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	Quarter	Site	Title	Description
05000327/FIN-2017008-06	2017Q4	Sequoyah	Potential Unjustified Qualified Life for ASCO Solenoid Operated Valves	Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for changing the activation energy and calculating a new qualified life for ASCO NP-1 valves assemblies. Description: The manufacturer, ASCO, conservatively established a 1.0 eV activation energy for the valve coil assemblies. The activation energy appeared to be determined by test and realistic coil failure modes. The conservative methodology used by ASCO, that used the most limiting activation energy, met the requirements in 10 CFR Part 50. By memorandum dated 8/19/2004, the nuclear utility user group for environmental qualification (NUGEQ), to which the licensee was a member, provided information supporting the use of revised activation energy values from 1.0 eV to a less limiting 1.37 eV. The memorandum (memo) specified that NUGEQ was tasked to revise the activation energy values for ASCO NP series SOVs to a less limiting one. The inspectors determined that the data and conclusions reported by NUGEQ did not appear to be justified by design control measures in accordance with 10 CFR Part 50 Appendix B Criterion III and 50.49. Adequate design control measures were specified in the Category 1 specifications established in NUREG-0588 Section 4, Aging and IEEE 323-1974 Section 6.3.3, Aging, as supplemented by RG 1.89 revision 1, Regulatory Position 5 Aging. The NUGEQ memo specified that they obtained their data through the research of information acquired from various sources. The use of the 1.37 eV value was for significantly increasing the qualified life of the ASCO coils. The inspectors are concerned that this did not meet the requirement to prove conservative extrapolations and use of the most limiting activation energies. Based on the inspectors review, NUGEQ did not demonstrate that the more limiting activation energies were unrealistic and could be discounted. The memo specified that the information NUGEQ used to derive 1.37 eV was based on emailed recollections of past DuPont testing. The DuPont email appeared to be supported by some identifiable test data, but was not quality related, was not commercial grade dedicated, and performed without any identifiable design control measures. In addition, the memo disregarded other coil components with more limiting activation energies by discounting the failure modes associated with them and the coil. The manufacturer ASCO found these discounted failure modes relevant to the coil safety functions. The inspectors are concerned that the licensee disregarded realistic, more limiting, failure modes without proper justification. The design control requirements in NUREG-0588 Section 4(5) specified, in part, that known material phase changes and reactions should be defined to insure that no known changes occur within the extrapolation limits, (staff position: claims that conservative extrapolation limits have been implemented must be supported), and Section 4(6) required, the aging acceleration rate used during qualification testing and the basis upon which the rate was established should be described and justified, (staff position: testing of the equipment should be conducted using the most limiting (lowest) activation energy of the components). Additionally, RG 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Revision 1, Regulatory Position 5.c, Section 6.3.3, Aging, of IEEE Std. 323-1974, specified, in part, that the aging acceleration rate and the basis upon which it was established be described, documented, and justified. The licensee has captured these concerns in their corrective action program as CR 1366020. The inspectors need to review the licensees analysis and justification for discounting realistic failure modes, changing the activation energy, and calculating a new qualified life for ASCO NP-1 valves assemblies. This URI is opened to determine if the performance deficiency for not providing adequate justification for changing the activation energy, is more than minor. (URI 05000327/2017008-06, 05000328/2017008-06, Potential Unjustified Qualified Life for ASCO Solenoid Operated Valves)
05000327/FIN-2017008-05	2017Q4	Sequoyah	Potential Inadequate Justification for Eliminating Preventative Maintenance for ASCO Valves	Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for eliminating the replacement of components that have a shorter life than the qualified life of the ASCO NP-1 valves assemblies. Description: The inspectors reviewed records in environmental qualification data package (EQDP), SQNEQ-SOL-005, Revision 47. After the valve manufacturer (ASCO) stopped providing rebuild kits, the licensee eliminated the replacement schedule for subcomponents that had a shorter life than the valve assembly. The licensee changed the inputs to the accelerated aging calculation and recalculated the life of these subcomponents from the approximate eight-year replacement schedule to approximately 32.5 years. The licensee changed the activation energies from 0.94eV and 0.96eV for ethylene-propylene-diene-monomer (EPDM) and Viton-A, respectively, to 1.1eV for both. Both, EPDM and Viton-A are rubber elastomers used within the ASCO valve assemblies. The licensees written justification in the EQDP referenced a review of several studies for each elastomers, which identified less limiting activation energies than the activation energies ASCO selected in their qualification test reports. Each of these studies used different material degradation mechanisms and end of life failure mechanisms to derive different activation energies. The conclusion in the EQDP, for the change justification analysis, stated in part, that these studies show that Sequoyah's original values were, in many cases, very overly conservative. The inspectors identified that the qualification of record, report AQR-67368, selected the qualification testing criteria based on the maintenance requirements (replacement schedule) specified in Appendix C of the report. The activation energies determined applicable in the ASCO test reports, (EPDM 0.94 eV and for Viton 0.96 eV) were determined by material testing and do not appear to inspectors to be very overly conservative (unrealistically low) or lacking in technical merit. In addition, the licensee did not consider the effect the various formulations for EPDM and Viton-A elastomers. The different formulations could non-conservatively affect the activation energies reviewed in their justification. The licensee subsequently replaced the accelerated aging rate used in ASCO qualification test report AQR-67368 with the more thermally severe acceleration rate used in AQS-21678/TR, ASCO Qualification Test Report, dated 7/1/1979. AQS-21678/TR does not appear to meet Category 1 requirements, yet its accelerated aging rate was used to replace the Category 1 qualification-aging rate in AQR-67368. The AQS-21678/TR report, specified, in part, that coils and elastomeric components shall be replaced every 4 years as noted in the Valve Design Specification Sheets. The thermal 12 aging in AQR-67368 simulated a minimum of 2,000 cycles (~4.5 cycles/hr.), which was more limiting than the once every 6 hours (96 cycles) specified in AQS-21678/TR. The inspector noted that the test program specified in AQS-21678/TR used IEEE 382-1972, which did not meet Category 1 qualification requirements. The test program in AQR-67368 used IEEE 382-1980, which did meet Category 1 requirements. The forward to IEEE 382-1980, stated, in part, that the testing in the report satisfies the latest issued requirements and standards, which were the NUREG-0588 Category 1 requirements issued for comment December 1979 and published in July 1981. The licensee also used a lower self-heating temperature than which was specified in ASCO letter ASCO Solenoid Valve Coil Heat Rise Data, dated 5/8/1986. The licensee determined that the ASCO heat-rise data in the above document was too conservative and used heat-rise data from the first testing done by the Franklin Research Center (FRC). Later, FRC completed NUREG/CR-5141 RV, Aging and Qualification Research on Solenoid Operated Valves, dated 4/1/1988, which was referenced in the EQDP. The NUREG specified that: Aging in forced air ovens significantly limited heat rise from self-heating. The NUREG further specified that the qualified lives of the subcomponents were significantly reduced after accounting for the differences between forced air vs less turbulent air flow and the actual temperature measurements made by ASCO vs approximate temperature measurements made by other testing. Additional difference between ASCO heat rise testing and other testing including FRCs was that ASCO drilled holes in the valves to measure the actual subcomponent temperature while others only measured externally near the subcomponents to avoid damaging the valves. This produced lower temperature readings than ASCOs. The inspectors noted that even applying the 1.1 eV currently used, the life of the elastomers appeared to be approximately 4 to 8 years, not the 32.5 years identified in the EQDP The ASCO qualification testing used nitrogen during the qualification testing. The NUREG/CR 5141 RV specified, that oxygen exposure from plant compressed air systems produced more degradation than did the nitrogen used in the ASCO qualification testing. The licensee determined that radiation margin from the qualification tests could be used to mitigate the differences between these two gases. The inspectors determined that both ASCO qualification reports, AQS-21678/TR and AQR-67368, used nitrogen instead of air, which limited the aging degradation. The inspectors question whether radiation margin can be applied to account for the difference between oxidizing gases and inert gases. In addition, the ASCO report AQR-67368, specified, that Viton elastomers significantly degraded above 18E6 test dose not 200E6 test dose used for the margin. The inspectors are concerned that the licensee failed to meet the Category 1 requirements as specified in NUREG-0588 and IEEE 323-1974. Category 1 specified proof of conservative extrapolations, and use of the most limiting activation energy. Additionally, that users of IEEE 382-1972 must meet Category 1 requirements. Furthermore, the inspectors noted that the licensee was made aware of similar deficiencies and NRC staff positions in Technical Evaluation Report (TER)-C5257-532, Implementation Guidance for New and Corrective Equipment Environmental Qualification, dated 4/22/1983. The licensee has captured these concerns in their corrective action program as CR 1366024. The inspectors need to evaluate: (1) the licensees justification for changing the activation energy; (2) the licensees assessment of how AQS-21678/TR met Category 1 requirements; (3) the adequacy of the licensees heat rise data; and (4) the licensees evaluation of elastomer degradation from oxygen vs nitrogen gas, and their use of apparent radiation dose margin to account for these differences. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-05, 05000328/2017008-05, Potential Inadequate Justification for Eliminating Preventative Maintenance for ASCO Valves)
05000327/FIN-2017008-04	2017Q4	Sequoyah	Potential Inadequate Determination of Failure Modes for Qualified Life for Foxboro/Weed Instrument	Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for failure modes and the degradation leading to them in the determination of the qualified life for the Foxboro/Weed Instrument transmitters documented in the licensees EQ Binder IPT-002. Description: The inspectors reviewed EQ Binder IPT-002 for Foxboro/Weed Instrument flow transmitter qualification. In reviewing qualification test report QOAACIO, Rev. A, the inspectors identified two concerns with the qualification. a. The inspectors noted that the qualification test report identified that polysulfone had the most limiting activation energy, 0.72 eV, for the Weed instrument assembly. However, the 0.72 eV was not being used by the licensee to determine the qualified life in the Arrhenius calculations in accordance with NUREG-0588 Section 4(6) and RG 1.89, Rev. 1, C.5.c. When inspectors questioned the licensees use of a less limiting activation energy (0.78 eV for resistors), the licensee determined that the 0.72 eV was based on the degradation and failure modes associated with the tensile strength material property for polysulfone. The licensee consulted the component manufacturer, and determined that creep was the correct material property to be evaluated for end of life. The activation energy for polysulfone creep was identified as 3.81 eV. The inspectors have challenged the licensees determination that creep is the only material property that can produce a failure of the sealing function for polysulfone. The inspectors noted that the requirements in IEEE Std. 323-1974, Section 5 require, in part, that assurance be provided that any extrapolation or inference be justified by allowances for known potential failure modes (i.e. loss of sealing function) and the mechanisms leading to them (i.e. the degradation in various material properties). If the degradation associated with creep is not the only degradation mechanism that could lead to a loss of sealing function over time, inspectors question if the degradation of other material properties would have more limiting activation energies than the licensees current activation energy for the Foxboro/Weed Instrument transmitter (0.78 eV). b. The 0.78 eV activation energy used by the licensee for qualified life was derived from an academic white paper that documented experiments performed in the early space program. The white paper specified that its experimental methods were not validated. The vendor that qualified the transmitter subsequently used this experimental information to determine the qualified life of the transmitters. This activation energy appeared not to be valid in the range of service temperatures that the transmitters are expected to age in prior to a DBA. The inspectors identified that these experimental tests did not follow any identifiable quality standard. The tests were conducted as early as 1963, the white paper was published in 1968, and the inspectors could not identify any subsequent verification of these experimental methods. Although no failure modes and effects analysis was evident, the table of 11 components in the qualification appeared to identify other components that could have much more limiting activation energies that were identified by qualification, as low as 0.5 eV. The licensee has captured these concerns in their corrective action program as CR1366039, CR 1363427. The inspectors need further information from the licensee and NRC technical staff to evaluate the concerns. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-04, 05000328/2017008-04, Potential Inadequate Determination of Failure Modes for Qualified Life for Foxboro/Weed Instrument)
05000327/FIN-2017008-03	2017Q4	Sequoyah	Potential Inadequate Use of Thermal Aging and the Arrhenius Methodology	Introduction: The inspectors identified a URI for the licensees use of the Arrhenius methodology without consideration for the limits of extrapolation and confidence bounds for statistical uncertainties. Description: The licensee did not consider the limits of extrapolation specified for Category 1 qualification in NUREG 0588 Section 4 and IEEE 323-1974 Section 6.5, Determination of Qualification. NUREG-0588, Section 4(5) required, in part, that known material phase changes and reactions should be defined to insure that no known changes occur within the extrapolation limits, (staff position: claims that conservative extrapolation limits have been implemented must be supported). Standard IEEE 323-1974, Section 6.5, specified, in part, that the qualified life shall be based upon the known limits of extrapolation of the time dependent environmental effects if an accelerated aging test was used to determine the mathematical model. Ancillary quality standards to IEEE 323-1974 and nuclear industry EPRI reports specified that 9 extrapolating beyond the extrapolation limits could invalidate the results of the Arrhenius methodology. The ancillary standards used for qualification of the various examples specified the limits of extrapolation to be no greater than 30 oC from the test data used to determine activation energies. In addition, the licensee did not consider adequate confidence bounds to account for the statistical uncertainties present when using the Arrhenius methodology. The inspectors noted that the uncertainties grow exponentially when exceeding the extrapolation limits. The Limitorque MOV motor life line appeared to have been extrapolated from test data at 240 C to 50 C, which is 190 C from the test data. The silicone rubber cable, life line appeared to have been extrapolated from test data at 210 C to 51.67 C, which is 188.3 C from the test data. The ASCO Valves, life line appeared to have been extrapolated from test data at 266 C to 40 C, which is 226 C from the test data. The Target Rock Valves, life line appeared to have been extrapolated upward from the test data. The Westinghouse RHR Motor rewind, life line appeared to have been extrapolated from test data at 180 C to 58.6 C, which is 121 C from the test data. 275 C is 216 C from test data The ancillary quality standards used in the qualification of these examples included IEEE 98-1972, IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials; IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data; IEEE 117-1974, IEEE Standard Test Procedure for Evaluation of Systems of Insulating Materials for Random-Wound AC Electric Machinery, and other quality standards. IEEE 98-1972, Section 10, Temperature Exposures, specified, in part, that the lowest test temperature shall be chosen so that the extrapolation necessary to establish the temperature index will not be more than 25 C. IEEE 101-1972, Section 1.3, Extrapolation, specified, in part, that extrapolation of the (qualified life) line below the range of test temperatures may cause erroneous predictions if the chemical reactions controlling the insulation aging are different at lower temperatures or if other conditions affecting the aging or the mode of failure are different. Therefore, the methods outlined in this guide are applicable only if all of the assumptions behind the use of the Arrhenius equation are met (identified in references). IEEE 117-1974, Section 3.3.1, Thermal Aging, specified, in part, for any system being evaluated, tests are made for at least three different temperatures. The lowest test temperature should be no more than 25 C above the system temperature rating. The highest temperature test should be at least 40 C above lowest temperature test, and temperature points should be selected to give approximately equal temperature intervals. The average life at the highest temperature shall be no less than 100 hours. The inspectors are concerned that the licensee did not meet the aforementioned Category 1 requirements in their licensing basis. The licensee has captured these concerns in their corrective action program as CR 1366022. The inspectors need further information from the licensee and NRC technical staff to evaluate the concerns. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-03, 05000328/2017008-03, Potential Inadequate use of thermal aging and the Arrhenius methodology)
05000327/FIN-2017008-02	2017Q4	Sequoyah	Inadequate Qualification for Unit 1 Reactor Lower Compartment Cooler Motors	The team identified a Green NCV of Title 10 Code of Federal Regulations 50.49(f) Electrical Equipment Qualification when the licensee failed to perform an adequate similarity analysis for the environmental qualification of their Reliance 75 horsepower reactor lower compartment cooling fan motors. The licensee entered this issue into their corrective action program as CR1366056 and performed an operability determination, which determined the reactor lower compartment cooling fan motors were operable but non-conforming in accordance with 10 CFR 50.49. The performance deficiency was determined to be more than minor because it was associated with the Design Control 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, failing to ensure the qualification of the reactor lower compartment cooling fan motors adversely affected their reliability and capability in the harsh environment of a design basis accident, which in turn adversely affected the reliability and capability of other environmentally qualified components that rely on the containment cooling system. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the team determined that the finding was of very low significance (Green) because it was a design deficiency that potentially affected the design or qualification of a mitigating system; however, the mitigating system maintained its operability. The team determined there was no cross-cutting aspect associated with this finding since it was not indicative of current licensee performance.
05000327/FIN-2017008-01	2017Q4	Sequoyah	Unjustified Qualified Life for Target Rock Power-Operated Relief Valves	The team identified a Green NCV of Title 10 Code of Federal Regulations 50.49(e)(5) Aging when the licensee failed to replace, refurbish, or demonstrate additional life for components that exceeded their qualified life. The licensee failed to justify changes to the accelerated aging calculations used for power operated relief valve harsh environmental qualification. The licensee entered this issue into their corrective action program as CRs 1365730 and 1366082, and performed operability determinations, which determined the systems were operable but non-conforming with 10 CFR 50.49. The performance deficiency was determined to be more than minor because it was associated with the Design Control 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, the failure to ensure that Target Rock power-operated relief valves were qualified for the duration that they were required to operate reduced the reliability of reactor coolant system in the harsh environments of design basis accidents. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the team determined that the finding was of very low significance (Green) because it was a design deficiency that affected the design or qualification of a mitigating system, however, the mitigating system maintained its operability. The team determined there was no cross-cutting aspect associated with this finding since it was not indicative of current licensee performance.
05000261/FIN-2017007-07	2017Q4	Robinson	Justification of Activation Energy of ASCO Solenoid Coil Assemblies	Introduction: 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-2017007-06	2017Q4	Robinson	Penetration F01 Submergence	Introduction: 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-2017007-05	2017Q4	Robinson	Questions Regarding EQDP-0401 Method Used to Determine Activation Energy and Responsibility for Verification	Introduction: 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-2017007-04	2017Q4	Robinson	Crouse-Hinds Qualification and Life Extension	Introduction: 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-2017007-03	2017Q4	Robinson	Failure to Determine Most Severe Containment Spray pH	The 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-2017007-02	2017Q4	Robinson	Failure to Perform Required O-ring Replacement to Maintain Qualification	The 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-2017007-01	2017Q4	Robinson	Failure to Correctly Determine Qualified Life	The 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.
05000425/FIN-2017009-01	2017Q4	Vogtle	Failure to Install Drain Hole	The NRC identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for licensees failure to verify drain holes were installed as assigned, following the licensees evaluation of Information Notice (IN) 89-63, Possible Submergence of Electrical Circuits Located Above the Flood Level Because of Water Intrusion and Lack of Drainage. Specifically, the licensee did not verify that that junction box 2BTJB0486 was equipped with a weep hole consistent with the assigned corrective action and the corrective action was closed without corrective action being taken. In response to the issue, the licensee initiated Condition Report (CR) 10439858, performed an immediate determination of operability, and determined that equipment associated with the cables in the junction box were Operable but Degraded/Non-conforming (OBDN), and plans to return the affected equipment to fully conforming status. 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 correcting the condition could cause submergence of the unqualified cables during events, which affects the reliability of the equipment. 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. No cross cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the error occurred on January 25, 1990.
05000321/FIN-2016010-02	2016Q2	Hatch	Failure to Identify N2E Nozzle Weld Through-Wall Flaw	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to promptly identify a condition adverse to quality regarding a through-wall flaw in the safe end-to-nozzle weld of the reactor coolant system N2E nozzle. The licensee has since repaired the flaw, completed all required postrepair examinations, and entered this issue entered this into their corrective action program as CR 10247856. 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. The inspectors screened this finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012. Because after a reasonable assessment of degradation, the finding could neither result in exceeding the RCS leak rate for a small LOCA, nor likely affected other systems used to mitigate a LOCA resulting in a total loss of their function, the finding screened as Green. This finding has a cross-cutting aspect of Challenge the Unknown in the area of Human Performance (H.11) because upon discovery of a less robust configuration of the N2E nozzle overlay, the licensee failed to consider the implications on the flaw that had existed in that component since 1988.
05000335/FIN-2015010-01	2015Q4	Saint Lucie	Implementation of Commitments and Aging Management Programs	The inspectors identified a URI associated with the implementation status of various commitments and AMPs. Description: The inspectors identified that there were pending actions for various regulatory commitments/AMPs as a result of commitment changes implemented by the licensee after the renewed operating license was issued. The licensee informed the NRC of such changes, and submitted correspondence to the NRC for review and approval. At the time of this inspection, the NRC was still in the process of reviewing the licensees submittals. While the licensee met its commitment to submit the proposed changes to the NRC prior to the PEO, the inspectors were unable to determine whether the licensees implementation of the affected AMPs was consistent with the staffs final position, which will be provided through the issuance of SERs. The affected commitment items, and their respective pending actions, are summarized below. Commitment 1, Condensate Storage Tank Cross-Connect Buried Piping Inspection On May 12, 2015, the licensee informed the NRC of a commitment change based on the as-found configuration of the cross-tie line after excavation. On September 1, 2015, the NRC issued a Request for Additional information, for which the licensee provided responses in letter L2015-258, dated October 6, 2015. At the time of this inspection the NRC was reviewing the licensees response to the Request for Additional Information, and no SER had yet been issued. Commitments 4 and 5, Reactor Vessel Internals Inspection Program As described in the inspection scope section of this report, the licensee submitted several letters to the NRC after the renewed operating license was issued describing the proposed program to manage the aging effects of the reactor vessel internals. At the time of this inspection, the NRC was reviewing the licensees submittals and no final SER had yet been issued. Commitment 6, Small Bore Class 1 Piping Inspection Program On May 11, 2015, the licensee submitted a revision to the previously approved Small Bore Class 1 Piping Inspection Program for NRC review and approval. The revision was related to the use of destructive examinations in lieu of volumetric examinations. At the time of this inspection, the NRC was reviewing the licensees submittal and no final SER had yet been issued. Commitment 20, Environmentally-Assisted Fatigue of the Pressurizer Surge Line On October 29, 2015, the licensee submitted its proposed program for managing environmentally-assisted fatigue of the pressurizer surge line to the NRC. The inspectors noted that the proposal detailed the licensees intent to utilize the ASME BPVC, Section XI ISI Program (UFSAR Section 18.2.2) to manage the recurring inspections, and the associated evaluations for any flaws noted. At the time of this inspection, the NRC was reviewing the licensees submittal and no SER had yet been issued. In addition to the commitment changes under NRC review, the inspectors identified a followup item for Commitment 17, Reactor Vessel Integrity Program. The inspectors noted that the licensee credited fleet procedure ER-AA-110 to meet the regulatory commitment associated with the integration of all four reactor vessel integrity subprograms into a single program document. Fleet procedure ER-AA-110 requires a plant-specific procedure be developed for each site describing the important parameters needed to meet the regulatory requirements specific to that station. The inspectors noted that the plant-specific procedure for Unit 1, procedure ADM 17.38, was still under development with a target completion date of March 1, 2016. Therefore, the inspectors concluded that there still were pending actions associated with the development of the site-specific program, and additional inspection was required to verify that the Reactor Vessel Integrity Program was implemented as intended. The licensee initiated AR 02094578 to enter this item in the CAP. The inspectors determined that it was necessary to open a URI to further review the implementation of the commitments/AMPs, and verify that the commitments were met as approved by the NRC in the final SERs. This issue requires followup inspection, and will be tracked as URI 05000335/2015010-001, Implementation of Commitments and Aging Management Programs.
05000280/FIN-2014007-01	2014Q3	Surry	Failure to Perform Required Preventative Maintenance on Class 1E Molded Case Circuit Breakers	The team identified a Green non-cited violation of Technical Specification 6.4.A.7, Unit Operating Procedures and Programs, for the licensees failure to implement written procedures to perform periodic tests for the Class 1E 125 volt direct current thermal-magnetic molded case circuit breakers (MCCBs). The licensee entered the issue into their corrective action program as condition reports CR558445 and CR560488 and performed an immediate determination of operability, in which they determined that the MCCBs were operable but not fully qualified. The licensees failure to conduct periodic tests to detect the deterioration of the system and to demonstrate that components not exercised during normal operation of the station are operable, as required by IEEE 308-1970, Section 6.3, was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, absent testing to detect deterioration and to demonstrate continued operability, the likelihood that these MCCBs will unpredictably fail when called upon increases with time in service. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and Inspection Manual Chapter 0612, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and 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, which maintained its operability or functionality. The team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance.
05000280/FIN-2014007-02	2014Q3	Surry	Failure to Evaluate the Range of Conditions that Effect Canal Level Probes	The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate and quantify the system response times and accuracies over the range of conditions under which the service water canal level probes must operate. The licensee entered the issue into their corrective action program as condition report CR558429 and performed an immediate determination of operability, in which they determined the canal level probes to be operable but not fully qualified. The licensees failure to evaluate conditions that affected system response times and accuracy of the canal level probes, as required by IEEE 279-1968, Section 4.1, 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 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, response time delays could allow the canal water level to fall below Technical Specification limits reducing the available heat removal required to mitigate Updated Final Safety Analysis Report chapter 14 design basis accidents. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and Inspection Manual Chapter 0612, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and 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, which maintained its operability or functionality. The team determined that the finding was associated with the Design Margin cross-cutting aspect of the Human Performance area because recent modification designs for the canal probes were completed and approved without evaluating effects on the canal level probe response times and accuracies.
05000280/FIN-2014007-03	2014Q3	Surry	Adequacy of Class 1E 125VDC Branch Circuit Breaker Design	The team identified an Unresolved Item (URI) regarding the adequacy of design of the Class 1E 125VDC power branch circuit breaker for the 1H 4160V Bus controls. The team reviewed the Class 1E 125VDC power distribution design to verify compliance with the licensing basis requirements in IEEE 308-1970, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations. The Surry licensing basis commitment to IEEE 308-1970 required the quality of the Class 1E power system design to be sufficient to ensure that multiple engineered safety features (ESF) would not lose power because of design vulnerabilities. Specifically IEEE 308-1970 stated, in part, The Class IE electric systems shall be designed to assure that any design basis event as listed in Table 1 will not cause: 1) A loss of electric power to a number of engineered safety features, surveillance devices, or protection system devices sufficient to jeopardize the safety of the plant. Table 1 stated, in part, that design basis events include Single act, event, component failure, or circuit fault that can cause multiple equipment malfunctions. The team identified design vulnerabilities in design basis documents and in the sampled branch circuitry. In Calculation EE-0499, DC Vital Bus Short Circuit Current, dated 11/30/1998, the licensee used AC power time current curve (TCC) data for HFB MCCBs (used in the 125VDC distribution system) instead of DC TCC data. In addition, in this calculation, the licensee did not de-rate components for the ambient temperature in the switchgear room. Furthermore, in 2009, the licensee replaced certain HFB MCCBs with model HFDDC MCCBs; however, did not evaluate the DC characteristics of these HFDDC MCCBs, and instead evaluated an AC model HFD MCCB. Because of these vulnerabilities the team questioned the coordination of the installed HFDDC breaker and whether it was adequate to protect the 1H branch circuit in the ambient temperature of the switchgear room. These calculational vulnerabilities were consistent across both trains A & B and for both Units 1 & 2. The licensee captured the inspectors questions in their corrective action program as CR559872 and CR559875. This issue is a URI pending further review of information provided by the licensee on November 4, 2014, and consultation with the Office of Nuclear Reactor Regulation to determine if this issue of concern constitutes a violation.
05000280/FIN-2014007-04	2014Q3	Surry	Qualification Basis for Safety-Related Molded Case Circuit Breakers	The team identified a URI regarding the licensees actions to maintain or extend the qualification basis for safety-related MCCBs installed in mild environments greater than vendor design life specifications. In 2004, the licensee received Westinghouse Electric Technical Bulletin TB-04-13, Replacement Solutions for Obsolete Classic MCCBs, UL (Underwriters Laboratory) Testing Issues, Breaker Design Life and Trip Band Adjustment, which was superseded in 2006 by TB-06-02, Aging Issues and Subsequent Operating Issues for Breakers That are at Their 20-Year Design/Qualified Lives; UL Certification/Testing Issues Update. These bulletins informed the licensee of MCCB aging and operating issues. Specifically, grease and red oil used in these breakers were found to be key limiting factors for continued operability within published specifications. As grease and red oil aged beyond 20 years, their lubrication properties were reduced, resulting in slower trip times beyond the published time-current curves. The bulletins further defined the design life of MCCBs in mild environments as 20 years. However, the inspectors noted that approximately 60 safety-related MCCBs installed in mild environments exceeded 20 years of service, and the licensee had not performed an engineering evaluation to justify continued operation beyond this design life. The affected MCCBs were associated with the Class 1E 125VDC distribution systems (switchgear) on both units. The licensee captured the inspectors questions in their corrective action program as CR558445 and CR560488. This issue is a URI pending further review, including consultation with the Office of Nuclear Reactor Regulation, to determine if this issue of concern constitutes a violation.
05000327/FIN-2014002-01	2014Q1	Sequoyah	Inadequate Clearance Causes Control Air System Transient	A self-revealing non-cited violation of Units 1 and 2 Technical Specification 6.8.1.a, Administrative Controls (Procedures), was documented for the licensees failure to establish an adequate clearance in preparation for maintenance activities on the B station air compressor. Implementation of this inadequate clearance on February 21, 2014, resulted in a reduction of control air pressure and a plant transient which challenged control room operators. Immediate corrective action was to revise the clearance to establish an adequate boundary. The licensee entered the issue into the corrective action program (CAP) for resolution as PER 850331. The performance deficiency was more than minor because it was associated with the configuration control and human performance attributes of the initiating events cornerstone and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inadequate clearance caused a plant transient during power operations that without operator action would have resulted in a loss of air operated plant components and ultimately require the operators to trip both units. The finding was determined to be of very low (green) safety significance based on Exhibit 1, Initiating Events Screening Questions, found in Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process for Findings At-Power, because the finding did not result in a complete or partial loss of a support system that contributed to the likelihood of, or cause, an initiating event and affected mitigation equipment. The inspectors determined the cause of this finding was associated with a cross cutting aspect of Work Management in the Human Performance area. Specifically, the licensee failed to implement their clearance process such that nuclear safety was the overriding priority.
05000250/FIN-2014007-01	2014Q1	Turkey Point	Failure to Properly Implement Time Critical Operator Action Program Procedure	The team identified a non-cited violation of Technical Specification 6.8.1, Procedures and Programs, for the licensees failure to implement procedure 0-ADM-232, Time Critical Action Program, to ensure time critical actions (TCAs) important to mitigate design basis events could be performed in the required time. The failure to implement this procedure was a performance deficiency. No documentation existed to demonstrate that the TCA to restore power to the battery chargers during a station blackout could be performed within the required time (30 minutes). The team also identified a TCA to locally isolate the auxiliary feedwater for a faulted steam generator that did not have a job performance measure to demonstrate the successful completion of the action. The licensee entered this issue into the corrective action program as action requests 01944453, 01945532, 01943321, 01943425, and 01943697. For TCAs where no validation documentation could be determined, the licensee completed tabletop exercises, simulator exercises, and field walkdowns to ensure that all of the TCAs to mitigate design basis events could be completed within the required action times. The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not implement 0-ADM-232 adequately to ensure that the TCAs listed in Attachment 1 of the procedure were properly validated; consequently, the licensee could not demonstrate that TCAs could be successfully executed in accordance with the design basis. The team determined the finding to be of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; and did not represent a loss of system and/or function. The team determined this finding was associated with the cross-cutting aspect of Procedure Adherence in the area of Human Performance because although the procedure was recently revised to include all necessary requirements to maintain the time critical action program, the licensee failed to follow procedure 0-ADM-232, which resulted in several TCAs not being properly validated.
05000327/FIN-2014002-02	2014Q1	Sequoyah	Licensee-Identified Violation	Failure to comply with technical specifications during refueling operations Unit 1 technical specification (TS) 3.3.9.4.c requires that during refueling operations, each penetration providing direct access from the containment atmosphere to the outside atmosphere be closed by a manual valve (if so equipped). Contrary to the above, between October 19 and 22, 2013, there were several instances where a Unit 1 containment penetration, X-108, to the additional equipment building was open (including its associated manual valve) during movement of irradiated fuel. This problem was entered into the licensees corrective action program as PER 800432, 806293, and 824224. Using Inspection Manual Chapter 0609, Appendix G, Shut-down Operations Significance Determination Process dated February 28, 2005 the inspectors determined that, the finding was Green because it did not: 1) involve a loss of reactor coolant system (RCS) inventory; 2) degrade ability to terminate a leak path or add RCS inventory as needed; or 3) degrade the ability to recover RHR once it was lost. This issue is also discussed under Section 4OA3 of this report.
05000327/FIN-2014002-03	2014Q1	Sequoyah	Licensee-Identified Violation	Failure to perform adequate post maintenance testing of the 1B EDG 10 CFR 50, Criterion XI, Test Control requires in part that an established testing program shall require that all testing of SSCs ensure that the SSC can perform its intended function. Contrary to the above, on February 23, 2014, adequate testing to ensure the EDG air start motors could fulfill their required functions was not performed. An adequate test was not performed until March 16 which was part of an annual testing program. This problem was entered into the licensees corrective action program as PER 859633. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the 1B EDG retained the capability to automatically start despite the improper air hose configuration of the air start motors.
05000327/FIN-2013007-05	2013Q3	Sequoyah	Failure to Document Deficiencies Discovered During Receipt Inspections in the Corrective Action Program	The team identified a non-cited violation of TS 6.8.1, Procedures and Programs, the licensees failure to properly implement maintenance procedures for performing receipt inspection of new 480V circuit breakers. Specifically, the licensees failure to evaluate the need to report defects and deficiencies, identified on new safety-related 480V circuit breakers, in the corrective action program as prescribed by procedure was a performance deficiency. The licensee corrected the deficiencies prior to putting the breakers in service. This issue was entered into the licensees corrective action program as PERs 763834 and 759238. This performance deficiency was determined to be more than minor because if left uncorrected could lead to a more significant safety concern. Specifically, not documenting deficiencies that could adversely affect the breakers in the corrective action program, would not ensure breaker issues were being properly trended and that the issues have been adequately corrected and are not recurring. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because it was not a design deficiency resulting in the loss of functionality or operability. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area Human Performance, component of Work Practices because the licensee failed to meet expectations regarding procedural compliance and did not follow procedures related to 480V safety-related breaker receipt inspections.
05000327/FIN-2013007-07	2013Q3	Sequoyah	Failure to Perform 50.59 Screens for Scaffolds and Clearances	The team identified a non-cited violation of TS 6.8.1, Procedures and Programs, for the licensees failure to implement procedures for equipment and maintenance control. The licensees failure to perform 10 CFR 50.59 reviews of temporary plant changes (e.g., scaffolding and clearances) that existed for greater than 90 days of plant operation was a performance deficiency. The licensee implemented corrective actions to review all of the temporary plant changes. The licensee generated PERs 756276, 753175, and 756308. This performance deficiency was determined to be more than minor because it affected the Design Control 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, the team identified multiple examples where the licensee failed to evaluate temporary plant changes to ensure those changes did not affect the availability, reliability, and capability of systems that respond to events. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because it was not a design deficiency resulting in the loss of functionality or operability. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area of Human Performance, component of Work Practices because licensee failed to meet expectations regarding procedural compliance and did not follow procedures related to performing 50.59 reviews of temporary plant changes that existed for greater than 90 days of plant operation.
05000327/FIN-2013007-08	2013Q3	Sequoyah	Inadequate Corrective Action for 2010 Degraded Voltage Issues	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct deficiencies in electrical calculations for the safety-related AC electrical distribution system identified during the 2010 CDBI. Specifically, the licensees failure to identify that safety-related motor operated valve (MOVs) needed to be evaluated for new lower calculated available voltage (degraded voltage) to ensure their operability was a performance deficiency. The licensee initiated PER 753504 and performed a prompt determination of operability (PDO). The team concluded that the evaluations and compensatory measures described in the PDO provided reasonable expectation of operability. The performance deficiency was determined to be more than minor because it affected the Design Control 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 identify and evaluate that safety-related MOVs could be affected by degraded voltage conditions did not ensure the availability, reliability, and capability of the MOVs to respond to initiating events. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area of Problem Identification and Resolution, component of Corrective Action Program because the licensee failed to identify that safety-related MOVs needed to be evaluated for new lower calculated available voltage (degraded voltage) to ensure their operability.
05000327/FIN-2013007-04	2013Q3	Sequoyah	Inadequate Basis for AFW MOV Motor Brake Alternate Voltage Criteria	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to check the adequacy of the design of the steam generator feedwater isolation valve motor brakes. Specifically, the licensee based voltage acceptance criterion of 74% of 460V for motor brakes used in a design basis calculation on inadequate testing and calculational methods. This was a performance deficiency. In response to the teams concerns, the licensee initiated PER 763818 and provided reasonable expectation of operability of the motor brakes, by use of administratively controlled voltage, pending restoration of full qualification. This performance deficiency was determined to be more than minor because it affected the Design Control 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, inadequate design criteria did not ensure the availability, reliability, and capability of the steam generator feedwater isolation valve motor brakes to operate under design basis degraded voltage conditions. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency was not indicative of present licensee performance.
05000327/FIN-2013007-03	2013Q3	Sequoyah	Failure to Properly Translate the Design and Licensing Bases for the 125 VDC System Into Design Calculations	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly translate the design and licensing bases for the 125 VDC system into design calculations. The licensee inappropriately credited the battery chargers for voltage support during accident scenarios in their voltage drop calculations, and failed to include vital inverters in the battery load profile. This was a performance deficiency. In response to the teams inquiries, the licensee initiated PER 758465 that provided reasonable expectation of operability by demonstrating that the required voltages would be available. This was based on interpolation of the vendor battery curves considering the maximum loading on the battery for the applicable portions of the duty cycle. This performance deficiency was determined to be more than minor because it affected the Design Control 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, the licensee failed to properly evaluate the 125 VDC system under accident conditions to ensure the capability and availability of 125V control circuits to operate during design basis events. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency was not indicative of present licensee performance.
05000327/FIN-2013007-02	2013Q3	Sequoyah	Failure to Evaluate Impact for Full Range of EDG Frequency	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate the entire range of allowable emergency diesel generator (EDG) frequencies into design basis documents. The failure to analyze the effects of the technical specification allowable EDG frequency range on the safety-related components powered by the EDGs was a performance deficiency. The licensee entered this issue in their corrective action program as PER 758761 and performed a prompt operability evaluation to determine that the safety-related equipment powered by the EDGs with a limited frequency range variation of 59.9 to 60.1 Hz, would be able to perform their design basis functions under accident conditions. In addition, a review of the results of the EDGs surveillances indicates no history of being outside the range of 59.9 to 60.1 Hz for the last three years. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to account for the allowable range of the EDG frequency and not evaluating the impact on safety related components powered by the EDGs did not ensure the availability and capability of safety-related components to respond to initiating events. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area of Human Performance, component of Resources because the licensee failed to ensure that design calculations affected by EDG frequency were complete and accurate.
05000327/FIN-2013007-01	2013Q3	Sequoyah	Failure to Evaluate a Potential Condition Adverse to Quality Prior to Mode Change	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow a test control procedure to evaluate indications of excessive check valve leakage prior to changing modes. Specifically, the licensee failed to evaluate the potential inoperability of residual heat removal check valve 2-63-563, which exhibited indications of excessive leakage, as required by procedure NPG-SPP- 06.9.1, Conduct of Testing, prior to transitioning to Mode 3, during startup. This was a performance deficiency. After conducting interviews with operations staff and performing a prompt determination of operability, the licensee concluded that the valve was never inoperable, since the valve subsequently passed its leak rate test in Mode 3 with no maintenance being performed. The operability determination was documented in PER 757559. This performance deficiency was determined to be more than minor because it affected the Human 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, failing to evaluate indications of excessive check valve leakage while performing procedure 2-SISXV- 063-206.0, ECCS Check Valve Leak Testing section 6.3.2, adversely affected the cornerstone objective of limiting the likelihood of events that challenge the critical safety function of maintaining the RCS pressure boundary. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding would not have affected other systems used to mitigate a LOCA resulting in a total loss of their functions. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area of Human Performance, component of Decision-Making because the licensee did not use conservative assumptions in their decision making when they failed to evaluate the potential inoperability of check valve 2-63-563 prior to transitioning to Mode 3.
05000327/FIN-2013007-09	2013Q3	Sequoyah	Insufficient EDG Starting Air Pressure Following SBO Coping Period	The team identified an unresolved item (URI) associated with licensee?s capability to meet their station blackout (SBO) mitigation strategy. Specifically, based on the allowable air start check valve leakage and the amount of air used during start attempts of the EDGs, the team found that the licensee did not ensure if adequate starting air pressure would exist to reliably start the EDGs following a SBO. Title 10 CFR 50.2, Definitions, defines a SBO as the complete loss of ac power to the essential and nonessential switchgear buses in a nuclear power plant, concurrent with turbine trip and unavailability of the onsite emergency power system. Essentially, this would involve the loss of the offsite power sources as well as the loss of emergency onsite AC power sources. The licensee is committed to coping with an SBO event for a duration of four hours, after which the licensee will recover AC power. The EDG air start system provides compressed air to start the EDGs. The compressed air is provided by non-safety related air compressors, and is stored in two safety-related air receiver tanks. Receiver tank ?A? is designed to maintain the air between 250 and 300 psig; tank ?B? is designed to maintain between 185 and 200 psig. The EDG air start system is equipped with check valves to maintain the integrity of the safety-related portion of the air start system. The licensee declares the EDG degraded if the receiver tank ?A? is less than 200 psig, due to the inability to meet the five start design basis requirement as described in UFSAR, Section 9.5.6, Diesel Generator Starting System. The EDG is declared inoperable at pressures below 150 psig on receiver ?B? due to the loss of start capability. This is based on the manufacturer?s value at which EDG starting and achieving rated speed and voltage has been demonstrated by testing. The team noted that the leak rate acceptance criterion outlined in procedure 0-PI-SXV- 082-203, Diesel Starting Air Valve Test, was 5 psig/minute for the EDG air start check valves. At this allowable leak rate, the EDG air start pressure could fall below 150 psig within 1 hour after an SBO and completely depressurize the air receiver within 3 hours after an SBO. This would not support the capability of the EDGs to start at the end of the 4-hour SBO coping period. In addition to concerns regarding check valve leak rate acceptance criteria, the team noted that postulated failed start attempts during an SBO event would also adversely impact the amount of air that would be available at the end of the 4- hour coping period. Specifically, in a SBO event, the initial failure of the onsite power sources would be followed by a failure of both onsite EDGs to start. The licensee?s procedures direct operators to attempt to start the EDGs a second time in the first few minutes of the SBO. The first and second start attempts are postulated to be unsuccessful during an SBO. The loss of offsite and onsite emergency ac power would prevent the air start compressors from recharging the tanks after the failed start attempts. Based on allowable check valve leakage and the amount of air used during two failed start attempts of the EDGs, the team found that the licensee did not ensure if adequate starting air pressure would exist to reliably start the EDGs in order to recover from an SBO after the 4 hour coping period. The team also found that the licensee had not developed procedural guidance to provide adequate air pressure to reliably start the EDG in order to recover from a SBO after the 4-hour coping period. The licensee captured these concerns in PER 763335. This issue remains unresolved pending inspector consultation with NRC headquarters technical staff for clarification of the licensee?s current license basis design requirements (with respect to 10 CFR 50.63 compliance), to determine if a performance deficiency exists. This issue is being identified as URI 05000327, 328/2013007-09, Insufficient EDG Starting Air Pressure following SBO Coping Period.
05000327/FIN-2013007-06	2013Q3	Sequoyah	Failure to Adequately Translate Design Basis Into Procedure Acceptance Criteria Time to Perform Operator Action	The team identified an Apparent Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate design basis requirements into emergency sub-procedure, ES- 1.3, Transfer to Residual Heat Removal Containment Sump, Rev. 19. Specifically, the time allotted for operators to perform time critical actions to swap emergency core cooling system (ECCS) pump suction from the refueling water storage tank (RWST) to the containment sump during a small break loss of coolant accident (SBLOCA) did not properly account for the instrument uncertainty and the design basis requirement in Updated Final Safety Analysis Report 15.3.1, to ensure the recovery of the core was demonstrated and to ensure continuous operation of the ECCS. This was a performance deficiency. As immediate corrective action, the licensee performed an operability review and documented the results in the corrective action program as PERs 760336 and 758761. The licensee concluded that there were no current operability concerns, and created Standing Order SO-13-025 to reinforce operator time performance requirements. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected objective of ensuring the availability, reliability, and capability of containment spray pumps, safety injection pumps, and charging pumps during a SBLOCA. Specifically, the licensee failed to demonstrate that operators would be able to successfully complete the time critical actions prior to reaching 8% RWST tank level, which required operators to secure all pumps taking suction from the RWST, because they did not consider the worst case allowable RWST level instrument uncertainty acceptance criteria along with the design pump flow rates. This action would result in the momentary loss of all ECCS high pressure injection during a SBLOCA and did not ensure the availability, reliability, and capability of the ECCS to respond to initiating events. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power. The safety significance is to be determined pending review and analysis of additional information from the licensee to determine if this finding is representative of an actual loss of the ECCS safety function. As a result, this finding is characterized as TBD. The finding did not represent an immediate safety concern because a review of past results indicated that operators were consistently performing the actions in times less than required, as documented by simulator testing. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of present licensee performance.
05000259/FIN-2013007-05	2013Q2	Browns Ferry	Security
05000261/FIN-2013007-05	2013Q2	Robinson	Failure to Have Appropriate Procedure to Verify Degraded Voltage Relay Circuit Status	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe an adequate procedure that verified DGVR circuit operability following degraded voltage disable switch operation for reactor coolant pump (RCP) starts. This was a performance deficiency. The licensee entered the issue into the corrective action program as Nuclear Condition Report 602516, developed a test procedure, and verified the DGVR operability on both emergency buses. The performance deficiency was more than minor because if left uncorrected, it could become a more significant safety concern. Specifically, by not properly testing the DGVR circuit to ensure continuity following switch manipulation for RCP starts, the circuit could unknowingly become inoperable and non-functional for an entire operating cycle. The finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system function, did not represent an actual loss of function of at least a single train for greater than its technical specification (TS) allowed outage time or two separate safety systems out-of-service for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-TS trains. No cross-cutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the modification that added the degraded voltage disable switches.
05000261/FIN-2013007-04	2013Q2	Robinson	Failure to Have Adequate Analyses for the E1 Bus Fast Transfer	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the plant design during fast bus transfers. Specifically, the licensee failed to have an adequate analysis that ensured a successful fast bus transfer of the safety-related E1 bus feeder from the Unit Auxiliary Transformer to the Startup Transformer when required. This was a performance deficiency. The licensee entered the issue into the corrective action program as Nuclear Condition Reports 603357 and 605562, and performed an additional fast bus transfer evaluation of the E1 feeder breaker to ensure that the breaker would not trip under fast bus transfer conditions. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of safety related loads on the E1 bus because the licensee did not verify the E1 feeder 4 breaker would not trip during a fast bus transfer. The finding was determined 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 and functionality. No cross-cutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the fast bus transfer evaluation.
05000259/FIN-2013007-04	2013Q2	Browns Ferry	Failure to Adequately Identify, Evaluate, and Correct the EECW Strainers Degraded/Nonconforming Condition	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and take corrective actions to address a non-conforming condition adverse to quality related to three faulted strainers in the safety related Emergency Equipment Cooling Water system. This was a performance deficiency. The licensee initiated Problem Evaluation Report 677627 to perform a new operability evaluation since the operability evaluation in Problem Evaluation Report 208636 was found to be inadequate. The licensee concluded that there were no current operability issues. The performance deficiency was determined to be more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the core spray system to respond to initiating events, in that, if left uncorrected could result in the plant not being able to sustain short-term heat removal under specific conditions. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. The team evaluated the finding for cross-cutting aspects and determined the finding was associated with the corrective action program component of the problem identification and resolution area, because the licensee did not perform a thorough evaluation of identified problems such that the resolutions addressed the underlying causes and extent of condition.
05000261/FIN-2013007-03	2013Q2	Robinson	Failure to Have Adequate Analyses Supporting the Degraded Voltage Relay Setpoints	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to have adequate analyses that supported safety-related load operation during a design basis accident while supplied by offsite power. This was a performance deficiency. The licensee entered the issue into the corrective action program as Nuclear Condition Reports 601201 and 605969, and performed an evaluation that determined the capability of starting the safety-related motors at degraded voltage conditions, as well as the capability of the electrical loads during the degraded grid voltage relay (DGVR) time delay to ensure equipment function was preserved. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of safety related loads to respond to a design basis accident under degraded voltage conditions. Evaluations of the effects of starting motors at the DGVR voltage dropout setpoint and the equipment survivability during the DGVR time delay were not performed. The team determined the finding required a detailed risk analysis, because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, and the team assumed the performance deficiency represented a loss of operability or functionality of the equipment that could be lost during the DGVR time delay. This assumption was made to bound the risk of the finding, because the licensee was still investigating whether or not there would be a loss of function of any equipment during the DGVR time delay period as of the date of this inspection report issuance. The team assumed a recoverable loss of function of all 480V motor control centers and assumed a degraded voltage condition exposure time of one hour per year. The one hour per year assumption is conservative relative to actual plant data which indicated a degraded voltage condition exposure of 44 seconds over the past 3 operating years. The results of the detailed risk analysis indicated an increase in core damage frequency <1E-6/year, which is representative of a finding of very low safety significance (Green). No crosscutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the degraded voltage evaluation.
05000259/FIN-2013007-03	2013Q2	Browns Ferry	Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to perform analyses demonstrating that the degraded voltage relay (DVR) set points specified in technical specifications (TS) would ensure adequate voltage to safety-related equipment. This was a performance deficiency. The licensee entered this issue into their corrective action program as PERs 676678 and 696876. As immediate corrective actions, the licensee performed a sensitivity study to verify that the voltage at the DVR set points specified in TS could provide adequate starting voltage to a sample of limiting safety-related equipment. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the 4160 volts alternating current buses. Specifically, the finding challenged the assurance that safety-related loads had adequate motor starting voltage during required degraded voltage scenarios. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since 1993 and was not indicative of current licensee performance.
05000261/FIN-2013007-02	2013Q2	Robinson	Failure to Evaluate SBO Coping Equipment for Environmental Conditions	The team identified a Green finding for the licensees failure to follow NRC Regulatory Guide 1.155, Station Blackout, guidance (to which they are committed in the Updated Final Safety Analysis Report) for evaluating equipment needed to cope with a station blackout for the required duration for associated environmental conditions. This was a performance deficiency. The licensee entered the issue into their corrective action program as Nuclear Condition Report 600522, and established a calculation that determined the maximum expected temperature inside the compartment housing the dedicated shutdown diesel generator (DSDG) and evaluated the equipment to determine its capability to perform its function for the station blackout coping duration. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the capability and reliability of the equipment located in the DSDG compartment was not ensured since a comparison of equipment temperature ratings and expected DSDG compartment temperatures was not performed. The finding was determined 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, and the structure, system, or component maintained its functionality. No cross-cutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the installation of the DSDG.
05000259/FIN-2013007-02	2013Q2	Browns Ferry	Failure to Evaluate the Effects of the Failure of Non-Class 1E Load Center Transformer Cooling Fans on the Class 1E 4160-480V Load Center Transformers and 480V Shutdown Boards	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, involving the failure to evaluate the effects of a postulated failure of the load center transformer non-safety-related, non-Class 1E cooling fans, which includes the fan power wiring and fan control equipment, on the safety-related Class 1E shutdown board load center transformers and 480V shutdown boards. This was a performance deficiency. The licensee tested the fans and performed an operability evaluation as documented in Problem Evaluation Report 682254 to provide reasonable assurance that the safety-related transformers would not be damaged from postulated failures from the non-safety-related fans and be capable of operating when required for the design basis accident conditions. The performance deficiency was determined to be more than minor because the finding affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the load center transformers TS1A and TS1B and the 480V shutdown boards 1A and 1B respectively. Specifically, the licensee had not evaluated the effects of the failure of non-safety-related transformer cooling fans, on both the safetyrelated load center transformer and 480V shutdown board and resulted in a reasonable doubt of operability. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since November 2004; therefore, not indicative of current licensee performance.
05000261/FIN-2013007-01	2013Q2	Robinson	Failure to Account for Containment Temperature Measurement Uncertainty	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for instrument uncertainty on the containment bulk temperature instrumentation which was used to verify technical specification (TS) containment operability. This was a performance deficiency. The licensee entered this issue into their corrective action program as Nuclear Condition Report 603294 and performed an evaluation of the temperature instrumentation uncertainty. In addition, the licensee issued Standing Instruction 13-001 which specified the indicated containment temperature for entry into TS Limiting Condition for Operation 3.6.5 was to be 118 degrees Fahrenheit, a value that compensated for the temperature measurement uncertainty. The performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, if the licensee did not account for the temperature measurement accuracy, containment temperature could unknowingly exceed the TS operability limit, and the licensee may not declare containment inoperable. The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components and did not involve a reduction in function of hydrogen igniters in the reactor containment. The cause of the finding was indicative of current licensee performance because the licensee failed to consider instrument uncertainty when they performed a containment re-analysis in 2013. The cause of the finding was directly related to the maintaining long term plant safety by maintenance of design margins cross-cutting aspect of the resources component in the area of human performance because when the containment re-analysis was performed, the licensee reduced margin between the analyzed value for containment starting temperature and the TS limit, making the instrument uncertainty of the temperature instruments more significant.
05000259/FIN-2013007-01	2013Q2	Browns Ferry	Failure to Verify the Capability of HPCI to Achieve Required Flow and Pressure within 30 Seconds Under Accident Conditions	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure that post-maintenance and post-modification testing of the high pressure cooling injection (HPCI) pump adequately demonstrated that it could achieve design basis flow within 30 seconds from a cold, non-oil-primed, turbine quick start under design basis conditions. This was a performance deficiency. The test configuration was less limiting than the design basis accident configuration, and the licensee had not verified by calculation or testing that the acceptance criteria in the test was adequate to demonstrate the HPCI pump could perform its function under design basis conditions. The licensee performed an operability review and documented the results in the corrective action program as Problem Evaluation Report 690086. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the HPCI pumps. Specifically, using procedure 3-SR-3.5.1.7, the licensee failed to demonstrate that the HPCI pump could achieve the required flow and discharge pressure under accident conditions as required by the design basis. Additional analysis was required to verify system operability. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since the original design of the plant and was not indicative of current licensee performance.
05000261/FIN-2012005-02	2012Q4	Robinson	Failure to Effectively Implement Gas Intrusion Program	The inspectors identified a Finding for the licensees failure to perform the 18- month pre-refueling outage (RO) ultrasonic testing (UT) examinations on 47 potential gas accumulation locations required by plant operating manual PLP-085, Emergency Core Cooling Systems Gas Management Program (GL 2008-01). Compliance with PLP-085 ensures the capability of the safety injection (SI), residual heat removal (RHR), and containment spray (CS) systems to perform their safety-related functions, and effectively implements the licensees gas management program as committed to the NRC in response to Generic Letter 2008-01. The licensee entered the issue into the corrective action program (CAP) as nuclear condition report (NCR) 575063, and is evaluating corrective actions. The failure to perform pre-RO UT examinations on 47 potential gas accumulation locations, as required by PLP-085 was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, if the licensee continued to miss pre-RO UT examinations, conditions that result in the formation of voids in the SI, RHR, and CS systems could go undetected with the potential to adversely affect the systems capability to perform their functions. The inspectors assessed the finding using IMC 0609 Attachment 4, Initial Characterization of Findings; and IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, and determined the finding was of very low safety significance (Green) because it was not a design deficiency, it did not represent the loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The inspectors identified a cross-cutting aspect in the work practices component of the human performance area, because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel following procedures. Specifically, on two occasions, the licensee did not perform pre-RO UTs in accordance with their gas management program, as described in PLP-085.
05000261/FIN-2012005-01	2012Q4	Robinson	Adequacy of Preventative Maintenance for the Dedicated Shutdown Diesel Generator Cooling System	On October 2, 2012, during monthly testing of the DSDG in accordance with OST-910, Dedicated Shutdown Diesel Generator Monthly, the control room received a DSDG Trouble alarm. Shortly after the alarm was received, the DSDG tripped. The licensee determined that the DSDG automatically tripped due to an engine jacket water over temperature condition. After the trip, licensee personnel inspected the engine and discovered that the drive belts for the belt driven radiator fan had come off the pulleys which prevented proper heat removal from the engine cooling system. All three drive belts were found to have varying degrees of wear and degradation. The last visual inspection of the fan belts was performed on September 12, 2011 and the last satisfactory surveillance run was performed on August 28, 2012. The DSDG is required to supply back-up power during a 10 CFR 50.65 Station Blackout condition and Appendix R conditions. Following the discovery of the thrown belts, the licensee replaced all three belts and performed a root cause evaluation. The root cause team determined that the cause of the failure was the lack of a time based replacement of the fan belts. The belts were last replaced in 2003. The inspectors reviewed the licensees root cause and asked additional questions regarding the expected service life of the fan belts. Additional inspection is required to review the licensees response to the inspectors questions and determine if a performance deficiency exists.
05000261/FIN-2012005-03	2012Q4	Robinson	Questions Regarding Whether GOTHIC is Sufficiently Qualified for Use in Operability Determinations	Information Notice 2011-17, issued July 26, 2011, informed addressees of recent instances of gas accumulation in safety-related systems in which the resulting operability determination of the as-found condition relied on computer models (i.e., GOTHIC) that were not demonstrated to be technically appropriate for the intended application. Specifically, the computer models had not been sufficiently qualified by benchmarking against test or plant data. The inspectors reviewed information related to the licensees response to GL 2008-01 and determined that the licensee had found voids in the SI system, RHR system, and CS piping. In most instances, the licensee had used GOTHIC to evaluate the past operability of the subject systems with voids, and then vented the gas prior to returning the subject systems back to service. The licensee had also evaluated the continued operability of the subject systems with a void left in place until corrective actions were implemented. Specifically, in 2008, the licensee evaluated eight gas voids found following filling and venting of the subject systems that could not be successfully removed during RO-25. The inspectors observed that the licensee used the GOTHIC as part of these evaluations to perform analysis of gas movement to predict how a void volume in piping is translated into a transient void fraction at the entrance of the pumps. The evaluations were the basis for the continued operability until corrective actions could be taken to remove the voids during the following RO-26, approximately 19 months later. While acknowledging the NRCs concerns that the GOTHIC models may not be sufficiently qualified by benchmarking against test or plant data for the particular gas transport scenario and piping configuration being analyzed, the licensee prepared engineering change document EC 86423 to document their justifications for continued use of the GOTHIC models to support operability determinations. The inspectors determined that this issue will remain unresolved pending additional inspection and consultation with a GOTHIC subject matter expert at NRC headquarters to evaluate the licensees use of GOTHIC to support operability determinations. This issue will be identified as URI 05000261/2012005-03, Questions Regarding Whether GOTHIC is Sufficiently Qualified for Use in Operability Determinations.
05000261/FIN-2012005-04	2012Q4	Robinson	Questions Regarding the Adequacy of the Fill and Vent Procedure for the Residual Heat Removal Heat Exchangers	Procedure OP-201-1, RHR System Venting directs system venting by a series of static and dynamic venting evolutions. The inspectors noted that the procedure did not specify the minimum flowrates necessary to ensure an adequate dynamic flush of the HXs. Specifically, the inspectors identified that dynamic venting of the system is performed by establishing flow via both the RHR HXs and its bypass line, which reduces the effective flow available to dynamically vent the HXs. The licensee indicated that following the fill and vent procedure, operations performs a post maintenance test (per OST-253, Comprehensive Flow Test for the RHR Pumps ), before returning the system to service, that establishes full flow through the HXs and would completely vent the HXs if the initial fill and vent was not successful. The inspectors was concerned because establishing full flow through the HXs with a large enough void size inside the HXs could potentially result in a water hammer condition that exceeds the structural design limitations of the system. The licensee is performing an evaluation to determine if any voids could be left in the HXs after fill and vent, and what the potential effects on the system could be. The inspectors determined that this issue will remain unresolved pending additional inspection to evaluate the licensees evaluation. This issue will be identified as URI 05000261/2012005-04, Questions Regarding the Adequacy of the Fill and Vent Procedure for the Residual Heat Removal Heat Exchangers
05000335/FIN-2012004-01	2012Q3	Saint Lucie	Failure to Ship Radioactive Material in Accordance with DOT Regulations	A self-revealing, Green non-cited violation (NCV) of 10 CFR 71.5 was identified for the licensees failure to ship radioactive material in accordance with Department of Transportation (DOT) requirements as specified in 49 CFR Parts 171-180. Specifically, upon receipt at its destination, a radioactive shipment classified as an excepted package for limited quantities was found to have external surface package dose rates exceeding the limit of 0.5 millirem per hour (mrem/h) as specified in 49 CFR 173.421)(a)(2). The package recipient identified a maximum dose rate of 3.95 mrem/h on the exterior surface of the package and notified the licensee of the discrepancy. The licensee entered the event into their corrective action program as Action Request (AR)- 01628106. The performance deficiency was more than minor because it was associated with the Program & Process Procedures attribute (DOT package limits) of the Public Radiation Safety Cornerstone. The inspectors determined the cornerstones objective was adversely affected based on the fact that shipment of radioactive material in excess of DOT limits in the public domain is contrary to NRC and DOT regulations. Assurance that the public will not receive unnecessary dose is decreased if packages are not prepared so that dose rates in accessible areas remain below regulatory limits during transit. The finding is of very low safety significance (Green) because there was little to no risk to members of the public. This finding involved the cross-cutting area of Human Performance with the aspect of conservative decision-making, in that the licensee assumptions failed to ensure that equipment packaged for shipment would not exceed DOT limits during transport.
05000335/FIN-2012004-02	2012Q3	Saint Lucie	Failure to Implement Procedure EN-AA-205, Design Change Packages	A self-revealing, non-cited violation (NCV) of Technical Specification (TS) 6.8.1, was identified which requires written procedures be established, implemented, and maintained covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, dated February 1978, including safety-related activities carried out during operation of the reactor plant. The licensees safety-related design control procedure EN-AA-205, Design Change Packages, was not implemented as written when a plant modification was performed on the reactor regulating system and steam bypass control system that affected a safety-related maintenance procedure that was not revised to reflect the design change. The licensee entered this violation in their corrective action program as action request 1786565. The licensees failure to fully implement procedure EN-AA-205, Design Change Packages, was a performance deficiency. The finding was determined to be more than minor because if left uncorrected, the deficiency could lead to a more significant safety concern. The inspectors evaluated the risk of this finding under the initiating events cornerstone using IMC 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined that the finding was of very low safety significance because it did not require a quantitative assessment as determined in Checklist 1. The finding involved a crosscutting aspect of complete and accurate procedures in the resources component of the human performance area (H.2.(c)). Specifically, the licensee failed to ensure that an adequate maintenance procedure was up to date to prevent an unexpected reactor plant temperature transient.
05000335/FIN-2012004-03	2012Q3	Saint Lucie	Licensee-Identified Violation	10 CFR 20.1501(a)(2) requires licensees to make or cause to be made surveys that are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels, concentrations or quantities of radioactive material, and the potential radiological hazards. Furthermore, 10 CFR 20.1003 defines a survey as an evaluation of the radiological conditions and potential hazards incident to the presence of radioactive material. Contrary to the above, on August 12, 2012, the licensee performed Tri-Nuc filter (vacuum) maintenance activities in the Unit 2 containment lower cavity without an adequate evaluation of the potential for the contamination to disperse and impact workers performing maintenance activities in the upper cavity area and containment. Specifically, the workers within the lower cavity who were supplied with bubble hood respiratory protective equipment disturbed elevated levels of alpha contamination within the lower cavity while moving their tangled respirator lines. Dispersion of these contaminants to the upper cavity area and containment was exacerbated by operation of the containment coolers and purge exhaust. The dispersed contamination resulted in unanticipated elevated airborne concentrations of radionuclides in the upper cavity and containment with subsequent intakes by workers involved with polar crane operation and upper cavity reactor head maintenance activities. The elevated airborne levels were discovered approximately one hour after the start of the lower cavity work through the evaluation of routine air samples collected for the work in the upper cavity. Immediate corrective actions taken upon discovery included evacuation of the Unit 2 containment and whole-body counting of all potentially impacted workers. The whole-body count evaluations identified eight workers with potential intakes of radioactive materials. Detailed analyses of whole-body count data and air sample results to identify hard-todetect radionuclides (alpha-emitters) for the affected workers resulted in a maximum assigned committed effective dose equivalent (CEDE) of 24.4 millirem (mrem) to one individual. For the other seven individuals identified with positive intakes, licensee estimates of dose (CEDE) were less than 10 mrem. An apparent cause evaluation performed by the licensee determined the causes to be inadequate work practices and planning. The corrective actions were documented under AR 01793148. The violation was evaluated using the Occupational Radiation Safety Significance Determination Process and was determined to be of very low safety significance (Green) because this finding was not an over-exposure, did not have a substantial potential for over-exposure because of continuous air monitors (CAMs) that would have alarmed with increasing airborne levels, and the ability of the licensee to assess dose was not compromised.

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