was declared due to reactor coolant system leakage greater than 25 gpm, the Shift Emergency Communicator (SEC) did not notify the States, Locals, and Tribal Community within 15 minutes of the classification.The inspectors reviewed IMC 0612, Appendix B, and determined that the finding was more than minor because it adversely affected the Emergency Response Performance attribute of the EP 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. Since the finding involved a failure to implement emergency preparedness requirements, the inspectors reviewed IMC 0609, Appendix B, Attachment 1, and determined that this was a finding of very-low significance (Green) because it involved the failure to notify the offsite response organizations as required in the Emergency Plan after the classification of an NOUE. The cause of this finding involved the cross-cutting area of human performance, with the aspect of procedure use and adherence because the SEC did not appropriately follow the notification procedure. (H.8)
05000282/FIN-2017002-042017Q2Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(8) states, Adequate emergency facilities and equipment to support the emergency response are provided and maintained. Section 8.2.2 of the Prairie Island Emergency Plan, Revision 52, states All supplies are inventoried quarterly and dated equipment and material are periodically replaced according to surveillance and testing program. Contrary to the above, from the fourth quarter of 2015 to fourth quarter of 2016, the licensee failed to maintain the effectiveness of the Emergency Plan by failing to complete the quarterly inventory of supplies and equipment in the alternative emergency response facility at their Red Wing Service Center. Specifically, for approximately five quarters, the licensee had not been conducting required quarterly inventory and equipment checks at the Alternative Emergency Response Facility due to several site procedures and supporting forms that verify continued facility readiness that were not updated or created following the 2014 Hostile Action Based Exercise.The inspectors determined that the finding was of very-low significance (Green) in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix B, Emergency Preparedness Significance Determination Process, Attachment 2, because this is a failure to comply with the Emergency Plan that does not result in a loss of a planning standard function. The licensee determined that the alternative emergency response facility remained functional during the time period when the inventories were missed. Because this finding is of very low safety significance, and has been entered into the licenseesCorrective Action Program under CAP 1513061, this violation is being treated as a Green NCV consistent with Section 2.3.2 of the NRCs Enforcement Policy.
05000282/FIN-2017001-012017Q1Prairie IslandFailure to Evaluate Changes to NRC Approved MethodologySeverity Level IV/Green. The inspectors identified a Green finding and associated Severity Level IV Violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.59(d)(1), for the licensees failure to perform a written evaluation which provided the bases for t he determination that a change in the NRC approved Westinghouse methodology referenced in the Updated Safety Analysis Report (USAR) for evaluating the acceptability of reactor pressure vessel internals baffle former bolting distributions did not require a license amendment. This finding was entered into the licensees Correction Action Program ( CAP ) as CAP documents 1539487, Documentation Missing in 50.59 Screening 4443, dated October 26, 2016; 1552331, BFB Screen Referenced Eval for SER Limitation 4 No n-Existent, dated March 6, 2017; and 1552314, BFB Screening Lacks Documentation for SER Limitation 3, dated March 6, 2017. The licensee performed an operability determination and determined the baffle bolts were operable. The inspectors reviewed the operability determination and no performance deficiencies were identified in this determination. The inspectors determined that the licensees failure to perform a written evaluation, providing the bases for the determination that a change in the NRC approved Westinghouse methodology for evaluating the acceptability of baffle former bolting distributions did not require a license amendment, was a performance deficiency. This finding was also evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The performance deficiency was determined to be more -than -minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, compliance with the NRC approved methodology of WCAP 15029 PA ensured the baffle former assembly maintained its structural integrity, avoiding a failure or excessive deflection of the baffle plates, and hence the primary concern of ensuring the emergency core cooling system could continue to perform its design function of cooling the reactor core. The inspectors determined the finding could be evaluated using the Significance 3 Determination Process (SDP) in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At -Pow er, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone. The finding screened as having very- low safety significance (Green) because the emergency core cooling system maintained its operability , specifically with respect to performing its safety function of ensuring adequate core cooling. As such, the finding corresponded to a Severity Level IV Violation in accordance with Example 6.1.d.2 of the NRC Enforcement Policy. The inspectors did not identify a cross cutting aspect because the performance deficiency was from 2013, and hence the issue did not represent current performance
05000282/FIN-2017001-022017Q1Prairie IslandLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix R, Section III.G.2 requires, in part, that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, up until April 21, 2016, the licensee failed to ensure that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions we re located within the same fire area outside of primary containment, one means of ensuring that one of the redundant trains is free of fire damage was provided. Specifically, the requirement was to provide separation of cables and equipment and associated non -safety circuits of redundant trains by a fire barrier having a 3 hour rating. However, fire barriers with unsealed combustible pathway penetrations existed between FA 85 (Holdup Tank Area/Demineralizer Area) and adjacent FAs 59 (Auxiliary Building Mezzanine Level Unit 1) and FA 74 (Auxiliary Building Mezzanine Level Unit 2) for Units 1 and 2 respectively. Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non compliances identified as a result of a licensees transition to the new risk informed, performance based fire protection approach included in 10 CFR 50.48(c), and for certain existing noncompliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk informed, performance based approach is referred to as NFPA 805, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. The licensee is in transition to NFPA 805, and therefore, the licensee- identified violation was evaluated in accordance with the criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation: (1) the licensee identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. The finding also met additional criteria established in section 12.01.b of IMC 0305, Operating Assessment Program. In addition, in order for the NRC to consider granting enforcement discretion the violation must not be associated with a finding of high safety significance (i.e., Red). The licensee provided the Fire PRA Multi -Compartment Analysis Notebook (FPRA PIMCA) for review, and concluded that this issue was not associated with a finding of high safety significance. An NRC Region III Senior Reactor Analyst (SRA) reviewed the evaluation and discussed it with licensee staff. The evaluation documents the results of fire modeling that concludes the fire 29 scenarios screen from further consideration because a damaging hot gas layer that could affect both compartments is not generated. The SRA concluded that the licensees result was reasonable and that the finding was less than Red and eligible for Enforcement Discretion. In addition, the licensee entered this issue into their CAP as 1519659. As a result, the inspectors concluded that the violation met all four criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues and that the NRC was exercising enforcement discretion to not cite this violation in accordance with the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues.
05000282/FIN-2016004-022016Q4Prairie IslandFailure to Properly Implement a Post-Maintenance Test Procedure during Safety Injection System Valve TestingGreen. A finding of very low safety significance was self-revealed, and an associated NCV of Technical Specification (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly implement surveillance procedure (SP) 1088B, Train B Safety Injection Quarterly Test, Revision 24, while performing a post-maintenance valve stroke test. Specifically, on November 14, 2016, while cycling a safety injection (SI) system pump suction valve, operators exposed the SI suction header to reactor coolant system (RCS) pressure, causing a relief valve to lift as designed, a subsequent unexpected RCS pressure drop below 240 pounds per square inch (psig), and requiring operators to trip both reactor coolant pumps (RCPs). The licensee entered the issue into the Corrective Action Program (CAP) as CAP 1541821. The inspectors determined that the licensees failure to properly implement procedure SP 1088B as required by TS 5.4.1.a was a performance deficiency (PD). The PD was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Initiating Events Cornerstone attribute of Configuration Control and affected the associated Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. Since the finding pertained to an event while the plant was shut down, the inspectors transitioned to IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Since the inspectors answered No to all questions within IMC 0609, Appendix G, Attachment 1, Exhibit 2, Initiating Events Screening Questions, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the PD was associated with the cross-cutting aspect of Teamwork in the Human Performance cross-cutting area, and involved individuals and work groups not communicating and coordinating their activities within and across organizational boundaries to ensure nuclear safety was maintained. (H.4)
05000282/FIN-2016004-032016Q4Prairie IslandFailure to Adequately Calibrate an ElectrometerGreen. A finding of very low safety significance, and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 20.1501(c) was identified by the inspectors for the failure to adequately calibrate the electrometer utilized in the validation of a JL Shepherd Calibrator. Specifically on November 30, 2015, the licensee performed a validation of a JL Shepherd Calibrator to ensure its correct operation. The electrometer used was incorrectly calibrated. The electronics and the detectors were required to be calibrated as a set, and this was not performed. The licensee entered this issue into their CAP as CAP 1543432. The inspectors determined that the licensees failure to properly calibrate the electrometer was a PD. The PD was more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Occupational Radiation Safety Cornerstone attribute of Program and Process and affected the Cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors applied IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to this finding. Since the finding was not associated with as-low-as-reasonably-achievable (ALARA) planning or work controls, nor was there an overexposure or a substantial potential for an over exposure and the ability to assess dose was not compromised, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the PD was associated with the cross-cutting aspect of Challenge the Unknown in the Human Performance cross-cutting area, and involved the licensee not challenging an unauthorized substitution for part of the electrometer that was damaged during shipment. (H.11)
05000282/FIN-2016004-042016Q4Prairie IslandLicensee-Identified ViolationPrairie Island Technical Specification 5.4.1, Procedures, required, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A contains, in part under Section 1, Administrative Procedures, Subsection e., Procedure Review and Approval. Contrary to the above, on December 12, 2013, and June 11, 2013, the licensee failed to properly implement FPGDOC04, Procedure Processing, Revision 19, to ensure that validation reviews were performed to ensure usability of C18.1, Engineered Safeguards Equipment Support Systems, following revisions to the procedure. Specifically, validation reviews were not performed during procedure revisions of C18.1 which lead to inadequate instructions to ensure that SCWSsupported system operability was properly addressed when SCWS functions were affected. This led to seven instances of conditions prohibited by TS for safeguards buses 15 and 16 between January of 2013 and May of 2015. The licensee later determined that although conditions prohibited by TS did occur based on the inadequate C18.1 instructions, an equally correct application of TS would have been to enter a 30day action statement for one SCWS inoperable per TS 3.7.11 and apply Surveillance Requirement 3.0.6 by performing a SFDP evaluation. This would not have resulted in conditions prohibited by TS for the supported AC or DC systems. Because the inspectors answered No to all questions under Exhibit 2 of Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CAP 1488482. Corrective actions included changes to C18.1 to implement the SFDP for future SCWS removals from service, and revisions to the FPGDOC04 job familiarization guide to ensure validation reviews are properly performed.
05000306/FIN-2016004-052016Q4Prairie IslandLicensee-Identified ViolationNorthern States Power CompanyMinnesota (NSPM), Prairie Island Nuclear Generating Plant Renewed Facility Operating License, Appendix B, Additional Conditions, Facility Operating License No. DPR42 and DPR60 (Amendment Nos. 206 and 193, respectively), required, in part, that The Alternate Source Term (AST) License Amendments 206/193 will be implemented after installation of the Unit 2 Replacement Steam Generators (RSGs) within 90 days after the completion of the outage in which the Unit 2 RSGs are installed. Further, implementation requirements incorporated within License Amendment 206/193 stated, in part, that prior to implementation of the AST license amendment, NSPM will revise the Prairie Island Nuclear Generating Plant design and licensing bases to indicate that the Steam Generator Water LevelNarrow Range Instruments are required to meet Regulatory Guide 1.97, Revision 2 requirements. Contrary to the above, on March 27, 2014, the licensee failed to revise the Prairie Island Nuclear Generating Plant design and licensing bases to indicate that the SGNR instruments were required to meet Regulatory Guide 1.97, Revision 2 requirements. Because the inspectors answered Yes to Question 1 under Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CAP 1424460. Corrective actions included replacement of the SGNR instrumentation with Regulatory Guide 1.97 compliant equipment.
05000282/FIN-2016004-012016Q4Prairie IslandBaffle Former Bolting Acceptance CriteriaFrom October 17November 28, 2016, the inspectors conducted a review of the implementation of the licensees inservice inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS), risk-significant piping and components and containment systems. This inspection constituted one ISI sample (see Sections 1R08.1, 1R08.3 and 1R08.5 below), as defined in IP 71111.0805. .1 Piping Systems Inservice Inspection a. Inspection Scope The inspectors either observed or reviewed records of the following Non-Destructive Examinations (NDEs) mandated by the American Society of Mechanical Engineers (ASME), Section XI Code, to evaluate compliance with the ASME Code Section XI and Section V requirements, and if any indications and defects were detected, to determine if these were dispositioned in accordance with the ASME Code or an NRC-approved alternative requirement. Ultrasonic examination of tubesheet to shell for steam generator (SG) 11; Magnetic particle examination of an integral attachment support rod for SG 11; Visual examination of reactor vessel nuts and washers (1 through 16); and Unit 1 metallic containment liner visual examination in 2012. During non-destructive surface and volumetric examinations performed since the previous refueling outage, the licensee had not identified any recordable indications. Therefore, no NRC review was completed for this inspection procedure attribute. The inspectors either observed or reviewed the following pressure boundary welds completed for risk-significant systems since the beginning of the last refueling outage to determine if the licensee applied the preservice NDEs, and acceptance criteria required by the Construction Code and ASME Code, Section XI. Additionally, the inspectors reviewed the welding procedure specification and supporting weld procedure qualification records to determine if the weld procedures were qualified in accordance with the requirements of Construction Code and ASME Code Section IX. Unit 1 reactor coolant pump (RCP) seal replacements. b. Findings No findings were identified. .2 Reactor Pressure Vessel Upper Head Penetration Inspection Activities a. Inspection Scope The licensee did not perform any welded repairs to vessel head penetrations since the beginning of the preceding outage for Unit 1. Therefore, no NRC review was completed for this inspection procedure attribute. For the Unit 1 vessel head, no examination was required pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50.55a(g)(6)(ii)(D) for the current refueling outage. Therefore, no NRC review was completed for this inspection attribute. b. Findings No findings were identified. .3 Boric Acid Corrosion Control a. Inspection Scope The inspectors performed an independent walkdown of the RCS and related lines in the containment, which had received a recent licensee boric acid walkdown, and verified whether the licensees boric acid corrosion control visual examinations emphasized locations where boric acid leaks can cause degradation of safety significant components. The inspectors reviewed the following licensee evaluations of RCS components with boric acid deposits to determine if degraded components were documented in the CAP. The inspectors also evaluated corrective actions for any degraded RCS components to determine if they met the ASME Section XI Code. 11 RCP seal bowl. The inspectors reviewed the following corrective actions related to evidence of boric acid leakage to determine if the corrective actions completed were consistent with the requirements of the ASME Code Section XI and 10 CFR Part 50, Appendix B, Criterion XVI. CAP 1465567; 12 RCP Seal Leakage. b. Findings No findings were identified. .4 Steam Generator Tube Inspection Activities a. Inspection Scope The licensee did not perform in-situ pressure testing of SG tubes. Therefore, no NRC review was completed for this inspection attribute. For the Unit 1 SGs, no examination was required pursuant to the TSs during the current refueling outage. Therefore, no NRC review was completed for this inspection procedure attribute. b. Findings No findings were identified. .5 Identification and Resolution of Problems a. Inspection Scope The inspectors performed a review of ISI/SG-related problems entered into the licensees CAP, and conducted interviews with licensee staff to determine if: the licensee had established an appropriate threshold for identifying ISI/SG-related problems; the licensee had performed a root cause evaluation (if applicable) and taken appropriate corrective actions; and the licensee had evaluated operating experience and industry generic issues related to ISI and pressure boundary integrity. The inspectors performed these reviews to evaluate compliance with 10 CFR Part 50, Appendix B, Criterion XVI requirements. Documents reviewed are listed in the Attachment to this report. b. Findings (1) Baffle Former Bolting Analysis Acceptance Criteria Introduction: The inspectors identified an Unresolved Item (URI) concerning the analysis that demonstrated the design adequacy of the baffle former bolting under design and licensing basis loading conditions. Description: The inspectors reviewed WCAP 17586P, Determination of Acceptable Baffle-Barrel Bolting for Prairie Island Units 1 and 2, Revision 0; WCAP15030NPA, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions under Faulted Load Conditions, dated March 2, 1999; and Safety Evaluation by the Office of Nuclear Reactor Regulation of WCAP15029, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, dated November 10, 1998. The inspectors were concerned that the licensee had evaluated the baffle former bolting using acceptance criteria different than what was reviewed and approved by the Office of Nuclear Reactor Regulation. In WCAP15030NPA, Section 4.3.2 stated that the stress allowable for primary membrane and bending of irradiated bolt material is taken to 0.9 times Sy (yield stress of baffle bolt material) for the faulted load condition. The stress allowable used in WCAP 17586P was based on ASME, Section III, Appendix F, specifically: (minimum of (0.9 times Su) ultimate stress of baffle bolt material), maximum of (0.67 times Su, Sy + 1/3 (Su - Sy)). The inspectors also reviewed 10 CFR 50.59 Screening No. 4443, Determination of Acceptable Baffle-Barrel Bolting, dated January 24, 2013, to determine whether the licensee performed a 50.59 evaluation for the use of ASME, Section III, Appendix F acceptance criteria. However, the inspectors identified that the change for the use of ASME, Section III, Appendix F acceptance criteria in lieu of the acceptance criteria contained in Section 4.3.2 of WCAP15030NPA was not explicitly reviewed in 50.59 Screening No. 4443. In response to the inspectors concern, the licensee initiated CAP 1539487, Documentation Missing in 50.59 Screening 4443, dated October 26, 2016. This issue is an URI pending evaluation of these concerns by the licensee, subsequent inspector review, and discussion with the licensee and Office of Nuclear Reactor Regulation (URI 05000282/201600401; 05000306/201600401; Baffle Former Bolting Analysis Acceptance Criteria).
05000282/FIN-2016003-012016Q3Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.48(b)(2) requires, in part, that all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O. Appendix R, Section III.G.1 of 10 CFR Part 50, requires, in part, that systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station(s) can be repaired within 72 hours. Contrary to the above, on January 7, 2016, the licensee failed to ensure that the Units 1 and 2 B RCS vent valves (necessary to achieve and maintain cold shutdown) could be repaired within 72 hours following a postulated fire. Specifically, the B RCS vent valves were credited within the licensees SSA following a postulated fire in the Units 1 and 2 auxiliary building mezzanine areas and could have been rendered unavailable for operation from the control room or emergency control station(s). Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non compliances identified as a result of a licensees transition to the new risk informed, performance based fire protection approach included in 10 CFR 50.48(c), and for certain existing non compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk informed, performance based approach is referred to as NFPA 805, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. The licensee is in transition to NFPA 805 and therefore the licensee-identified violation was evaluated in accordance with the criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation: (1) the licensee would have identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures (see Section 4OA3.3) within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. The finding also met additional criteria established in section 12.01.b of IMC 0305, Operating Assessment Program. In addition, in order for the NRC to consider granting enforcement discretion the violation must not be associated with a finding of high safety significance (i.e., Red). The issue was of very low safety significance (Green) because it did not impact the licensees ability to reach hot shutdown. The licensee entered this issue into their corrective action program as CAP 01507901. As a result, the inspectors concluded that the violation met all four criteria established by Section 9.1 of the NRCs Enforcement Policy and the NRC was exercising enforcement discretion to not cite this violation in accordance with the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues.
05000282/FIN-2016002-012016Q2Prairie IslandLicensee-Identified ViolationPrairie Island TS 3.6.3, Containment Isolation Valves, Required Action A.1 required, in part, isolation of the affected penetration flow path within 4 hours if one or more penetration flow paths with one containment isolation valve inoperable. Contrary to the above, since August 4, 2012 on 21 occasions for Unit 1 and 23 occasions for Unit 2 (three year reporting window), the licensee failed to isolate containment spray header penetration flow paths within 4 hours during the performance of quarterly containment spray pump surveillance procedures SP 1090A & 1090B and SP 2090A & 2090B. Specifically, the SPs inappropriately credited Note 1 of TS 3.6.3 and created open flow paths from the Unit 1 and 2 containments under administrative control while vent and/or drain valves connected to the containment spray header were opened. The opening of these valves was to facilitate draining of the header and to verify no leakage past manual isolation valves during containment spray pump operation in recirculation mode. On August 4, 2015, the licensee generated CAP 01488454 which questioned whether use of TS 3.6.3 Note 1 to open the containment spray header vent and drain valves under administrative control was permissible. The licensee performed an apparent cause evaluation and determined that because the vent and drain valves were not considered part of a containment penetration flow path, Note 1 could not be applied. A past operability review was performed and it was determined that on multiple occasions (at 1-10 hour durations) over the prior three years, the vent/drain opening resulted in a 3/8 opening in the containment pressure boundary. Because the resultant leakage at peak containment pressure during a design basis accident (approximately 4 percent of the containment volume per day) would have exceeded the maximum allowable leakage rate, conditions that could have prevented the fulfillment of the safety function of the Units 1 and 2 containments and, conditions that were prohibited by TS, had occurred. Because the inspectors answered Yes to question B.1 under Exhibit 3, Barrier Integrity Screening Questions of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors transitioned to IMC 0609, Appendix H, Containment Integrity Significance Determination Process. Because the leak rate through the vent/drain openings would not have exceeded greater than 100 percent of the containment volume per day at calculated peak containment internal pressure, the finding screened as very low safety significance (Green). The issues were entered into the licensees CAP as CAP 01488454. Corrective actions included immediate quarantine of the affected SPs and subsequent revisions to the SPs and TS Bases.
05000282/FIN-2016007-022016Q2Prairie IslandInadequate Operability DeterminationsA finding of very low safety significance with two examples and an associated non-cited violation of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FPOPOL01, Operability/Functionality Determination, Revisions 14 and 15. Specifically, on two occasions, the licensee failed to properly evaluate potential operability concerns associated with the Unit 2 emergency diesel generator (EDG) day tanks and the Unit 2 train A cooling water (CL) system piping. The licensee entered the issues into the Corrective Action Program as Action Requests 1525842 and 1526070. The inspectors determined that the licensees failure to accomplish the requirements of procedure FPOPOL01, Operability/Functionality Determination, Revisions 14 and 15, to properly evaluate the operability issues associated with the Unit 2 EDG day tank fuel oil level and the Unit 2 CL system piping (both safety-related, mitigating systems) was a performance deficiency. The performance deficiency, with two examples, was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," it was associated with the Mitigating Systems Cornerstone attributes of Equipment Performance (for the Unit 2 EDGs) and Protection against External Factors (for the Unit 2 CL piping) and adversely affected the Cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events. The inspectors utilized IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding screened as very low safety significance (Green) since the inspectors answered Yes to Question 1 of Section A of Exhibit 2, Mitigating Systems Screening Questions. The inspectors concluded that this issue was cross-cutting in the area of Problem Identification and Resolution in the aspect of Evaluation. As defined in IMC 0310, Aspects Within the Cross-Cutting Areas, this aspect states, The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee had not thoroughly evaluated the operability issues associated with the Unit 2 EDG day tank levels and the Unit 2 CL piping structural integrity.
05000282/FIN-2016007-012016Q2Prairie IslandFailure to Ensure Breaker Main Contacts are Fully AlignedA finding of very low safety significance and associated non-cited violation of Technical Specification Section 5.4.1, Procedures, was identified by the inspectors for the licensees failure to ensure the 21 safeguards diesel exhaust fan main contact connectors were fully engaged and aligned as required per electrical maintenance procedures to ensure proper operation of the breaker. As part of their corrective actions, the licensee aligned and re-engaged the main contact connectors as necessary. In addition, the licensee ensured maintenance personnel were aware of the operating experience to prevent the same issue from occurring in the future. The violation was entered into the licensees corrective action program as Action Request 1525844. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone and the breaker failure led to the inoperability of the 21 safeguards diesel exhaust fan and impacted the availability of the 22 cooling water system diesel driven pump. This finding represented a loss of the 22 safeguards diesel cooling water pump function for longer than the Technical Specification allowed outage time of 7 days and therefore required a detailed risk evaluation. The regional senior reactor analyst performed a detailed risk evaluation of this finding using the Prairie Island Standardized Plant Analysis Risk Model revision 8.19 and determined the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because it was not indicative of current performance.
05000263/FIN-2016001-012016Q1MonticelloFailure to Use Procedures While Performing Activities Affecting QualityAn NRC identified finding of very low safety significance (Green) and associated of 10 CFR 50, Appendix B, Criterion V; Instructions, Procedures, and Drawings, was identified on February 5, 2016, as a result of the licensees failure to use procedures while performing activities affecting quality. Specifically, the licensee failed to accomplish activities affecting quality in accordance with FP-G-DOC-03; Procedure and Work Instruction Use and Adherence, in that documented procedures were not used to install a conduit support on safety related Emergency Filtration Train (EFT) Division II conduits. Immediate corrective actions included removal of the support and entering the issue into the licensees Corrective Action Program (CAP) 1511349. The finding was determined to be more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the inspectors based this determination on the fact that performing activities affecting quality without using procedures has the potential to adversely affect the design/qualification of a Structure, System, and Component (SSC) or impact the operability or functionality of a system or component. The inspectors determined the finding to have very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, teamwork because of the licensees work group failures to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained.
05000306/FIN-2015008-012015Q4Prairie IslandFailure to Correct an NCV Associated with Inadequate Gas Monitoring of Inaccessible RHR Gas Susceptible LocationsThe inspectors identified a finding of very low safety significance (Green), and an associated cited violation of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to correct a condition adverse to quality (CAQ). Specifically, on August 1, 2011, the NRC issued an NCV for the failure to monitor five safety-related gas susceptible locations considered to be inaccessible, which is a CAQ. As of November 24, 2015, the licensee had not corrected this CAQ for two of those locations and did not have plans to restore compliance. The licensee captured this issue into their Corrective Action Program (CAP) with a proposed corrective action to develop an alternative monitoring method for these locations when the unit is operating. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee was able to access and inspect these locations during the refueling outage that was ongoing when this issue was identified and confirmed that they were full of water during the previous operating cycle. In addition, a historical review did not find information that challenged operability due to gas accumulation at these locations. The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate their discovery that the CAQ was not been corrected on July 29, 2013. Specifically, on 2013, the licensee initiated a condition evaluation (CE) to determine if the action plan at the time addressed the NCV associated with the CAQ. However, the CE was closed by crediting actions that were similar to those that resulted in the NCV and other documented observations associated with the inappropriate resolution of the issue.
05000282/FIN-2015004-032015Q4Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.48(b)(2) requires, in part, that all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O. Appendix R, Section III.G.3 of 10 CFR Part 50, requires, in part, that alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section. In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration. Contrary to the above, on April 19, 2015, the licensee failed to ensure that alternative or dedicated shutdown capability and its associated circuits were independent of cables in the area. Specifically, procedure F5 Appendix B, Control Room Evacuation (Fire), Revision 31, did not contain actions to isolate the RCP breaker circuits to prevent restarting due to a fire induced loss of remote trip and loss of RCP seal cooling water that could lead to an increased rate of seal degradation and a small break loss of coolant accident. These actions were required to achieve and maintain safe shutdown in the event of a fire that resulted in functional loss and/or evacuation of the control/relay and cable spreading rooms. Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non compliances identified as a result of a licensees transition to the new risk informed, performance based fire protection approach included in 10 CFR 50.48(c), and for certain existing non compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk informed, performance based approach is referred to as NFPA 805, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. The licensee is in transition to NFPA 805 and therefore the licensee-identified violation was evaluated in accordance with the criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation: (1) the licensee would have identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. The finding also met additional criteria established in section 12.01.b of IMC 0305, Operating Assessment Program. In addition, in order for the NRC to consider granting enforcement discretion the violation must not be associated with a finding of high safety significance (i.e., Red). The licensee performed risk evaluation V.SPA.15.012, Revision 3, dated December 18, 2015, and determined that this issue was not associated with a finding of high safety significance. A region III senior reactor analyst (SRA) reviewed the evaluation and concluded that the result was reasonable and that the finding was less than Red and eligible for enforcement discretion. The dominant core damage sequence from the licensees evaluation involved an electrical cabinet fire in the relay room involving the cables that could cause spurious operation of the RCPs and that would lead to alternate shutdown. The licensee identified several conservative assumptions in the analysis. The SRA agreed that some were conservative, notably that any fire affecting the cables in the relay room that could cause a spurious start of an RCP would also result in a loss of all seal cooling due to fire damage. The SRA used IMC 0609, Appendix F, Fire Protection Significance Determination Process, to review the results of the licensees evaluation. The relay room is similar to a cable spreading room with electrical cabinets. The fire frequency for this room in Appendix F is 6E3/yr. The probability of non-suppression was estimated to be 2E2 and the spurious operation probability was assumed to be 0.6. The product of these values (7.2E5/yr) represents a bounding relay room fire scenario delta core damage frequency (CDF) for this finding. Since the bounding result is consistent with the licensees conclusion, the SRA determined that the delta core damage frequency for the finding was less than 1E4/yr, which is less than Red. In addition, the licensee entered this issue into their corrective action program as CAP 01475242. As a result, the inspectors concluded that the violation met all four criteria established by Section 9.1(a) and that the NRC was exercising enforcement discretion to not cite this violation in accordance with the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues.
05000282/FIN-2015004-022015Q4Prairie IslandFailure to Adequately Calibrate Liquid Effluent MonitorsThe inspectors identified a finding of very low safety significance (Green) and associated NCV of TS 5.5.1.a for the failure to comply with the Offsite Dose Calculation Manual (ODCM) for not using calibration sources that were traceable to the National Institute of Standards and Technology (NIST) or equivalent during the calibration of station effluent monitors. The licensee entered the issues into the CAP as CAPs 01490581 and 01500149. Immediate corrective actions included the re-calibration of impacted monitors and the performance of an extent of condition evaluation for other radiation monitor calibrations. The PD was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the plant facilities/equipment and instrumentation attribute of Public Radiation Safety and it adversely impacted the cornerstone objective of ensuring adequate protection of public health and safety due to failure to properly calibrate certain effluent monitors. Subsequent calibrations of the monitors determined that the monitor efficiency was previously overstated. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, but did not identify any similar examples. The finding was assessed using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated, February 12, 2008, and determined to be of very low safety significance (Green), because it was associated with the effluent release program but was not a failure to implement an effluent program, public dose did not exceed Appendix I criteria, and the limits in Title 10 CFR 20.1301(e) were not exceeded. A cross-cutting aspect was not assigned as this issue occurred numerous years ago. The station has since performed monitor calibrations with radioactive sources with known quality.
05000306/FIN-2015004-012015Q4Prairie IslandFailure to Meet ANSI N14.6 Section 5.3.1 RequirementsThe inspectors identified a finding of very low safety significance (Green), and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to incorporate the American National Standards Institute (ANSI) N14.61978, Section 5.3.1 required testing frequency for the reactor vessel head and reactor vessel internals lifting devices into the controlling preventive maintenance procedure. Compliance with the ANSI standard was documented in the Safety Evaluation Report (SER) for the licensees control of heavy loads. The licensee documented the issue in the corrective action program (CAP) as CAP 01497779 and performed testing on the reactor vessel head and internals lifting devices during the outage. The inspectors determined the licensees failure to comply with ANSI N14.61978, Section 5.3.1, for the continued use testing of special lifting devices was a performance deficiency (PD). The PD was determined to be more-than-minor and a finding because the PD was associated with the Initiating Events Cornerstone attribute of design control, and adversely affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, compliance with ANSI N14.61978, Section 5.3.1 ensured safe load handling of heavy loads over the reactor core, and/or over safety-related systems through established testing for the continued functionality of the special lifting devices. The failure to perform the required frequency of testing on special lifting devices could increase the likelihood of a load drop and could decrease the load handling reliability of the lifting device if the device were returned to service with potentially unacceptable flaws. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase I - Initial Screening and Characterization of Findings, Table 3. Since the finding was associated with shutdown conditions, the inspectors used Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined that none of the conditions constituting a loss of control were met, as described in Appendix G, Attachment 1, Phase I Operational Checklists for Both PWRs (Pressurized Water Reactors) and BWRs (Boiling Water Reactors), for this finding, and neither a Phase II nor a Phase III analysis was required. Therefore, the inspectors determined that this finding was of very low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Resources, for the licensees failure to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, the licensee staff evaluated NRC Information Notice (IN) 201412, Crane and Heavy Lift Issues Identified during NRC Inspections, in corrective action program (CAP) document 01457469. However, in CAP 01457469, the licensee concluded that issues identified in IN 201412 related to other licensees not performing testing in accordance with ANSI N14.6 requirements were not applicable to the licensee at the Prairie Island Nuclear Generating Plant. Therefore, the inspectors determined that there was a recent missed opportunity for the licensee to have reasonably identified that the current preventive maintenance procedure for special lifting devices was not in accordance with the ANSI N14.61978 requirements, as referenced in the SER.
05000263/FIN-2015003-032015Q3MonticelloFailure to Identify Safe Shutdown Equipment Impacts in Fire Strategy ProceduresThe inspectors identified a finding of very low safety significance and an NCV of TS 5.4.1.d when the licensee failed to maintain procedures associated with Fire Protection Program Implementation, consistent with the Updated Safety Analysis Report (USAR), to ensure that fire strategy procedures accurately indicated safe shutdown (SSD) equipment. Specifically, on June 25, 2015, the licensee failed to maintain A.3-12-C, Condenser Room Fire Strategy, to ensure SSD equipment was appropriately identified. In this case, fire strategy A.3-12-C failed to identify any SSD equipment in the room, despite the fact that SSD cabling ran through the room and was included in the USAR Fire Hazards Analysis. Corrective actions included performance of an extent of condition review which identified 40 other fire strategies where safe shutdown cabling was not identified, and initiation of procedure changes to include the appropriate SSD equipment. This issue was entered into the licensees CAP (CAP 1484142). The inspectors determined that the failure to maintain fire strategy procedures to ensure that SSD equipment was identified was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Self-Assessment aspect because of the licensees failure to conduct self-critical and objective assessments of its programs and practices.
05000282/FIN-2015003-012015Q3Prairie IslandFailure to Determine Compensatory MeasuresA finding of very low safety significance with two examples and an associated non-cited violation of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14. Specifically, on two occasions, the licensee failed to determine compensatory measures following the identification of a Updated Safety Analysis Report (USAR) non-conforming condition associated with the Units 1 and 2 residual heat removal (RHR) recirculation sump valves on August 31, 2015, and for a degraded condition of the Unit 1 B RHR recirculation sump valves on September 14, 2015. The licensee entered the issues into the Corrective Action Program (CAP) as CAPs 01491302 and 01491900. The inspectors determined that the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14, to properly determine compensatory measures for operable but degraded and operable but non-conforming conditions was a performance deficiency. The performance deficiency, with two examples, was determined to be more than minor and a finding in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed on two occasions to properly determine compensatory measures to maintain or enhance operability of Technical Specification (TS) Systems, Structures, and Components (SSCs) that were not fully qualified until final corrective actions were taken. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that because the finding was a qualification deficiency and did not impact operability of the TS SSCs, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor for the performance deficiency was associated with the cross-cutting aspect of Consistent Process in the Human Performance cross-cutting area, involving individuals using a consistent, systematic approach to make decisions. Specifically, the licensee did not apply a consistent, systematic approach for determining the appropriateness of compensatory measures while making operability decisions for the degraded and non-conforming conditions associated with the RHR recirculation sump valves.
05000282/FIN-2015003-022015Q3Prairie IslandImproper Operability DeterminationA finding of very low safety significance and an associated non-cited violation of Title 10, CFR, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14. Specifically, on August 9, 2015, following the discovery of a non-functional D6 building ventilation system and declaration of inoperability of Buses 26, 221, 222, and the D6 DG, the licensee improperly declared the affected TS SSCs operable and fully qualified without restoring functionality of the ventilation TS support system or implementing appropriate compensatory measures per the requirements of FP-OP-OL-01. The licensee entered the issue into the Corrective Action Program as CAP 01490027. The inspectors determined that the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14 was a performance deficiency. The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee improperly declared the TS SSCs operable and fully qualified without restoring functionality of a TS support system or implementing appropriate compensatory measures. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered No to all questions under Section A, therefore the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor for the performance deficiency was associated with the cross-cutting aspect of Challenge the Unknown in the Human Performance cross-cutting area, involving individuals stopping when faced with uncertain conditions and evaluating and managing risk prior to proceeding. Specifically, the licensee did not properly evaluate and manage uncertain conditions associated with the ventilation system and impacts on TS SSC operability prior to proceeding with declaration of full qualification.
05000282/FIN-2015003-032015Q3Prairie IslandLicensee-Identified ViolationTitle 10, CFR Part 50.72(b)(3)(xiii) states, in part, a licensee shall report (notify the NRC as soon as practical and in all cases within 8 hours of the occurrence) any event that results in a major loss of emergency assessment capability. Contrary to this requirement, over the past 3 years, the licensee identified six instances (on August 14, 2012; November 16, 2012; November 18, 2012; November 21, 2012; December 5, 2012; and January 16, 2013) of a failure to report the major loss of emergency assessment capability when the Seismic Monitoring Panel was non-functional for unplanned events. The licensee also identified three instances (on December 14, 2012; September 3, 2014; and September 30, 2014) of a failure to report the major loss of emergency assessment capability when the Seismic Monitoring Panel was non-functional for planned events for greater than 24 hours. The system degradation adversely impacted the sites ability to make an ALERT and a Notice of Unusual Event Emergency Action Level assessment in accordance with PINGP-1575, Emergency Action Level Matrix, and F3-2.1, Emergency Action Level Technical Bases. The licensee entered the issue into the corrective action program as CAP 01472229, OE Review of NRC Event Reports Related to Seismic Monitors, CAP 01472731, Missed Reportability for Seismic Monitor Out of Service, and CAP 01486147, Potential Licensee ID Violation from EP Inspection. The licensee completed the required report to the NRC on April 2, 2015 (Event Number 50948, Seismic Monitor Not Available for Emergency Plan Assessment). The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the generic communications input that 10 CFR 50.72 reports serve. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement process. The inspectors determined that this issue was a Severity Level IV violation based upon Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, example d.9 in the NRC Enforcement Policy. Example d.9 specifically states, A licensee fails to make a report requirement by 10 CFR 50.72 or 10 CFR 50.73. Because the issues were entered into the licensees corrective action program as CAPs 01472229, 01472731, and 01486147, the violation is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000263/FIN-2015003-052015Q3MonticelloFailure to Provide Complete and Accurate Information in LER 05000263/2015-002-00The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9 due to the licensees failure to provide information to the NRC that was complete and accurate in all material respects in accordance with the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, on June 29, 2015, the licensee failed to include an accurate assessment of the safety consequences and implications of a loss of shutdown cooling event when they issued LER 05000263/2015-002-00. This LER included an inaccurate assessment of safety implications, stating that engineering calculations show a potential worst case maximum temperature of 115 degrees Fahrenheit (F). The inspectors identified that engineering models actually showed potential worst case temperatures of 25-26 degrees F higher, which could have challenged or exceeded fuel pool cooling design specifications. Corrective actions included issuance of a revision to LER 2015-002-00 which contained the correct engineering modeling results and associated discussion of safety implications. The licensee entered this issue into its CAP (CAP 1484633). This issue was of more than minor significance under the Traditional Enforcement Process because the NRC relies on licensees to identify and correctly report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., loss of shutdown cooling) was evaluated separately and determined to be a finding of very low safety significance as documented in the 2015 2nd Quarter Integrated Inspection Report (05000263/2015002-01). In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was of more than minor concern with relatively inappreciable potential safety significance and is related to a finding that was determined to be a more than minor issue. Consistent with Example 6.9.d.1, this represented an example where the licensee submitted inaccurate information in a required report, which resulted in expansion of the scope of the next regularly scheduled inspection and required LER revision. Because there was no finding evaluated with this violation, the inspectors did not assign a cross-cutting aspect to this issue.
05000263/FIN-2015003-022015Q3MonticelloFailure to Perform High Radiation Area Portable Fire Extinguisher SurveillancesThe inspectors identified a finding of very low safety significance and an NCV of Technical Specification (TS) 5.4.1.d when the licensee failed to implement procedures associated with Fire Protection Program Implementation, to ensure that required refueling outage surveillances were performed for fire extinguishers located in high radiation areas (HRAs). Specifically, between March 2007 and May 2015, the licensee failed to implement steps 9 and 10 of 1123, Portable Fire Extinguishers, which required weighing and verifying adequate hydrostatic testing of the fire extinguishers in HRAs on a refueling outage frequency. Corrective actions included surveillance process changes and evaluation of the current status of the high radiation area fire extinguishers which resulted in the determination that outside of the surveillance process, a separate work activity had exchanged all the affected extinguishers with ones that were current on their surveillances in May 2015. This issue was entered into the licensees Corrective Action Program (CAP) 1484257 The inspectors determined that the failure to implement HRA fire extinguisher surveillances was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Work Management aspect because of the failure to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority and the failure to identify the need for coordination with different groups or job activities
05000263/FIN-2015003-042015Q3MonticelloDrywell to Torus Vacuum Breaker Past OperabilityDuring the cycle preceding the 2015 refueling outage, two evaluations associated with torus to drywell vacuum breaker operation were developed due to issues identified in the first quarter 2014. These included: CAP 1417977, Failure of drywell-torus vacuum breaker to close, which identified an occasion of dual indication during Procedure 0143 procedure. A second occurrence was observed several days later and was documented in CAP 1418471, AO-2382A Torus-to-DW vacuum breaker closed indication anomaly. CAP 1420318, DW-Torus vacuum breaker work performed with inadequate PMT, identified the PMT following shaft sealing component (O-ring) replacement during the 2013 outage was not performed as planned. The licensee evaluations for these CAP conditions concluded the Drywell to Torus vacuum breakers were operable. However, neither evaluation specifically considered the effect of an interference between the vacuum breaker test lever and vacuum breaker test actuator stem. Since this specific mechanism was not addressed in these two evaluations, past operability of the torus to drywell vacuum breakers was questioned. As a result, the licensee established a past operability evaluation be conducted via CAPs 1479198 and 1478212. The licensee completed its past operability evaluation on June 26, 2015. After review, the inspectors conveyed a number of questions to the licensees engineering staff in regard to the past operability evaluation. Although the licensee provided responses for the majority of these questions during the remainder inspection quarter, the licensee had requested external input in regard to one of the inspectors questions. Specifically, inspectors questioned whether it was possible for the bottom of the lever arm to be at an elevation above the top of the actuator stem at valve disc full open and if so, could the valve test lever arm have come to rest on top of the actuator stem, potentially impacting the ability of the vacuum breaker valve to close. Upon the close of this inspection period, that input had not yet been finalized and made available to the inspectors. As a result, this issue was considered to be an unresolved item pending a review of the licensees response and past operability for CAPs 1479198 and 1478212, including and the licensee response to open inspector questions.
05000263/FIN-2015003-012015Q3MonticelloInadequate Evaluation of Refueling Floor Structural Steel BeamsThe inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to 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. Specifically, on September 3, 2008, licensee personnel failed to verify the adequacy of design when they failed to use correct section properties in their calculation of stresses on structural steel beams supporting the refueling floor for the increased spent fuel cask loading. Reevaluation of the beams using correct methodology resulted in the conclusion that the beams would not meet the design basis stress limits. Immediate corrective actions for this issue included initiation of a CAP, performance of a functionality assessment which concluded that the refueling floor remained functional but non-conforming, and creating compensatory measures which limited the refueling floor live load in the cask loading area (CAP 1492837). The inspectors determined that the licensees calculational methodology was contrary to the standard engineering principles applicable for determination of stresses in structural members, which resulted in a failure to meet Criterion III, Design Control, and was a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612 because it was associated with the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers (reactor building) protect the public from radionuclide releases caused by accidents or events. Additionally, More than Minor Example 3.j of IMC 0612, Appendix E, Examples of Minor Issues, was used to inform the more than minor screening. The inspectors used IMC 0609, SDP, Attachment 4, Initial Characterization of Findings, and Appendix A of IMC 0609 to screen this finding. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. As a result, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000282/FIN-2015002-032015Q2Prairie IslandFailure to Make an 8-Hour Report Required by 10 CFR 50.72(b)(3)(ii)(B)The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.72(b)(3)(ii)(B) due to the licensees failure on August 8, 2014, to report an unanalyzed condition within eight hours of discovery. Specifically, the lack of fuse protection for the emergency bearing oil pump control circuitry created an unanalyzed condition due to the potential for a fire that impacted the licensees safe shutdown capabilities. The inspectors determined that the failure to submit a report required by 10 CFR 50.72 for the unanalyzed condition described above was a performance deficiency. The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the information that 10 CFR 50.72 reporting serves. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement process. The inspectors determined that this issue was a Severity Level IV violation based on Example 6.9.d.9 in the NRC Enforcement Policy. Example 6.9.d.9 specifically states, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. Because the licensee identified the technical issue as part of their NFPA-805 transition process, and no additional or separate NRC-identified or self-revealed more-than-minor Reactor Oversight Process findings were noted, there was no cross-cutting aspect associated with this violation.
05000282/FIN-2015002-062015Q2Prairie IslandLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix R, requires, in part, that fire protection features shall be provided for SSCs important to safe shutdown. These features shall be capable of limiting fire damage so that one train of systems needed to achieve and maintain hot shutdown from either the control room or emergency control station(s) is free of fire damage and that equipment needed to achieve and maintain cold shutdown can be repaired within 72 hours. In addition, where cables and equipment located outside of containment could prevent equipment operation or cause miss-operation due to hot shorts, open circuits or shorts to ground of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, separation of cables and equipment must be maintained through the use of a fire barrier with a three-hour rating to ensure one train of redundant equipment remains free of fire damage. On October 13, 2011, the licensee failed to provide fire protection features for SSCs important to safe shutdown that limited fire damage such that one train of systems remained free from fire damage, in accordance with 10CFR Part 50, Appendix R. Specifically, the licensee identified that Rodofoam material present in the auxiliary building seismic joint seals failed to provide a three hour fire barrier to ensure that one train of redundant safe shutdown equipment remained free from fire damage following a fire due to Rodofoam being a combustible material. The inspectors reviewed this issue and determined that the Rodofoam material was part of the initial plant design. In addition, these seals were not identified as fire penetration seals. As a result, the seals were not considered to be part of the fire protection program. Once the Rodofoam material was found, the licensee initiated periodic fire watches in the impacted areas and initiated CAP 1308129. The fire watches remained in place until the Rodofoam seals were replaced with a non-combustible seal material. The inspectors determined that the failure to ensure that equipment was protected from fire, such that one train of equipment remained free from fire damage was a performance deficiency and a violation of 10 CFR 50, Appendix R. However, Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non-compliances identified as a result of a licensees transition to the new risk-informed, performance-based fire protection approach included in 10 CFR 50.48(c) and for certain existing non-compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk-informed, performance-based approach is referred to as NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. In 2005, the licensee began the process of transitioning from the requirements of 10 CFR 50, Appendix R to NFPA 805. This process included submitting a licensing amendment to the NRC for review and approval in September 2012. The inspectors reviewed the criteria included in Section 9.1 of the NRC Enforcement Policy and concluded that the licensee had met the criteria for enforcement discretion. Specifically, the licensee entered the noncompliance into the CAP as CAP 1308129 and implemented compensatory fire watches in the area until the seals were replaced with an appropriate material. Additionally, this issue would not have been identified under normal surveillance or QA activities. This issue was not willful since the seismic gap seals were installed prior to the development of the fire protection requirements. The inspectors evaluated the significance of this finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Finding, dated June 19, 2012, and determined that the finding affected the Mitigating System cornerstone. The inspectors determined that the finding degraded fire protection defense-in-depth strategies so IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, was used to determine the safety significance. The inspectors concluded that this issue was of very low safety significance because the credited safe shutdown equipment was located more than ten feet horizontally or vertically away from the flammable seal material. As a result, a credible fire on either side of the flammable seal material would not result in damage to the redundant safe shutdown equipment on the other side. Because there were no redundant cables or equipment penetrating the seal area, the inspectors concluded that hot gases, which could penetrate the seal, would cool and disperse, such that redundant cables and equipment would not have been damaged. Therefore, no credible fire could affect the ability to achieve and maintain safe shutdown. Because each of the criteria listed in Section 9.1 of the NRC Enforcement Policy was met, the NRC concluded that enforcement discretion should be granted for this issue. No enforcement action will be documented unless the licensee fails to address this non-compliance after completing their transition activities.
05000282/FIN-2015002-052015Q2Prairie IslandLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix R, requires, in part, that safe shutdown equipment and systems for each fire area shall be known to be isolated from associated non-safety circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the circuit will not prevent operation of the safe shutdown equipment. The isolation of these associated circuits from the safe shutdown equipment shall be such that a postulated fire involving the associated circuits will not prevent safe shutdown. On August 8, 2014, the licensee identified an Appendix R non-compliance in that the emergency bearing oil pumps were not properly isolated (fuse protected) from safe shutdown equipment, in accordance with 10 CFR Part 50, Appendix R. As a result, an overload condition in the emergency bearing oil pump circuitry could result in a fire that damages other cabling and prevents the licensee from achieving safe shutdown following a fire. The inspectors reviewed this issue and determined that the improper fuse protection was part of the initial plant design. Specifically, the design philosophy in the late 1960s was to maximize the reliability and availability of the emergency bearing oil pumps to protect the main turbines. The potential impact that this design philosophy had on fire protection of safe shutdown equipment was also not recognized as 10 CFR 50, Appendix R, did not exist until the early 1980s. The licensee documented this issue in CAP 1442220. The licensee also implemented hourly fire watches in the impacted fire areas to ensure that any potential fires were identified prior to it impacting safe shutdown capability. Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non-compliances identified as a result of a licensees transition to the new risk-informed, performance-based fire protection approach included in 10 CFR 50.48(c) and for certain existing non-compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk-informed, performance-based approach is referred to as NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. In 2005, the licensee submitted a letter of intent to transition to 10 CFR 50.48(c). This licensee submitted a license amendment request to the NRC for review and approval in September 2012. The inspectors reviewed the remaining criteria included in Section 9.1 of the NRC Enforcement Policy and concluded that the licensee had met the criteria. Specifically, the licensee entered the noncompliance into the CAP as CAP 1442220, implemented compensatory fire watches in the area and the noncompliance was not willful. In addition, this issue would not have been identified under normal surveillance or quality assurance activities. Lastly, a regional SRA reviewed an analysis performed by the licensee to show that the risk of the condition was less than high safety significance (i.e., less than red). The licensee identified the cable routing for the six cables of concern (three for each unit) and the fire scenarios where an initial fire could cause a secondary fire in a separate fire area due to the inadequate fusing of the emergency bearing oil pumps. The licensees evaluation assumed that a secondary fire would be limited to the cable tray that contained the faulted cable and would not propagate beyond that tray. The licensee cited NFPA 805 FAQ13005, Close-out of Fire Probabilistic Risk Assessment Frequently Asked Question 13005 on Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting, that provided similar guidance for self-ignited cable fires as the basis for the assumption. The SRA consulted with NRC Headquarters staff and concluded that the FAQ guidance did not specifically apply to cable fires resulting from inadequate fusing. However, there currently is no available method for estimating the likelihood and extent of a secondary cable fire caused by inadequate fusing. Given the lack of an acceptable method, the SRA also performed a walk down of the control cable routing to observe the potential for secondary fires to impact additional targets. In all cases, there did not appear to be a significant potential for a secondary fire to damage additional targets beyond the cable tray of interest. The licensee provided other reasons why secondary fire damage would be limited, such as existing fire detection and suppression systems and the fact that the cables are thermoset rather than thermoplastic material. The SRA determined that the likelihood of significant secondary fire spread for these particular scenarios was low. The licensee also determined that some scenarios did not impact any unique targets. For those scenarios, there was no change in risk due to the inadequate fusing. For scenarios that did have the potential for additional target damage from a secondary fire, the licensee calculated the change in risk of this condition. The change in risk was determined to be less than 1E4/yr. The dominant fire scenarios involved a fire starting in the fire area 18 with a secondary fire propagating to either Fire Area 58, 31, or 32. Because each of the criteria listed in Section 9.1 of the NRC Enforcement Policy was met, the NRC concluded that enforcement discretion should be granted for this issue. No enforcement action will be documented unless the licensee fails to address this non-compliance after completing their transition activities.
05000282/FIN-2015002-042015Q2Prairie IslandDesign Control Measures not Implemented to Ensure Group E Pressurizer Heaters Remain Operational Post-FireThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, on October 13, 2014, for the licensees failure to ensure the design requirements of the fire protection program were maintained. Specifically, the licensee had not ensured that Group E pressurizer heaters would continue to operate following a fire in Fire Area 32 (the Unit 1 side of the auxiliary feedwater pump room). As a result, the licensee was unable to ensure that the Unit 1 reactor would be able to achieve and maintain a cold shutdown condition following a fire in this area. The inspectors determined that the failure to ensure the design requirements of the fire protection program were maintained was contrary to 10 CFR 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The finding was more than minor because it was associated with the Protection from External Factors attribute of the Mitigating Systems cornerstone. The finding also impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors utilized IMC 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and determined that this finding was best assessed for safety significance by using IMC 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors used IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, dated September 20, 2013, and assigned a Post-Fire Safe Shutdown fire inspection finding category to the issue per Step 1.2. Based upon the information contained in Step 1.3 of IMC 0609, Appendix F, Attachment 1, the finding was determined to be of very low safety significance because any fire related damage to the Group E pressurizer heater cables did not impact the licensees ability to reach and maintain a safe shutdown condition (either hot or cold). No cross-cutting aspect was assigned to this issue since the missed opportunities to identify this issue occurred more than three years ago and were not reflective of current performance.
05000282/FIN-2015002-012015Q2Prairie Island
  1. 12 Battery Charger Design Control
The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to ensure the design requirements of the #12 battery charger were maintained. Specifically, the licensee failed to address the impact that previously identified additional electrical loads had on the design capacity of the battery chargers from May of 2010 until April of 2015. The inspectors determined that the failure to maintain the design basis for the battery charger was contrary to 10 CFR 50 Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to account for the additional electrical load of the inverters on the #12 battery charger. This additional load exceeded the battery chargers design capacity and as a result, the licensee could not demonstrate that the #12 battery charger would be capable of responding to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, issued June 19, 2012, the inspectors answered Yes to Question 2 of the Mitigating SSCs and Functionality screening questions because the finding represented a loss of function to the #12 battery charger. Thus the inspectors consulted the regional senior reactor analyst (SRA) for additional assistance and the finding was determined to be of very low safety significance (Green). No cross-cutting aspect was assigned to this issue as the actions taken in 2011 were not reflective of current performance.
05000282/FIN-2015002-022015Q2Prairie IslandFailure to Correct #12 Battery NonconformanceThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to correct a non-conforming issue for the #12 battery that was discovered in February 2011. The inspectors determined that the failure to correct the non-conformance in a timely manner was contrary to 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and was a performance deficiency. The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee did not take timely corrective actions to resolve the #12 battery non-conformance. Additionally, no corrective action was taken to correct the occurrence of the inverters AC circuit breakers tripping of the normal load and becoming an additional load on to the DC system; thereby causing the battery to be non-conforming. In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, issued June 19, 2012, the inspectors answered No to all of the questions. The inspectors confirmed that the finding did not result in a loss of operability or functionality per IMC 0326, Operability Determination & Functionality Assessments for Conditions Adverse to Quality or Safety, since the capacity of the battery had been tested above the 88.5 percent capacity factor per battery calculation and evaluation. Therefore, this finding was of very low safety significance (Green). The inspectors determined the finding was cross-cutting in the Problem, Identification and Resolution, Resolution area because of the licensees failure to implement effective corrective actions to restore operability of the #12 battery.
05000282/FIN-2015001-032015Q1Prairie IslandUntimely Resolution of Environmental Qualification IssuesA self-revealing finding of very low safety-significance and a non-cited violation of 10 CFR 50.49 was identified on March 5, 2015, for the licensees failure to keep environmental qualification (EQ) files current and the failure to replace or refurbish EQ electrical equipment at the end of its designated life. Specifically, the licensee initiated CAP 1431268 in May 2014 to document numerous EQ file errors identified during an in-depth review of the EQ program. These file errors resulted in the EQ designated life for multiple safety-related solenoid valves being non-conservative such that some solenoids were installed beyond their designated life. Corrective actions included taking action to revise the incorrect EQ files and replacing the safety-related solenoids installed beyond their designated life. The inspectors determined that this issue was more than minor because if left uncorrected the failure to maintain the EQ files and to replace or refurbish EQ equipment could result in a more significant safety concern. Specifically, the inaccurate files could result in EQ equipment not being refurbished or replaced as required. In addition, the failure to replace or refurbish EQ equipment installed beyond its designated life could result in equipment failure during normal operation or post-accident conditions. The inspectors utilized IMC 0609, Attachment 0609.04, Initial Characterization of Findings, and determined this issue was of very low safety significance because each of the questions provided in IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, was answered No. The inspectors concluded that this issue was cross-cutting in the Problem Identification and Resolution, Evaluation area because the licensee had not thoroughly evaluated CAP 1431268 to ensure that the resolution addressed the causes and extent of condition commensurate with the safety significance (P.2).
05000282/FIN-2015001-022015Q1Prairie IslandFailure to Follow Foreign Material Exclusion Procedure during Reactor Coolant Pump Seal ReplacementA self-revealing finding of very low safety significance and associated NCV of TS 5.4.1 was identified on December 19, 2014, due to the licensees failure to follow Procedure FPMAFME01, Foreign Material Exclusion and Control. Specifically, workers failed to implement and adhere to the foreign material exclusion (FME) control requirements for a Level 1 foreign material exclusion area when replacing the Unit 1 reactor coolant pump (RCP) seals and associated piping during Refueling Outage 1R29. The failure to implement and adhere to the FME control requirements resulted in introducing foreign material into the reactor coolant system and the subsequent degradation of the #12 RCP seal in December 2014 and January 2015. The seal degradation led to two Unit 1 reactor shutdowns. Corrective actions for this issue included replacing the RCP seal, flushing the seal piping and establishing a process to review work document quality to ensure that appropriate programmatic requirements were included. The inspectors determined that the failure to follow Procedure FPMAFME01 was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors utilized Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Work Management area, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. In addition, the work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities (H.5).
05000306/FIN-2015001-012015Q1Prairie IslandFailure to Perform Immediate Operability Determination for 14 CFCU as Required by ProcedureAn inspector identified finding of very low safety significance and a NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," occurred on January 27, 2015, due to operations personnel failing to follow Procedure FPOPOL01, Operability/Functionality Determination, while assessing the operability of the 14 containment fan coil unit (CFCU) and the Unit 1 containment. Specifically, personnel failed to perform an immediate operability determination for the 14 CFCU and the Unit 1 containment after the inspectors identified that the 14 CFCU was potentially leaking. Corrective actions for this issue included documenting the immediate operability determination after the inspectors brought this issue to the attention of the operations department and sharing the details of this event with other operations personnel. The inspectors determined that the failure to perform an immediate operability determination on the 14 CFCU and the Unit 1 containment as required by Step 5.3.1 of Procedure FPOPOL01 was more than minor because if left uncorrected, the failure to perform operability determinations, as required by procedure could result in incorrect/untimely operability conclusions and the failure to take action to correct degraded or deficient conditions, as required by the technical specifications (TS). In addition, this is the second example of an untimely CFCU operability determination identified by the inspectors in the last ten months. The inspectors utilized IMC 0609, Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, Part B, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Teamwork area because individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4).
05000282/FIN-2014005-012014Q4Prairie IslandFailure to Implement the Winter Plant Operation ProcedureThe inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, on December 4, 2014, due to the licensees failure to follow procedure during the performance of test procedure (TP) 1637, Winter Plant Operation. Specifically, maintenance personnel failed to comply with a step within TP 1637 which directed that a tent and heater be installed around the Unit 2 cooling water (CL) discharge to grade header to prevent ice buildup and subsequent blockage during freezing conditions. Consequently, the inspectors identified ice buildup on the CL header discharge orifice which if left uncorrected, could result in header blockage and subsequent inoperability. Corrective actions for this issue included removing the ice buildup on the cooling water discharge header, installing a tent and heater in accordance with TP 1637, revising the associated procedures and performing an apparent cause evaluation. The inspectors determined that this issue impacted the Mitigating Systems cornerstone and was more than minor because if left uncorrected, this issue could become a more significant safety concern. Specifically, with freezing conditions present coupled with the existence of leakage and resultant ice buildup on 20CL61, the potential existed for subsequent ice blockage if left uncorrected and resultant inoperability of the cooling water system. This issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. The inspectors concluded that this finding was associated with a conservative bias cross cutting aspect in the human performance cross cutting area. Specifically, operations and maintenance personnel did not utilize prudent decision making practices to ensure the cooling water header was adequately protected against freezing conditions (H.14).
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