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05000261/FIN-2018410-0130 June 2018 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2017007-0431 December 2017 23:59:59RobinsonNRC identifiedCrouse-Hinds Qualification and Life ExtensionIntroduction: The inspectors identified an unresolved item (URI) involving three separate concerns that could affect the qualification of Robinsons Crouse-Hinds (C-H) electrical penetration assemblies (EPAs). First, the inspectors were concerned that a similarity analysis, which fulfilled the requirements of Commission memorandum and Order CLI 80-21, In the matter of Petition for Emergency and Remedial Action, and 10 CFR 50.49, Environmental Qualification for Electric Equipment Important to Safety for Nuclear Power Plants, may not have been completed. Second, the inspectors were concerned that Robinson may not have demonstrated that the penetrations electrical performance specifications were met using appropriate IEEE standards, as stated in the UFSAR. Third, the inspectors were concerned that the licensee may not have used appropriate methods when extending the qualified life of the C-H EPAs. Description: (1) In Robinsons initial Bulletin 79-01 response dated June 1980, to justify the qualification of the C-H EPAs by similarity, Robinson submitted a Westinghouse (WEC) qualification report AB-11/12/73, Qualification Tests for a Modular Penetration 5 dia. (Prototype B1), obtained from Brunswick nuclear station; a record of a phone conversation between Robinson and WEC, CPL-77-550, dated 11/29/1977; and a WEC design specification for the C-H EPAs, CPL-R2-E3, dated 6/26/1968. In the technical evaluation report (TER) dated July 8, 1982, that accompanied the NRC staff safety evaluation report (SER) dated January 5, 1983, 10 regarding the Robinson EQ Program, the C-H EPAs qualification was identified as Category IV Documentation Not Available. In the 1982 TER and NRC SER, these specific submitted documents were listed as reviewed and, the qualification of the C-H EPAs remained Category IV. In a licensee letter, dated March 2, 1984, the licensee documented a meeting with the NRC staff discussing Robinsons proposed methods of resolution for each of the EQ deficiencies identified. Robinson appeared to commit to documenting a similarity analysis between their C-H manufactured EPAs and other similar EPAs found acceptable by the NRC staff. In the 1985 final NRC SER, the staff found Robinsons proposed method of resolution specified in the March 2, 1984 letter, acceptable. However, the 1984 submittal summarized a January 18, 1984 meeting with NRC where it was stated the NRC would not perform any additional equipment review and it was left up to the utility to state the adequacy of the documentation. During the inspection, Robinson provided the documents originally submitted (AB- 11/12/73, CPL-77-550, and CPL-R2-E3) to the inspectors to justify qualification by similarity. The inspectors had concerns with these documents justifying similarity between the WEC and C-H EPAs. a) In a review of AB-11/12/73 and comparing it to what was known about the C-H EPAs, the inspectors identified that the materials used in the WEC EPAs were not identical or sufficiently similar in material composition or performance specifications. The WEC tested EPAs used silicone rubber O-rings, a proprietary WEC composition Q epoxy resin potting material as the internal filler, and had a 5 diameter. The C- H EPAs did not use O-rings, used room temperature vulcanized (RTV) silicone rubber potting material as the internal filler, a thin layer of Sty-Cast epoxy resin to seal the end opening exposed to a DBA, and has an approximately 11 diameter. b) The inspectors noted the performance requirements demonstrated by the WEC pressure tests did not appear to envelope the required Robinson DBA pressure performance. The WEC maximum pressure only developed 1286.9lbf at 105psig, and the C-H EPA would develop 3955.2lbf at 42 psig. The effects of the more substantial forces on the C-H EPAs was not addressed. c) In the review of specification, CPL-R2-E3, the inspectors noted that specification CPL-R2-E3 was actually an EBASCO specification rather than a WEC specification as had been stated, and that C-H had taken exception to the specification due to chemical incompatibilities between the RTV potting material and cable insulations specified by EBASC O. Many of the Robinson documents still specify these incompatible cable insulations for use with the C-H EPAs without justification. d) In the review of CPL-77-550, the inspectors noted that the record of the phone call did not have any suitably specific information that could justify similarity to the C-H in materials, performance specifications, or manufacturing methods. The inspectors are concerned that Robinson was unable to provide an acceptable similarity analysis to address the deviati ons between the tested and installed EPAs. The licensee entered this concern into t heir corrective action program as NCR 2161911, and determined the equipment was operable. 11 (2) Robinsons UFSAR Section 3.8.1.2 stated, in part, that electrical penetrations are designed and demonstrated by test to withstand, without loss of leak tightness, the containment post-accident environment and to meet the National Electric Code, IEEE - Proposed Guide for Electrical Penetration Assemblies in Containment Structures for Stationary Nuclear Power Reactors or subsequent issues of this standard, IEEE Electric Penetration Assemblies in Containment Structure for Nuclear Power Generating Stations (IEEE 317). In accordance with the IEEE 317 versions reviewed from 1971 to 1976, the performance requirements are to be met by test during all conditions from mild plant conditions (normal) to the most limiting environmental conditions produced during DBAs (accident), and post-accident conditions. When asked to provide the test documentation that met these original requirements, Robinson was not able to provide them. In addition, the inspectors noted that electrical calculation RNP-E-5. 30, Crouse-Hinds Electrical Penetration Ampacity, Short Circuit, and Heat Generation Calculation, revision 6, indicated that the current plant design exceeded the electr ical performance specification for some of the C-H EPAs, and thus these EPAs would not meet the UFSAR and IEEE 317 specifications. The inspectors requested evidence that Robinson met the required verifications testing specified in the UFSAR Section 3.8.1.2, and that those test conditions are bounding of the current electrical plant design described in RNP-E-5.30. The inspectors are concerned that Robinson may not be in conformance with statements in the UFSAR and 10 CFR 50, Appendix B, Criterion III, Design Control, which required, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. The licensee entered this issue into their corrective action program as NCRs 2159165 and 2164589. (3) The inspectors identified two concerns with the way Robinson extended the qualified life of the C-H EPAs. First, Robinson reverse calculated an activation energy which appears to be outside of known acceptable Arrhenius techniques. Second, Robinson derived activation energies from EPAs with materials that were not the same as in the C-H EPAs. The inspectors noted that the Division of Operating Reactors (DOR) guidelines, Guidelines for Evaluating Qualification of Class 1E Electrical Equipment in Operating Reactors, and NUREG 0588 both accepted Arrhenius techniques as acceptable methods for determining the qualified lives of components, and required that the materials be identical or be justified by analysis. For the first concern, UFSAR Section 3.11.3, Qualification Tests Results, specified the EQDPs contained the qualification justification analysis for EQ components. The EQDP-0900, for the C-H EPA, credited the WEC EQ report AB-11/12/73 for thermal aging life calculation. The WEC EQ report applied Arrhenius techniques in accordance with IEEE 98-1972, IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials, and IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data. The WEC EQ report indicated that they had determined an activation energy and the confidence bounds, but they did not include this information or the data used to derive it. The omitted information would be required to identify the limitations of what WEC had derived for their thermal aging. To derive the pseudo activation energy and extend the life of the C-H EPAs from 40 to 60 years, Robinson applied 12 an Arrhenius equation and discounted the limitations involved with using the Arrhenius extrapolation techniques as specified in known quality standards. For the second concern, the inspectors determined that there were material deviations between the WEC and C-H EPAs that could potentially invalidate the pseudo activation energy Robinson derived. Robinson derived a 1.018eV activation energy, when the silicone RTV known to be used in construction of the C-H EPA had a more limiting activation energy of 0.63eV. The 0.63eV would have significant negative effect on the qualified life of the C-H EPA, invalidating the life extension and current EQ status. In addition, the inspectors noted that in the Robinson license renewal application and safety evaluation report, NUREG 1785, Section 4.4.1.1, Summary of Technical Information in the Application, the licensee appeared to commit to using the Arrhenius method, as described in Electric Power Research Institute (EPRI) NP- 1558, A Review of Equipment Aging Theory and Technology. The inspector noted that NP-1558 was not a quality standard as required by general design criteria 1 and 10 CFR 50.54(jj); however, its use would have likewise invalidated the WEC information for the C-H life extension. The inspectors are concerned that despite the specifications in the IEEE quality standards and the information in EPRI report NP-1558, Robinson extrapolated an invalid qualified life for the EPAs possibly making them unqualified to withstand a DBA. The licensee entered this concern into their corrective action program as NCR 2164567. This URI is opened to determine if a performance deficiency or a violation exists. To resolve the various aspects of this URI, the inspectors need: (1) Actual material and performance specification similarity analysis or confirmation of licensing basis; (2) The documented verification testing that satisfies statements in UFSAR 3.8.1.2, and confirmation that the electrical performance specifications tested are bounding of the current plant design; and 3) Confirmation that the actual penetration materials needed to be used when extending the qualified life, and what is required for appropriate application of Arrhenius techniques. (URI 05000261/2017007-04, Crouse-Hinds Qualification and Life Extension)
05000261/FIN-2017007-0131 December 2017 23:59:59RobinsonNRC identifiedFailure to Correctly Determine Qualified LifeThe NRC identified a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish a qualified life for the motors covered by Environmental Qualification Documentation Package (EQDP)-0 803 in accordance with their administrative procedure AD-EG-ALL-1612, Environmental Qualification (EQ) Program. Specifically, the licensee did not correctly establish a qualified life for the motors covered by EQDP-0803 due to a calculational error. In response to the issue, Robinson staff placed the issue in their corrective action program as NCRs 2155050 and 2158467, and demonstrated operability by removing conservatisms regarding assumptions for cumulative energized time of the motors. Additionally, the licensee plans to replace the affected motors. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not establishing the correct qualified life for the motors resulted in a reduction in margin that impacted the reliability of the equipment. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The inspectors determined that the finding was indicative of current licensee performance, because the error occurred on June 28, 2017. A cross-cutting aspect of Documentation (H.7) in the Human Performance Area was assigned because the organization did not create and maintain complete, accurate and up to-date documentation.
05000261/FIN-2017007-0331 December 2017 23:59:59RobinsonNRC identifiedFailure to Determine Most Severe Containment Spray pHThe NRC identified a non-cited violation of 10 CFR Part 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, for the licensees failure to correctly determine the most severe composition of chemicals for containment spray for the purposes of environment al qualification of equipment in containment. Specifically, the licensee did not identify that the pH of the chemical spray could have been more severe than what was identified in the Environmental Qualification zone maps if the Spray Additive Tank (SAT) had been operated at its limits provided in procedures CP-001 and OST- 023. In response to this issue, the licensee placed the issue into their corrective action program as NCR 2162081, demonstrated operability by reviewing current and historical operating conditions of the tank, and implemented administrative controls to prevent exceeding the qualified pH limit. This performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the containment spray pH could have exceeded the pH to which equipment inside containment was qualified, if the SAT had been operated at its procedural limits. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. A cross-cutting aspect was not assigned because the finding was not indicative of current licensee performance.
05000261/FIN-2017007-0631 December 2017 23:59:59RobinsonNRC identifiedPenetration F01 SubmergenceIntroduction: The inspectors identified a URI concerning the submergence qualification of Robinson EPA F-01. The qualification may not have qualified the EPA in accordance with NUREG-0588, Category 1 requirements. Description: In 1988, the licensee determined that penetration F-01 would become submerged and subsequently contracted testing to demonstrate qualification. The inspectors reviewed Wyle qualification test report 41175-1, and EGS qualification test report, EGS-TR-903200-04-R000. These two reports were credited for submergence in EQDP-1700 for the CONAX penetrations. The inspectors were concerned that the CONAX penetration F-01 was not tested in its most limiting configuration. To place the penetration pigtails in a configuration that could support qualification, the licensee performed a modification, MOD 977, Repairs to Protect Penetration F-01, to re-terminate the pigtails by adding Raychem heat shrink to provide submergence protection. Modification, MOD 977, specifically figure 1, drawing number C20482, and feed through detail drawing number B190670 revision 1, appears to allow 36 conductors to be bundled together in a single pass through. The EGS and Wiley test reports did not test the 36 conductor configuration or demonstrate that the signals passing through these bundles would remain operable for the duration of submergence as required by NUREG-0588, Category 1 requirements. The inspectors were also concerned that while the termination procedures in MOD 977 required a two inch Raychem overlap, it also allowed a one-half inch overlap during Raychem installation. A one-half inch overlap may not ensure submergence qualification in accordance with EGS qualification report EGS-TR-903200-04-R000. In addition, the EGS qualification used an 8.3 pH caustic solution during submergence testing, which is less than what was required for Robinsons harsh environment design basis (10.5 pH). Title 10 CFR 50.49(d)(3) and (e)(6), RG 1.89 revision 1, C.d.3.a, and NUREG 0588 Section 2.2(5) Qualification by Test, required that equipment that could be submerged must be qualified by testing in a submerged condition to demonstrate operability for the duration required. The inspectors are concerned that F-01 is not qualified for submergence and the pigtails may not meet the requirements for submergence qualification. The licensee entered this concern into their corrective action program as NCR 2167136. This URI is opened to determine if a performance deficiency or violation exists. To resolve this URI, the inspectors need the licensee to address the apparent lack of qualification required by NUREG-0588, Category 1 EQ requirements. (URI 05000261/2017007-06, Penetration F01 Submergence)
05000261/FIN-2017007-0231 December 2017 23:59:59RobinsonNRC identifiedFailure to Perform Required O-ring Replacement to Maintain QualificationThe NRC identified a non-cited violation of 10 CFR Part 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, for the licensees failure to correctly identify the maintenance required to maintain the core exit thermocouple reference junction box in a qualified state. Specifically, the licensee did not identify that the qualifying entity required that the cover O-ring be replaced on a 5 year frequency in addition to being replaced any time the junction box cover was removed, and due to this, the O-rings have not been replaced since original installation. In response to the issue, Robinson staff placed the issue in their corrective action program as NCRs 2157897 and 2161580, and demonstrated operability via analysis of the qualification test results. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining the equipment in its qualified configuration affected its reliability. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. A cross-cutting aspect was not assigned because the finding was not indicative of current licensee performance.
05000261/FIN-2017007-0731 December 2017 23:59:59RobinsonNRC identifiedJustification of Activation Energy of ASCO Solenoid Coil AssembliesIntroduction: The inspectors identified a URI concerning the qualified life of ASCO solenoid operated valves. The qualified life determined by the licensee utilized unvalidated information provided by a third-party, non-Appendix B vendor and discounted other critical materials in their weak-link analysis without providing justification in accordance with Regulatory Guide 1.89, Rev. 1. Description: In 2006, the Nuclear Utility Group for Environmental Qualification (NUGEQ) provided a letter suggesting methods to extend the qualified lives of the solenoid operated valves. The licensee modified the qualified life of their ASCO valves as described by NUGEQ and failed to validate and justify the informations acceptability for use. Inspectors determined that the use of MW-35 magnet wires activation energy in place of MW-16 was not appropriate as activation energies are material and failure specific, and are not transferrable between different material compositions. Furthermore, the inspectors determined that the licensee (and NUGEQ) failed to adequately justify the discounting of the other materials in the ASCO solenoid coils, which had lower activation energies than the MW-16 magnet wire as reported by ASCO in their qualification test reports. The failure to justify the discounting of MW-16 magnet wire and other identified limiting component of the ASCO coil assembly was a performance deficiency and a violation of 10 CFR 50.49. Regulatory Guide 1.89, Rev. 1, Regulatory Position 5.c requires, in part, that the basis upon which the rate and activation energy were established should be defined, justified, and documented. Contrary to the above, the licensee failed to justify and document their use of the MW-35 activation energy in place of all other identified limiting activation energies in the ASCO solenoid coil assembly. Additionally, 10 CFR 50.49(e)(5) requires, in part, that equipment be replaced before the expiration of its qualified life unless ongoing testing can demonstrate that the equipment has additional life. Contrary to the above, the licensee failed to demonstrate that the ASCO solendoid coil assemblies have additional life when they failed to justify their departure from ASCOs limiting activation energies. This URI is being opened to determine if this performance deficiency is more than minor. To resolve this URI, the inspectors need to review the licensees response to proposed questions regarding the validation and justification of the appropriate activation energy that will be used in determining the qualified life. (URI 05000261/2017007-07, Justification of Activation Energy of ASCO Solenoid Coil Assemblies)
05000261/FIN-2017007-0531 December 2017 23:59:59RobinsonNRC identifiedQuestions Regarding EQDP-0401 Method Used to Determine Activation Energy and Responsibility for VerificationIntroduction: The inspectors identified a URI concerning Robinsons requirement to verify the qualification of components (e.g., Rosemount transmitters) required to meet 10 CFR 50.49. Description: The Rosemount transmitters EQ described by Robinson EQDP-0401, referenced Wyle test report 45592-3 for qualification, which referenced NUREG 0588 Category 1 requirements. The Wyle report, Table III Aging Matrix, identified electronic components along with their respective activation energies (eV) and the references that identified the source of this information. The report specified that thin film metal resistors were the most limiting of these components. The reference for the thin film metal resistor activation energy was an IEEE white paper published in 1965, The Determination and Application of Aging Mechanisms Data in Accelerated Testing of Selected Semiconductors, Capacitors and Resistors. The validity of Wyles determination of activation energies was in question because their methods had not been validated, as stated in the IEEE white paper. The inspectors reviewed the other components in Table III of the Wyle report to verify what components were more limiting and determined that the metal film resistors were not the most limiting. The inspectors identified that the activation energy in the Wyle report for transistors was for metal enclosed transistors, 1.02eV, but the transistors used in the transmitter construction were actually plastic enclosed transistors with activation energies ranging from 0.5eV to 0.66eV. The transmitters used some carbon resistors that were more limiting than metal film resistors and were more sensitive to radiation synergisms. Further, the information in the IEEE white paper seemed to indicate a phase change with an associated more limiting activation energy in the range of the normal plant environmental temperatures. The licensee appeared to not have evaluated this phase change and used the less conservative activation energy from the IEEE white paper throughout their extrapolations. Finally, Robinson may not have reviewed the actual activation energy test data, the test plan and acceptance criteria for the activation energy, or information about the test program, or if any equivalent App. B program supported the informations quality. NUREG 0588 Section 5(2), specified that independent verification of similarity or equivalence must be established, and that it was incumbent on the applicant to have the necessary documentation to justify the adequacy of using data from similar or equivalent equipment. In addition, this Section 5(2) and NUREG 0588, Appendix E, specified, that for electrical equipment that will experience the environmental conditions of design basis accidents for-which it-must function, the licensee must provide: the qualification test plan, test setup, test procedures, acceptance criteria and a summary of test results that demonstrates the adequacy of the qualification program. Additionally, if analysis is used for qualification, justification of all analysis assumptions must be provided. Further, NUREG 0588 Section 4(5) specified that known material phase changes must be addressed; and Section 4(6) specified that the aging acceleration rate used during qualification testing, and the basis upon which the rate was established, should be described and justified. In NUREG 0588 Part II, the comment resolution to Section 4(6), it was specified that the testing of the equipment should be conducted using the most limiting (lowest) activation energy of the components. Standard IEEE 323-1974 Section 5, Principles of Qualification, specified, that principles and procedures for demonstrating qualification include assurance that any extrapolation or inference be justified by allowances for known potential failure modes and the mechanism leading to them. Section 5.1, Type Testing, specified that test alone satisfies qualification only if the equipment to be tested is aged, subjected to all environmental influences, and operated under post-event conditions to provide assurance that all such equipment will be able to perform their intended function for at least the required operating time. The inspectors identified other known failure mechanisms were not considered. For instance, electro-migration of aluminum in diodes, transistors, and Zener diodes present in the electronics has an activation energy between 0.5eV and 0.63eV, which is more limiting than what was used. This failure mechanism was identified in EPRI NP-1558, A Review of Equipment Aging Theory and Technology, and in many IEEE documents that were known at the time of qualification. Robinson used what appeared to be an unvalidated activation energy that also appeared to overlook a phase change that occurs within the licensees service conditions to extend the qualified life. The activation energy value and the method used to arrive at this value are in question. This URI is opened to determine if a performance deficiency or violation exists. To resolve the various aspects of this URI, the inspectors need to: (1) assess the validity of the methods used in the IEEE white paper, which includes addressing the apparent phase change; (2) assess the difference of the more limiting activation energies for the resistors used in the Robinson transmitters compared to the value the licensee is using (including addressing the more limiting activation energies for the other electronics in question); and (3) evaluate the self-heating effects of the junctions in the electronic components and its impact on activation energy. Finally, the inspectors need to assess what responsibilities and to what extent, the licensee has to ensure the activation energies provided by an Appendix B vendor, are accurate and reasonable. The licensee entered this concern into their corrective action program as NCR 2164598. (URI 05000261/2017007-05, Questions Regarding EQDP-0401 Method Used to Determine Activation Energy and Responsibility for Verification)
05000261/FIN-2017001-0331 March 2017 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non- confor mances are promptly identified. Contrary to the above, in March 2014, while performing examinations in steam generator C during forced shutdown RFO229F3, the licensee failed to identify a loose part lodged in contact with tube R37C22. The licensee identified the loose part in March 2017 during refueling outage RO30. The licensee verified that indications of the part were detectable during RFO229F3, retrieved the part, verified that degradation caused by the part met all structural integrity requirement s, plugged the tube, and removed it from service. This issue was identified in the licensees CAP as NCR 0210725. The inspectors evaluated this violation using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings At -Power, and determined that the violation was of very low safety significance (Green) because evaluations demonstrated that the tube could sustain three times the differential pressure across it during normal full power steady state operation and that the steam generator did not violate the accident leakage performance criterion
05000261/FIN-2017001-0131 March 2017 23:59:59RobinsonNRC identifiedFailure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak -Chase Test ConnectionGreen . An NRC- identified Green non -cited violation ( NCV ) of 10 CFR Part 50.55a, Codes and Standards, was identified for the failure to perform general visual examinations of moisture barriers in the containment leak -chase channel test connections in accordance with the American Societ y of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC), Section XI, Subsection IWE , Requirements for Class MC and Metallic Liners of Class CC Components of Light -Water Cooled Plants . Following the inspectors identification of this issue, t he licensee initiated actions to conduct the re quired visual examinations during the March 2017 refueling outage and initiated actions to revise the containment inservice inspection (ISI) plan such that the required examinations will be performed in the future . This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 02109909. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME B PVC , Section XI, Subsection IWE , was a performance deficiency (PD) . The finding was of more than minor significance because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, visual examinations of mois ture barriers associated with the containment leak -chase channel test connections provide assurance that the containment metal liner and liner seam welds remain capable of performing its intended safety function. In the absence of such examinations, corro sive conditions at the moisture barrier (concrete -to-tubing interface) could go undetected. As a result, degradation of inaccessible portions of the containment liner could progress to challenge the containment operational capability. Using IMC 0609, A ttachment 4, Initial Characterization of Findings, the finding was determined to affect the Barrier Integrity Cornerstone because it involved ISI program examinations designed to identify degradation of the containment metal liner. The inspectors screen ed the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At -Power, Exhibit 3 Barrier Integrity Screening Questions, and determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the containment. The inspectors reviewed this performance deficiency for cross -cutting aspects as required by IMC 0310, Components With Cross -Cutting Aspects. The finding was determined to be reflective of present licensee performance because in 2014, the licensee did not take effective corrective actions to implement the ASME BPVC 3 requirements in the Subsection IWE P rogram , when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014- 07, which highlighted this industry -wide problem. Therefore, the finding was assigned a cross - cutting aspect in the resolution component of the problem identification and resolution cross -cutting area (P.3)
05000261/FIN-2017001-0231 March 2017 23:59:59RobinsonNRC identifiedFailure to Submit Complete and Accurate Information for a Requested License AmendmentSeverity Level IV. An NRC -identified severity level IV (SL IV) NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, was identified for the licensees failure to provide complete and accurate information in a license amendment request (LAR), dated November 19, 2015, requesting extension of the containment leak rate test frequencies required by various containment technical specifications (TS s). In this LAR, the licensee incorrectly stated that they had revised their ASME BPVC, Section XI, Subsection IWE program to include visual examinations of the test connections in the leak -chase channel penetration pressurization system ( PPS) , when in fact, the program had not been revised and the examinations had not been performed . This information was material to the NRC because it was used, in part, as the basis for the approval and issuance of License Amendment 247, dated October 11, 2016, extending the TS containment leak rate test frequencies. The licensees corrective actions included conducting the visual examinations of the test connections in the leak -chase channel PPS during the ongoing refueling outage in March 2017 and initiating actions to add the visual examination requirements to their Subsection IWE program. This issue was entered into the licensees CAP as NCR 02110516. The failure to provide complete and accurate information in accordance with 10 CFR 50.9(a) for the LAR associated with License Amendment 247 is a violation of NRC requirements . This violation was screened against the ROP guidance in IMC 0612, Appendix B, Issue Screening, and no associated ROP finding was identified. The inspectors evaluated this issue using the Traditional Enforcement process because it had the potential to impact the NRCs ability to perform its regulatory function. Specifically, the violation impacted the regulatory process, in that the inaccurate information was material to the NRCs review and acceptance of licensee actions to address the industry -wide operating experience discussed in NRC IN 2014- 07. Based on licensee inaccurate information that they had addressed IN 2014 -07 by revising their containment ISI program to perform visual inspections of accessible tubing in the containment leak -chase channel PPS system, the NRC staff concluded that the licensee was properly implementing the ASME BPVC, Section XI, Subsection IWE program. In accordance with the guidance in Sections 2.2 and 6.9 of the NRC Enforcement Policy, the inspectors determined this is an SL IV violation, because had the information been complete and accurate at the time provided, it likely would have resulted in the need for further clarification of the licensees actions to address NRC IN 2014- 07 , but would not have caused the NRC to change its decision to issue the license amendment or resulted in substantial further inquiry . Also, on March 23, 2017, the licensee completed the visual examinations of the subject tubing in the leak -chase channel system and did not identify any significant degradation. In accordance with IMC 0612, Appendix B, traditional enforcement issues are not assigned a cross -cutting aspect.
05000261/FIN-2017405-0130 September 2016 23:59:59RobinsonNRC identifiedSecurity
05000261/FIN-2016008-0130 September 2016 23:59:59RobinsonNRC identifiedFailure to Failure to Keep EOP FRP-H.1 in Conformance with Plant Specific GuidelinesThe NRC identified a non-cited violation of Technical Specification 5.4.1 for the licensees failure to maintain emergency operating procedure (EOP) FRP-H.1, Response to Loss of Secondary Heat Sink, in accordance with their commitment to implement EOPs based on plant specific technical guidelines. Specifically, the licensee was committed to upgrading their EOPs in accordance with the H.B. Robinson Unit 2 plant specific technical guidelines, and FRP-H.1 was not updated during implementation of engineering change (EC) 283171. In response, the licensee entered the issue into their corrective action program as action request 2047575 and updated FRP-H.1 to bring it into conformance with its basis document. This performance deficiency was more than minor because it could lead to a more significant safety concern if left uncorrected. Specifically, the procedure would have been implemented as written during an event that required bleed and feed, and it was not demonstrated that one SI pump was adequate for core cooling. The finding required a detailed risk evaluation to be performed because the finding was not a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the finding would represent a loss of system and/or function, because it was not demonstrated that one safety injection (SI) pump would be sufficient during bleed and feed operations. A detailed risk assessment determined the increase in core damage frequency due to the performance deficiency was less than1E-6/year, a GREEN finding of very low safety significance. The team determined that the finding was indicative of current licensee performance, because the issue resulted from inadequate implementation of EC 283171, which was completed in 2014. A cross-cutting aspect of Teamwork (H.4.) in the Human Performance area was assigned because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained.
05000261/FIN-2016003-0230 September 2016 23:59:59RobinsonSelf-revealingFailure to Assess and Manage Risk for Main Turbine Trip Maintenance Resulting in Turbine/Reactor TripA self-revealing Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) was identified for the failure to adequately assess and manage the increase in risk associated with online maintenance activities involving the removal of the cover to the main turbine trip mechanism in order to perform visual inspections. During removal of the cover, the turbine trip mechanism lever was contacted causing an automatic turbine/reactor trip. The licensee took immediate corrective actions to reemphasize the need to enter all applicable types of work activities into the work management process and to conduct formal risk assessments in accordance with the risk management program. The licensee entered this issue into the corrective action program (CAP) as condition report (CR) 2056554. The licensees failure to adequately assess and manage the risk of maintenance associated with visual inspection of the turbine trip mechanism was a performance deficiency (PD). The inspectors evaluated the PD in accordance with IMC 0612, Appendix B, Issue Screening, and determined it to be more than minor because it impacted the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to assess and manage the risk associated with removing the turbine trip mechanism cover to conduct visual inspections resulted in a turbine/reactor trip. The inspectors evaluated the finding in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K, the inspectors requested that a regional Senior Reactor Analyst (SRA) independently evaluate the risk. A Region II SRA performed an analysis of the risk deficit for the unevaluated condition associated with the work activity on the turbine trip mechanism. The latest Robinson Standardized Plant Analysis Risk (SPAR) model was used to calculate an incremental core damage probability deficit (ICDPD). The result was an ICDPD of 3.74E-7 and represented the increase in core damage probability associated with a turbine/reactor trip coincident with the dedicated shutdown diesel generator being out of service at the time of the event. In accordance with IMC 0609, Appendix K, because the calculated ICDPD was not greater than 1E-6, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of work management in the cross-cutting area of human performance because the licensee failed to adequately implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. Specifically, the licensee failed to adequately assess, manage, and implement risk management actions for activities associated with trip sensitive equipment.
05000261/FIN-2016008-0230 September 2016 23:59:59RobinsonNRC identifiedFailure to Perform Appropriate Maintenance or Testing for the Dedicated Shutdown TransformerThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations (10 CFR) Part 50.63, Loss of all alternating current power, for the licensees failure to meet their commitment to the guidance in NRC RG 1.155, Station Blackout. Specifically, the licensees preventive maintenance and testing program did not identify required tests and inspections, and was not implemented such that it demonstrated system readiness and reliability requirements would be met as required by RG 1.155. In response, the licensee entered the issue into their corrective action program as action request 2053938, and initiated actions to determine which vendor recommended activities were needed to be performed to meet their RG 1.155 commitments and began updating their PM schedule and maintenance procedures. This performance deficiency was more than minor because it could lead to a more significant safety concern if left uncorrected. Specifically, transformer components degrade over time, and in the absence of appropriate testing and maintenance, could degrade to the point where the transformer may fail when called upon to mitigate an SBO. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The team determined that the finding was indicative of current licensee performance, because AR 643531 was written on November 11, 2013, which described that appropriate maintenance and testing was not being performed on the DS transformer, however, the impact on the stations RG 1.155 commitments was not evaluated. A cross-cutting aspect of Evaluation (P.2) in the Problem Identification and Resolution area was assigned because the licensee did not thoroughly evaluate the issue to ensure that the resolution addressed the cause and extent of condition commensurate with its safety significance.
05000261/FIN-2016008-0330 September 2016 23:59:59RobinsonNRC identifiedFailure to Comply with TS Requirements for Containment High Range Radiation MonitorsThe NRC identified a non-cited violation of Technical Specification (TS) 3.3.3, Post Accident Monitoring (PAM) Instrumentation, for the licensees failure to maintain the operability of the containment radiation monitors (high range) (CHRRMs). In response to this issue, the licensee generated AR 2062735 and made appropriate staff aware of the expected radiation monitor response and re-evaluated the IDO/PDO in NCR 2052758, and determined the CHRRMs were inoperable, and entered the appropriate TS action statement. This performance deficiency was determined to be 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 that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team determined the finding was of very low safety significance (Green) using the flowchart in IMC 0609, App. B, Attachment 2, because the finding was a failure to comply with a non-risk significant planning standard and no planning standard function failure occurred since other parameters could be used to validate the indications from the CHRRMs. This finding was not assigned a cross-cutting aspect because the issue was not indicative of current licensee performance. Specifically, the failure to properly evaluate the operability implications of IN 97-45 on the Robinsons CHRRMs occurred in 1997 and 1998.
05000261/FIN-2016008-0430 September 2016 23:59:59RobinsonNRC identifiedFailure to Follow Operability Determination ProcessThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow their operability determination procedure. Specifically, the licensee did not provide a high degree of assurance of operability in their immediate determination of operability (IDO) and did not perform a prompt determination of operability (PDO) as required when evaluating the operability of the containment radiation monitors (high range) (CHRRMs). In response to this issue, the licensee entered the issue into their corrective action program as AR 2055160, re-evaluated the IDO in NCR 2052758, and performed a detailed determination of operability in a PDO as required by their procedure. This performance deficiency 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 that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, an inadequate operability determination regarding the CHRRMs would adversely impact the licensees ability to classify, assess, and develop the correct protective measures following an accident. The team determined the finding was of very low safety significance (Green) using the flowchart in IMC 0609, App. B, Attachment 2, because the finding resulted in a failure to comply with a non-risk significant planning standard and no planning standard function failure occurred. Specifically, failure to follow the operability determination procedure and adequately determine the operability of the CHRRMs resulted in the failure to provide and maintain adequate emergency equipment that supports the emergency response, however, no failure of the planning standard occurred because other parameters could be used to validate the indications from the CHRRMs. The team determined that the finding was indicative of current licensee performance, because the issue resulted from inadequate implementation of the licensees operability determination process during the course of the inspection. A crosscutting aspect of Operating Experience (P.5.) in the Problem Identification and Resolution Area was assigned because the organization did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience (OE) in a timely manner.
05000261/FIN-2016404-0130 September 2016 23:59:59RobinsonNRC identifiedSecurity Safeguards Control
05000261/FIN-2016003-0130 September 2016 23:59:59RobinsonSelf-revealingFailure to Scope Tainter Gate Flood Protection Features in Maintenance Rule Resulting in Degraded PerformanceA self-revealing Green NCV of 10 CFR 50.65(b)(2)(ii) was identified for the failure to scope the external flood protection function of the Robinson Lake Dam spillway (Tainter) gates in the maintenance rule (MR) monitoring program. The failure to include the Tainter gates in the MR program resulted in ineffective maintenance being performed and subsequent degraded opening capability which challenged the availability of safety-related equipment during design basis rainfall events due to site flooding. The licensee took immediate corrective actions to replace/refurbish the chains to both gates and completed full open testing to restore their functionality. In addition, the licensee has developed and initiated implementation of an action plan to improve and ensure reliability of the gates, and initiated actions to revise the MR scoping program to include the Tainter gates. The issue was entered into the licensees CAP as CR 2035500. The failure to scope the flood protection function of the Lake Robinson Dam Tainter gates in the maintenance rule monitoring program was a PD. The finding is more than minor because it is associated with the protection against external factors (i.e., flood hazard) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to monitor flood protection features associated with the Tainter gates resulted in degraded gate opening performance that could have resulted in site flooding during design basis rainfall events and adversely impact multiple trains of safety-related equipment due to water intrusion. Using IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding involved the degradation of equipment specifically designed to mitigate flooding events. In accordance with Exhibit 4, External Events Screening Questions, the inspectors determined that the finding represented a degradation of two or more trains of a multi-train system or function during an external flooding event, therefore it required a detailed risk evaluation. A regional senior reactor analyst completed a detailed risk evaluation in accordance with NRC IMC 0609 Appendix A, and Appendix M, Significance Determination Process Using Qualitative Criteria, using the latest NRC Robinson Standardized Plant Analysis Risk model. The high uncertainty associated with estimating flood frequencies was the reason for using the NRC IMC Appendix M approach. The major analysis assumptions included a one-year exposure interval, recovery credit for opening the Tainter gates subsequent to binding of the chain, and limited credit for FLEX flooding mitigation strategies. If the rainfall produced a water surface elevation which would overtop the dam, the dam was considered failed and the ultimate heat sink lost. The rainfall frequencies requiring gate operation were estimated using a combination of National Oceanographic and Atmospheric Administration rainfall data and a probabilistic technique to establish precipitation frequency estimates performed by the licensee. The dominant sequence was a flood event inducing a non-recoverable loss of offsite power and loss of the emergency buses with a failure of the operators to manually recover the Tainter gates and failure of the operators to depressurize the steam generators to facilitate FLEX injection leading to a loss of core heat removal and core damage. The risk was mitigated by the low flood frequency, and the likely recovery of the Tainter gates prior to site flooding. There were additional conservatisms which were not applied to the result but would reduce the risk. These included the fact that the plant would be shutdown prior to flooding impacting safety-related equipment, which would reduce decay heat cooling required, and additional FLEX flooding strategies which could provide cooling even if the dam was lost. The risk increase due to the performance deficiency was < 1.0E-6/year, a Green finding of very low safety significance. The licensees analysis and full scope probabilistic risk assessment model produced a similar result. The inspectors determined that since the scoping of plant systems had occurred more than three years in the past, the finding did not represent current plant performance and therefore did not have a cross-cutting aspect associated with it.
05000261/FIN-2016002-0130 June 2016 23:59:59RobinsonLicensee-identifiedLicensee-Identified ViolationSection 50.55a(h)(2) of 10 CFR states in part, for nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of Institute of Electrical and Electronic Engineers (IEEE) Std. 6031991 and the correction sheet dated January 30, 1995. The Robinson FSAR (current licensing basis) Section 3.1.2.20, states in part that, reactor protection is designed to meet all presently defined reactor protection criteria and is in accordance with the proposed Institute of Electrical and Electronic Engineers (IEEE) 279 Standard for Nuclear Plant Protection Systems, August 1968. IEEE-279, Section 4.2, requires that any single failure within the protection system shall not prevent proper protection system action when required. Contrary to this requirement, from initial startup, until April 13, 2016, when using a FRBV (i.e., FRBV in the open position in Modes 1, 2, and 3), and a MSLB occurred, the protection system would not provide the proper system protection action. Specifically, with a single failure of the FRBV to close, the protective system action to isolate feedwater could not be accomplished. This would cause an increase in secondary mass available for release in containment structure, resulting in a higher peak containment pressure that would challenge the containment design pressure. As corrective actions, the licensee implemented a standing instruction and placed caution tags on the FRBVs to ensure the valves remain closed/isolated while operating in Modes 1, 2, and 3. Additionally, the licensee completed an engineering change to update the containment analysis and licensing basis. The licensee entered this issue into the CAP as CRs 2012658, 2020495, and 2018710. The failure to meet the single failure criterion for feedwater isolation following a main steam line break inside containment was a performance deficiency (PD). Significance Determination Process (SDP) screening in accordance with NRC IMC 0609.04 determined that the PD affected the secondary short term heat removal safety function of the mitigating systems cornerstone. The finding was determined to represent a loss of function and a detailed risk assessment was performed per NRC IMC 0609 Appendix A. The bounding analysis assumed a conditional core damage probability of 1.0, a 14 day exposure period estimated from surveillance and outage schedules, and main steam line break inside containment (MSLBIC) initiating event probability and main feedwater regulating valve bypass (MFWRVBV) failure to close probabilities from the NRC SPAR model data. The dominant sequence was an MSLBIC with a failure to close of the MFWRVBV which was assumed to lead to core damage and large early release. The risk was mitigated by short exposure period and the low likelihood of the MSLBIC and the failure to close of the MFWRVBV. The bounding analysis determined that the PD represented a risk increase of < 1.0E-7/year, a GREEN finding of very low safety significance for both core damage frequency and large early release frequency.
05000261/FIN-2016001-0231 March 2016 23:59:59RobinsonNRC identifiedFailure to Scope Safety-Related Auxiliary Building Ventilation Fans into the Maintenance RuleThe inspectors identified a Green NCV of 10 CFR 50.65(b)(1), for the licensees failure to include safety-related structures, systems and components (SSCs) within the scope of the maintenance rule program. Specifically, the licensee failed to include auxiliary building ventilation fans, which are required to remain functional during and following a design bases event to mitigate the consequences of an accident, within the scope of the maintenance rule monitoring program. The licensee initiated corrective actions to include the auxiliary building ventilation exhaust fans within the maintenance rule monitoring program. The licensee entered this issue into their CAP as CR 1997952. The failure to appropriately scope the safety-related auxiliary building ventilation fans within the maintenance rule was a performance deficiency. This performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to include auxiliary building ventilation fans in the maintenance rule affects the licensees ability to effectively monitor the performance or condition of the SSCs such that SSCs remain capable of fulfilling their intended function. Using IMC 0609, Appendix A, issued June 19, 2012, SDP for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The finding does not have a cross-cutting aspect since the failure to scope this equipment into the maintenance rule was not recognized during the initial maintenance rule scoping activities in 1997 and, as a result, is not indicative of current licensee performance.
05000261/FIN-2016001-0331 March 2016 23:59:59RobinsonNRC identifiedFailure to Follow Procedure for a Light Indication ReplacementThe inspectors identified a Green NCV of TS 5.4.1.a, for the licensees failure to adequately implement procedure OMM-001-11, Logkeeping, while performing maintenance. Specifically, the licensee replaced a local light indication for PCV-1716, a containment instrument air isolation valve, which resulted in a plant transient, without a senior reactor operator (SRO) being contacted, as required per procedure. As corrective action, the licensee replaced the blown fuse, issued a standing instruction to initiate a work request for all light bulb replacements, and submitted a procedure revision request to add more detailed guidance for lightbulb replacement. The licensee entered this issue into their CAP as CR 1991686. The failure to contact an SRO prior to changing out a local light indication for PCV-1716 as required by procedure OMM-001-11 was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, because the SRO was not contacted, an assessment and management of risk associated with the replacement of the light indication was not performed, and resulted in a plant transient. The inspectors evaluated the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Section B and determined the finding to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The performance deficiency had a cross-cutting aspect of Avoid Complacency in the area of Human Performance because the individual performing the lightbulb replacement did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk.
05000261/FIN-2016001-0431 March 2016 23:59:59RobinsonSelf-revealingFailure to Control Deviations from Design Specifications for the Service Water SystemA self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Specifically, the failure to control a modification of the service water (SW) system led to the installation of a non-conforming valve and resulted in the inoperability of the motor-driven auxiliary feedwater (MDAFW) system. As corrective action, the licensee performed a modification to replace the SW-115 valve. The licensee entered this issue into their CAP as CR 1993790. The failure to control deviations from design specifications for a modification to the SW system was the performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the installation of a valve outside of design specifications for the SW system contributed to the failure of SW-115 and reduction of cooling water flow to the MDAFW system. This degraded condition rendered the A train of the MDAFW system inoperable for greater than its TS AOT from December 15, 2015, to January 19, 2016. In addition, both trains of MDAFW were inoperable for greater than the TS AOT from January 19, 2016, to January 22, 2016. The inspectors used NRC IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to evaluate the significance of this issue and determined the finding required a Detailed Risk Evaluation because the finding represented an actual loss of function of at least a single train of MDAFW for greater than its TS AOT. A detailed risk evaluation was performed by a regional SRA in accordance with NRC IMC 0609 Appendix A using the NRC Robinson SPAR model and input from the licensees Robinson Fire PRA model. The major analysis assumptions included exposure periods for loss of a single train of MDAFW for 38 days and loss of both trains of MDAFW for a period of three hours. No recovery was assumed. The SDAFW and C AFW trains were not affected. The dominant sequence was a loss of main feedwater initiator followed by failures of both trains of MDAFW due to the performance deficiency, the SDAFW pump, the C train of AFW, and failure of the operator to implement feed and bleed cooling. The risk associated with this performance deficiency was mitigated by the availability of alternate AFW trains. The detailed risk evaluation determined that the increase in core damage frequency due to the performance deficiency was less than 1.0 E-6 per year and therefore the performance deficiency was characterized as a GREEN finding of very low safety significance. The finding does not have a cross-cutting aspect since the installation of a valve outside of design specifications into the SW system occurred prior to 1978 and, as a result, is not indicative of current licensee performance.
05000261/FIN-2016001-0531 March 2016 23:59:59RobinsonNRC identifiedFailure to Adequately Maintain Emergency Operating Procedures EOP-ECA-0.0 and EOP-E-0The NRC identified a Green NCV of TS 5.4.1.a for the licensees failure to adequately maintain procedures EOP-ECA-0.0, Loss of All AC Power, and EOP-E-0, Reactor Trip or Safety Injection, as recommended in Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operations), Revision 2, dated February 1978. Specifically, both procedures contained inadequate procedure steps. Revision four of EOP-ECA-0.0, contained a step that could delay or prevent the restoration of a charging pump when electrical power was available to do so. This could have led to a loss of reactor coolant system (RCS) pressure control. Revision six of EOP-E-0, contained a step that could have led to the restoration of seal injection to overheated reactor coolant pump (RCP) seals with subsequent RCP seal damage and RCS leakage. The licensee submitted procedure revision requests (PRRs) 2009136 and 2009217 to correct the procedures. The licensee entered this issue into their CAP as CR 2009602. The licensees failure to adequately maintain the emergency operating procedures by having inadequate procedure steps, was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, steps in EOP-ECA-0.0 and EOP-E-0 could lead to one or more of the following during an event: unnecessary reduction in core sub-cooling margin, loss of RCS pressure control, RCP seal damage, and/or excessive RCS leakage. The finding is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of a safety system and/or function; (3) did not represent an actual loss of function of at least a single train of a plant system for longer than its TS allowed outage time, or two separate safety systems out-of-service for longer than their TS allowed outage time; and (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program for greater than 24 hours. The finding has a cross-cutting aspect in the area of human performance associated with avoiding complacency because individuals did not recognize and plan for the possibility of mistakes and latent issues when performing EOP verification and validation.
05000261/FIN-2016001-0131 March 2016 23:59:59RobinsonNRC identifiedFailure to Adequately Establish and Implement Procedure During Tornado Watch/WarningThe inspectors identified a Green non-cited violation (NCV) of technical specification (TS) 5.4.1.a for the licensees failure to adequately establish and implement procedure OMM-021, Operation During Adverse Weather Conditions. Specifically, the licensee failed to include requirements to tie down or remove loose material in the area of Unit 1 adjacent to the switchyard. Additionally, the licensee failed to implement the procedural requirements to tie down or remove material in the vicinity of the turbine building ground level and secure doors to the chemical treatment room and as required by procedure OMM-021. As corrective action, the licensee secured or removed the material in the vicinity of the turbine building and issued a procedure change request to change procedure OMM-021 to include an action to secure or remove potential missile hazards in the vicinity of the switchyard in the Unit 1 area. The licensee entered this issue into their corrective action program (CAP) as condition report (CR) 2005141. The licensees failure to include requirements to tie down or remove loose material in the area of Unit 1 adjacent to the switchyard in procedure OMM-021, and failure to implement the procedural requirements to tie down or remove material in the vicinity of the turbine building ground level and secure doors to the chemical treatment room as required by procedure OMM-021 during a tornado watch/warning was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to secure or remove potential missile hazards in the areas adjacent to the switchyard increased the likelihood of a unit trip and/or loss of offsite power event. The inspectors evaluated the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Section B and determined the finding to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The performance deficiency had a cross-cutting aspect of Work Management in the area of Human Performance because the organization did not implement a process of planning, controlling, and executing work activities associated with a tornado watch/warning such that nuclear safety was the overriding priority.
05000261/FIN-2015405-0530 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2015405-0130 September 2015 23:59:59RobinsonNRC identifiedSecurity
05000261/FIN-2015405-0630 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2015003-0230 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation10 CFR 50.49, Environmental Qualification of electric equipment important to safety for nuclear power plants, states that each licensee shall establish a program for qualifying specified electric equipment. Section (e) (5) of 10 CFR 50.49 specifies, in part, that the licensee must replace or refurbish equipment at the end of its qualified life unless ongoing qualification demonstrates that the item has additional life. Contrary to the above, on February 18, 2015, the licensee determined that they failed to replace or refurbish the pressurizer PORV limit switches prior to the end of their qualified life, or by February 23, 2007. The licensee was also unable to demonstrate the limit switches had additional life. The licensee documented this condition in CR 738953. Following discovery of this condition, the licensee implemented an interim standing instruction for alternate methods of monitoring PORV positon, and finally replaced the switches during refueling outage 29 which began May 12, 2015. Using IMC 0609, Appendix A, SDP for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the PORV maintained its operability although the finding affected its qualification.
05000261/FIN-2015405-0230 September 2015 23:59:59RobinsonNRC identifiedSecurity
05000261/FIN-2015405-0730 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2015003-0130 September 2015 23:59:59RobinsonNRC identifiedFailure of Refueling Water Storage Tank (RWST) Discharge Valve to Close on DemandThe inspectors identified an unresolved item associated with the failure of SI-864A, Reactor Water Storage Tank discharge valve to close on demand during surveillance testing. The URI is being opened to review the licensees cause evaluation and determine if a performance deficiency exist. On 5/18/2015, with the plant in Mode 6, the A RWST discharge valve, SI864A failed to stroke close on demand, from the control board, during engineering surveillance testing. Troubleshooting revealed that the thermal overload relay within the breaker that supplies power to the MOV actuator was tripped, rendering the A emergency core cooling system train (ECCS) inoperable. Following discovery of this issue, maintenance personnel manually reset the thermal overload and cycled the valve closed. Additionally, the licensee replaced the thermal overload relay and performed post maintenance testing. The licensee documented this issue in CR 749789 and initiated a cause evaluation. The exact time and cause of the tripped thermal overload relay for SI-864A is unknown at this time. Additional inspection time is required to review the licensees evaluation and determine if a performance deficiency exist. This issue will be identified as URI 05000261/2015003-01, Failure of Refueling Water Storage Tank (RWST) Discharge.
05000261/FIN-2015405-0330 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2015406-0130 September 2015 23:59:59RobinsonNRC identifiedSecurity
05000261/FIN-2015404-0130 September 2015 23:59:59RobinsonNRC identifiedSecurity
05000261/FIN-2015405-0430 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2015404-0230 September 2015 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation
05000261/FIN-2015002-0130 June 2015 23:59:59RobinsonNRC identifiedFailure to Timely Report Required Information as Required by 10 CFR 50.73An NRC-identified Severity Level IV NCV of 10 CFR 50.73, Licensee Event Report System, was identified for the licensees failure to submit a licensee event report (LER) within 60 days after discovery of a condition which was prohibited by the plants Technical Specifications (TS). The issue was entered into the licensees CAP as condition report (CR) 743653. The licensee submitted the LER to restore compliance. The licensees failure perform an adequate reportability evaluation and subsequently submit an LER within 60 days after discovery of a condition which was prohibited by the plant TSs as required by 10 CFR 50.73 was a performance deficiency. This performance deficiency was assessed using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors determined the significance of this violation was a Severity Level IV NCV using Section 6.9.d.9 of the NRCs Enforcement Policy. Cross cutting aspects are not assigned to traditional enforcement violations.
05000261/FIN-2015002-0230 June 2015 23:59:59RobinsonSelf-revealingFailure to Follow Engineering Change Procedure for Modification of RPSA self-revealing Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to follow EGR-NGGC-0005, Engineering Change, during a modification of the reactor protection system (RPS). This resulted in having inadequate work instructions associated with engineering change (EC) 75690 and EC 86690, which resulted in a cross-tied configuration of independent trains of the RPS and the DC electrical system. The licensee entered this into the corrective action program (CAP) as action request (AR) 729926 and took immediate corrective actions to cut the cable and restore the independence of safety trains for both systems. The failure to have adequate work instructions for engineering changes as required by procedure EGR-NGGC-0005 was a performance deficiency. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the cross-tied configuration rendered the RPS and DC electrical subsystem inoperable because the required independence and redundancy of systems were eliminated. The finding was screened using IMC 0609 Appendix A Exhibit 2.C, Reactivity Control Systems, dated June 19, 2012, and was determined to be of very low safety significance (Green) because the finding did not result in a mismanagement of reactivity by the operators. The performance deficiency had a cross-cutting aspect of teamwork in the area of human performance because the licensee failed to coordinate their activities between the work control planners and engineering to ensure nuclear safety was maintained.
05000261/FIN-2015001-0131 March 2015 23:59:59RobinsonNRC identifiedInadequate 10 CFR 50.59 Evaluation Results in RPI System InoperabilityThe inspectors identified a severity level IV (SLIV) non-cited violation (NCV) of 10 CFR 50.59, Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment prior to implementing a change to licensee procedure OST-20, Shiftly Surveillances. Specifically, a note was added to procedure OST-20 to allow the use of the Emergency Response Facility Information System (ERFIS) as an acceptable alternate method to determine Analog Rod Positon Indication (ARPI) System operability if the position indicators were not indicating properly. This change resulted in an associated Green NCV of Technical Specification (TS) 3.1.7, Rod Position Indication, for failing to shut down the reactor or follow remedial actions permitted by a TS action requirement when a Limiting Condition for Operation (LCO) was not met. Upon determination that the practice of crediting ERFIS for rod position indication (RPI) operability was not allowed by the current licensing basis (CLB), Standing Instruction 14-023 was issued to suspend the practice and condition report (CR) 720726 was written to document the issue. The licensees failure to obtain a license amendment for a change that resulted in a change to technical specifications incorporated in the license was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the mitigating systems cornerstone attribute of procedure quality and adversely affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure change adversely impacted the availability and capability of systems to respond to a design basis event because it allowed the use of a non CLB method for determining rod position after failure of the ARPI system. Rod position indication is required to determine maximum rod misalignment which is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available shutdown margin. The finding was screened using IMC 0609 Appendix A Exhibit 2.C, Reactivity Control Systems, dated June 19, 2012, and was determined to be of very low safety significance (Green) because the finding did not result in a mismanagement of reactivity by operators. The violation was determined to be a SLIV violation using the Enforcement Policy example 6.1.d.2, because it resulted in a condition having very low safety significance. No cross-cutting aspect was assigned in association with the ROP finding because the change to the procedure was performed greater than three years ago and did not reflect current licensee performance.
05000261/FIN-2014201-0131 December 2014 23:59:59RobinsonNRC identifiedSecurity
05000261/FIN-2014005-0131 December 2014 23:59:59RobinsonNRC identifiedFailure to Protect Diesel Driven Equipment from Effects of Extreme Cold TemperaturesThe inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1, for failure to establish procedural guidance to protect diesel driven equipment important to safety from the effects of extreme cold temperatures. Specifically, the licensees cold weather procedures failed to include actions to maintain fuel oil temperatures above the diesel fuel oil cloud point for the dedicated shutdown diesel generator (DSDG) and/or the engine driven fire pump (EDFP). The licensee entered this into the corrective action program (CAP) as AR 715032 and took immediate corrective actions to revise station procedures to protect the diesel driven equipment during periods of extreme low temperatures. The failure to establish procedural guidance to protect diesel-driven equipment important to safety from the effects of extreme cold temperatures was a performance deficiency. This issue was more than minor because if left uncorrected this finding would have the potential to lead to a more significant safety concern. Specifically, failure to maintain the fuel oil temperatures for the DSDG and/or the EDFP greater than the measured cloud point, may impact the operation of the equipment during extreme low temperature conditions, due to the associated fuel oil transfer system becoming non-functional. A detailed risk assessment was performed by a regional Senior Reactor Analyst in accordance with NRC IMC 0609 Appendices A and F. The latest NRC Robinson SPAR risk model was used to quantify the internal events risk and a calculation was performed to estimate the fire risk. The major analysis assumptions included: both the EDFP and the DSDG were simultaneously considered unavailable without recovery for a 1-day exposure interval, DSDG fire scenarios were considered for the emergency switchgear room (ESWGR), the cable spreading room, and the main control room, where fire could cause a loss of offsite power and the emergency diesel generators (EDGs), compartment total ignition frequency data from the Robinson NFPA 805 project was used and a bounding Conditional Core Damage Probability for the fire scenarios of 1.0. The dominant sequence was a fire in the ESWGR which remained unsuppressed long enough to cause a loss of offsite power and the EDGs requiring use of alternate shutdown which failed due to the performance deficiency impact on the DSDG resulting in station blackout, and core damage due to an unmitigated reactor coolant pump seal loss of cooling accident. The risk was mitigated by the low likelihood of the initiators occurring during the specific cold weather vulnerability periods. The risk due to the performance deficiency was determined to be an increase in core damage frequency of <1E-6/year, a GREEN finding of very low safety significance. The performance deficiency had a cross-cutting aspect of Evaluation in the area of Problem Identification and Resolution because the licensee failed to thoroughly evaluate the effects of cold weather on the fuel system for diesel driven equipment to ensure that resolutions address the extent of conditions commensurate with their safety significance (P.2).
05000261/FIN-2014004-0130 September 2014 23:59:59RobinsonLicensee-identifiedLicensee-Identified Violation10 CFR 26.205(d)(7), Work Hour Controls, requires, in part that the licensee shall control the work hours of individuals to less than a weekly average of 54 hours, calculated using an average period of six weeks. Contrary to this requirement, seven covered workers violated the work hour limits on thirteen occasions. Specifically, following the Spring 2014 forced outage, the licensee failed to recode all covered workers to an on-line status. This finding was more than minor because if left uncorrected, the failure to change the individual coding from off-line to on-line status would have allowed all covered workers onsite to exceed work hour limits, and could lead to a more significant safety concern. This violation was determined to be of very low safety significance because no significant events or human performance issues were directly linked to personnel fatigue as a result of the hours worked. The licensee entered this issue into their CAP as CR 698782.
05000261/FIN-2014003-0230 June 2014 23:59:59RobinsonLicensee-identifiedLicensee-Identified ViolationSection 50.49 of 10 CFR, Environmental Qualification of electric equipment important to safety for nuclear power plants, states that each licensee shall establish a program for qualifying specified electric equipment. Section (a)(3) of 10 CFR 50.49 specifies the environmental qualification requirements for post-accident monitoring equipment. Section (f) of 10 CFR 50.49 requires, in part, that each item of electric equipment important to safety must be qualified by testing an identical item of equipment under identical conditions. Contrary to the above, since May 1992, the licensee failed to maintain the qualification of the limit switches for CVC-204B, letdown line isolation, in accordance with the tested configuration of the equipment which rendered the Post Accident Monitoring Instrumentation function inoperable. The licensee documented this condition in AR 640902 and AR 633207. The cause was determined to be associated with a human performance event in which the licensee failed to use the proper heat shrink insulators per procedure CM-309, Sealing Low Voltage Electrical Splices for Environmentally Qualified or Safety Related Splices. Following discovery of this condition, the licensee replaced the non-environmental qualified splice and returned the equipment to the test configuration. Using IMC 0609, Appendix A, issued June 19, 2012, The SDP for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours.
05000261/FIN-2014008-0130 June 2014 23:59:59RobinsonNRC identifiedFailure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with the Steam Generator Tube LeakThe team identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take adequate corrective action to prevent repetition of a significant condition adverse to quality regarding steam generator tube leakage due to poor maintenance practices. Specifically, on February 27, 2014, the C steam generator showed indications of a primary to secondary tube leak due to foreign material that was introduced during the fall 2013 refueling outage. As immediate corrective actions, on March 7, 2014, the licensee shutdown the plant and repaired the leak. This violation was entered into the licensees CAP as nuclear condition reports (NCRs) 683695, 683593, and 683591. The licensees failure to implement appropriate corrective actions to address poor worker practices to prevent recurrence of a steam generator tube leak was a performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone equipment performance attribute and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, foreign material entered the steam generator and damaged a steam generator tube, which increased the likelihood of a steam generator tube rupture. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At-Power, dated June 19, 2012. The finding screened as Green per Section D of Exhibit 1, Initiating Events Screening Questions, because testing showed that the affected steam generator tube could sustain three times the differential pressure across the tube during normal full power and that the steam generator did not violate the accident leakage performance criterion. The performance deficiency does not have a cross cutting aspect because the last revision of the root cause evaluation was completed in 2011 and it is not indicative of current licensee performance.
05000261/FIN-2014003-0130 June 2014 23:59:59RobinsonSelf-revealingFailure to Identify and Correct Degraded Wire Labels in the Reactor Protection Relay CabinetsA self-revealing Green non-cited violation (NCV) was identified for the licensees failure to promptly identify and correct degraded wire labels in the reactor protection cabinets, which were a condition adverse to quality, as required by 10 CFR Part 50, Criterion XVI, Corrective Action. This resulted in an automatic reactor trip. Immediate corrective actions included inspection of both trains of relay racks to identify and remove any potential foreign material. The licensee also tested both trains of reactor protection relays to verify no foreign material was present. Additionally, the licensee plans to replace the wire labels in the reactor protection and safeguards relay racks during refueling outages 29 and 30. The licensee documented the issue in the corrective action program as CR 654789. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the degraded wire labels became lodged between contact 2-6 on relay LC-496A1-X(B), which set up the half-trip condition to cause a reactor trip, during the surveillance testing. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because although the finding caused a reactor trip, it did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a crosscutting aspect of identification in the area of problem identification and resolution because the licensee failed to implement a corrective action program with a low enough threshold for identifying issues in that the licensee process did not recognize, during review of the work requests for the degraded wire labels, that this issue should have been entered into the corrective action program as a nuclear condition report.
05000261/FIN-2014002-0331 March 2014 23:59:59RobinsonSelf-revealingInadequate Preventive Maintenance on 4 KV Breaker 52/7 Results in an Automatic Reactor TripA self-revealing Green finding (FIN) was identified for the licensees failure to perform adequate preventive maintenance (PM) in accordance with, licensee procedure ADM-NGGC-107, Equipment Reliability Process, for 4 KV Breaker 52/7, Unit Auxiliary to 4 KV Bus 1. As a result, while transferring loads from the start-up transformer, a broken operating rod for breaker 52/7 prevented the breaker from closing and caused an automatic reactor trip. The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance, and it adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in breaker 52/7 failing to close and subsequently causing an automatic reactor trip from 19 percent power operations on November 5, 2013. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would not be available. The performance deficiency had a crosscutting aspect of Resolution in the area of Problem Identification and Resolution, because the licensee failed to take effective corrective actions to address a similar failure of an operating rod for the A circulating water (CW) pump breaker in 2011.
05000261/FIN-2014002-0531 March 2014 23:59:59RobinsonNRC identifiedDefective Motor Operated Potentiometer causes failure of the DSDG during surveillance testingAn URI was identified regarding the trip of the DSDG, on December 31, 2013, during monthly surveillance testing. The URI is being opened to provide for additional inspection of the equipment issues that led to the failure and to review the results of the vendors analysis of a defective motor operated potentiometer to determine if a performance deficiency exists. On December 31, 2013, during monthly testing of the DSDG in accordance with licensee procedure OST-910, Dedicated Shutdown Diesel Generator (Monthly), the output breaker tripped open on overcurrent while the operators were attempting to adjust DSDG output voltage. Operators in the field noted erratic voltage indication prior to the failure. Engineering identified that the likely cause was a failure of the motor operated potentiometer (MOP). The licensee replaced the MOP with a new part from stock and performed post maintenance testing. The MOP that was removed was sent offsite for forensic analysis. During examination, the licensee identified a manufacturing defect for the MOP. The licensees extent of condition investigation found the same manufacturing defect on the MOP installed in the DSDG and in a MOP in storage. The licensee replaced the MOP in the DSDG with a MOP that was verified to be acceptable. Engineering has sent the defective components back to the vendor for additional analysis. This issue will be identified as URI 05000261/2014002-05; Defective Motor Operated Potentiometer causes failure of the DSDG during surveillance testing.
05000261/FIN-2014002-0131 March 2014 23:59:59RobinsonNRC identifiedFailure to Adequately Critique Fire Brigade DrillsA Green NRC-Identified non-cited violation (NCV) of Facility Operating License DPR-23, Condition 3.E, Fire Protection Program, was identified for the licensees failure to identify, critique, and develop corrective actions for fire brigade performance weaknesses during two fire drills as required by procedure TPP-219, Fire Protection Training Program. Upon identification of these weaknesses by the inspectors, the licensee entered them into the corrective action program (CAP), performed an apparent cause evaluation, and revised procedure TPP-219 to further define the roles and responsibilities of the drill controllers as well as the standards used to critique the fire brigade. The licensees failure to identify, critique, and develop appropriate actions for fire brigade performance weaknesses during two fire drills as required by procedure TPP-219 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety Significance (Green) in accordance with question D.1 because although the finding involved fire brigade training requirements, the fire brigade demonstrated the ability to meet the required times for fire extinguishment for the fire drill scenarios and the finding did not significantly affect the fire brigades ability to respond to a fire. The performance deficiency had a cross-cutting aspect of Consistent Process in the area of Human Performance, because the licensee failed to use a consistent, systematic approach during conduct of fire brigade drills and during the subsequent critique process.
05000261/FIN-2014002-0431 March 2014 23:59:59RobinsonNRC identifiedFailure to Provide Adequate Design Control Measures for Diesel Fuel Oil Cloud PointThe inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to provide adequate design control measures to ensure appropriate specifications were translated into procedures for diesel fuel oil (DFO) to ensure that the DFO temperatures remained above the DFO cloud point. The licensee entered this into the CAP as action request (AR) 664223 and took immediate corrective actions to change the cloud point acceptance criteria from 23 degrees to 10 degrees Fahrenheit and revise procedure OP-925, Cold Weather, to install temporary heaters if outside temperatures fell below 15 degrees Fahrenheit. The licensees failure to provide design control measures to ensure that the DFO temperature was maintained such that the cloud point was not reached was a performance deficiency. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, during periods of cold weather the DFO temperature could have been allowed to fall below its cloud point and affect operation of the emergency diesel generator (EDG) and/or the dedicated shutdown diesel generator operation due to the DFO transfer system becoming inoperable. The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, dated June 19, 2012, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding is a deficiency affecting the design or qualification of a mitigating SSC; however, the SSC maintained its operability or functionality since the design conditions were not actually reached. The performance deficiency had a cross-cutting aspect of Design Margins in the area of Human Performance because the licensee failed to recognize that additional actions were required to maintain operability of the DFO system when ambient temperatures are below the maximum administrative limit even though samples are reviewed monthly per the DFO Testing Program.