Semantic search

This page or action requires JavaScript to work. Please enable JavaScript in your browser or use a browser where it is supported, so that functionality can be provided as requested. For further assistance, please have a look at the noscript help page.

Options | Search Clear all entries |

Condition

Printout selection

Options

Parameters [+]

limit:

The maximum number of results to return

offset:

The offset of the first result

link:

Show values as links

headers:

Display the headers/property names

mainlabel:

The label to give to the main page name

intro:

The text to display before the query results, if there are any

outro:

The text to display after the query results, if there are any

searchlabel:

Text for continuing the search

default:

The text to display if there are no query results

class:

An additional CSS class to set for the table

transpose:

Display table headers vertically and results horizontally

sep:

The separator between results

prefix:

Control display of namespace in printouts

Sort options

Delete

Add sorting condition

ResultCodeDebugExtra

To embed this query inline into a wiki page use the code below.

{{#ask: [[Category:Finding]] [[Inspector::Brendan Collins]]
 |?Quarter
 |?Site
 |?Title
 |?Description
 |format=table
 |limit=50
 |offset=0
 |sort=
 |order=asc
 |mainlabel=
 |searchlabel=[Findings List]
}}

Results 1 – 50 previous 50 20 50 100 250 500 next 50

JSON CSV RSS RDF

Query log

{
    "query_string": "[[Category:Finding]] [[Inspector::Brendan Collins]]",
    "query_source": "SMWSQLStore",
    "query_time": "0.0874",
    "from_cache": false
}

	Quarter	Site	Title	Description
05000261/FIN-2013005-01	2013Q4	Robinson	Inadequate Preparation for Cold Weather Conditions	The inspectors identified a Green non-cited violation (NCV) of Technical Specification 5.4.1 for the licensees failure to implement freeze protection requirements specified in station procedures. Specifically the inspectors found that the required temporary enclosures were not installed and work orders for freeze protection circuits were not repaired prior to November 1, 2013, in accordance with procedure OP-925, Cold Weather Operation. The licensee initiated CR 645333 and took immediate corrective actions to install the necessary enclosures and to verify the proper operation of freeze protection circuits for safety related and fire protection equipment. The licensees failure to implement freeze protection requirements as required by procedure OP-925 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external factors attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to implement the requirements of procedure OP-925 could limit the sites ability to detect, respond to, or mitigate the consequences of an accident. The finding was determined to be of very low safety significance (i.e. Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. More specifically, the site had not experienced freezing weather conditions of sufficient magnitude to challenge plant systems during this time period. The finding involved the cross-cutting area of Human Performance under the Work Control component in that the licensee failed to appropriately plan work activities by incorporating risk insights to ensure the activities required to prepare the plant for cold weather conditions were completed prior to the onset of cold weather.
05000261/FIN-2013005-02	2013Q4	Robinson	Transient Materials Not Removed from Containment Prior to Reactor Startup	The inspectors identified a Green non-cited violation of Technical Specification (TS) 5.4.1 for the failure to properly implement procedure PLP-006, Containment Vessel Inspection Closeout, prior to startup following RFO 28. The improper closeout resulted in various tools as well as bags of consumable items and debris left in containment that could impact the containment sump strainer following an accident. The licensee initiated CR 640903, removed the items identified by the inspectors, and re-performed procedure PLP-006, Containment Vessel Inspection/Closeout, to further identify materials that should have been previously removed. The failure to remove debris and various temporary materials as required by procedure PLP-006 was a performance deficiency. The performance deficiency was determined to be 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 the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability and availability of ECCS equipment would be degraded by the introduction of material in to the containment that would impact and reduce the available area on the recirculation sump strainer. The inspectors determined that this finding is of very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Tech Spec Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding had a cross-cutting aspect in the Work Practices component of the Human Performance area, because the licensee failed to ensure that supervisor and management oversight of procedure PLP-006 ensured that debris was removed as required during containment closeout prior to reactor startup.
05000261/FIN-2013005-03	2013Q4	Robinson	Unauthorized Entry Into a HRA	A self-revealing, Green, non-cited violation (NCV) of TS 5.7.1, High Radiation Area, was identified for an unauthorized entry into a High Radiation Area (HRA). Specifically, two workers entered the residual heat removal pump room without knowledge of current radiological conditions and without wearing the prescribed electronic dosimetry for the area. The licensee entered this issue into the Corrective Action Program as Nuclear Condition Report 524523 and took immediate corrective actions including restriction of the workers from access to the Radiologically Controlled Area. This finding was determined to be greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Human Performance, Work Practices because the HRA event was a direct result of inadequate pre-job briefings and a lack of self and peer checking on the part of the work crew.
05000261/FIN-2013007-07	2013Q2	Robinson	Questions Regarding License Basis Design Requirements for Degraded Voltage Relays	The team identified an unresolved item (URI) regarding the degraded voltage relays. Specifically, the effect of system and transient harmonics on proper operation of degraded voltage relays was not analyzed. The Robinson degraded voltage protection design features three ITE Type 27N relays for each 480V emergency bus E1 and E2, arranged in a two out of three tripping scheme. BBC Instruction Bulletin 7.4.1.7-7 states, the relay employs a peak voltage detector, and harmonic distortion on the AC waveform can have a noticeable effect on the relay operating point and the measuring instruments used to calibrate the relay. The bulletin also notes that the relay is available with an internal harmonic filter for applications where waveform distortion is a factor; however, harmonic filters are not installed on the degraded grid voltage relays based upon their model number and specification package. The inspectors questioned if persistent harmonics on the 480V system could cause the relays to fail to actuate at the set point specified in Technical Specifications 3.3.5, and if transient harmonics could cause the relays to spuriously reset during the time delay that occurs during an actual degraded voltage condition concurrent with a design basis accident. Persistent harmonics can be produced by factors external to the nuclear site or by internal phenomena. A typical internal source of harmonics at nuclear power plants is defects in rotating equipment. Persistent harmonics could cause dropout set point shift, and mask an actual degraded voltage condition. Transient harmonics could cause the relays to spuriously reset during an actual degraded voltage event, thereby delaying the protective function beyond the nominal value stipulated in Technical Specifications 3.3.5. The relay is susceptible to this type of mal-operation because it features an instantaneous voltage sensor that could reset in less than two cycles in the presence of harmonics, thereby reinitiating the relays internal timer. The licensee has entered this item into their corrective action program as NCR 601203. This issue is unresolved pending inspector consultation with NRC headquarters technical staff for clarification of license basis design requirements of degraded voltage relays to withstand the effects of harmonics. This issue is identified as URI 05000261/2013007-07, Questions Regarding License Basis Design Requirements for Degraded Voltage Relays.
05000261/FIN-2014002-01	2014Q1	Robinson	Failure to Adequately Critique Fire Brigade Drills	A Green NRC-Identified non-cited violation (NCV) of Facility Operating License DPR-23, Condition 3.E, Fire Protection Program, was identified for the licensees failure to identify, critique, and develop corrective actions for fire brigade performance weaknesses during two fire drills as required by procedure TPP-219, Fire Protection Training Program. Upon identification of these weaknesses by the inspectors, the licensee entered them into the corrective action program (CAP), performed an apparent cause evaluation, and revised procedure TPP-219 to further define the roles and responsibilities of the drill controllers as well as the standards used to critique the fire brigade. The licensees failure to identify, critique, and develop appropriate actions for fire brigade performance weaknesses during two fire drills as required by procedure TPP-219 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety Significance (Green) in accordance with question D.1 because although the finding involved fire brigade training requirements, the fire brigade demonstrated the ability to meet the required times for fire extinguishment for the fire drill scenarios and the finding did not significantly affect the fire brigades ability to respond to a fire. The performance deficiency had a cross-cutting aspect of Consistent Process in the area of Human Performance, because the licensee failed to use a consistent, systematic approach during conduct of fire brigade drills and during the subsequent critique process.
05000261/FIN-2014002-02	2014Q1	Robinson	Steam Generator Tube Leak Resulting from Foreign Material	A self-revealing Green FIN was identified for the licensees failure to thoroughly inspect and remove foreign material from feedwater piping after initial breach of the pipe, as required by licensee procedure MNT-NGGC-0007, Foreign Material Exclusion Program. As a result, foreign material entered the C Steam Generator (SG) and damaged a tube which created a primary-to-secondary leak condition. This finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone, and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, foreign material entered the SG and damaged a SG tube, which increased the likelihood of a SG tube rupture (SGTR) and challenged the reactor coolant system (RCS) integrity safety function during shutdown. The inspectors used IMC 0609, Significance Determination Process, Attachment 0609.04, issued June 19, 2012, Initial Characterization of Findings, and Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, and determined that the finding was of low safety significance (Green) because testing showed that the affected SG tube could sustain three times the differential pressure across the tube during normal full power and that the SG did not violate the accident leakage performance criterion. The performance deficiency had a cross-cutting aspect of Challenge the Unknown in the area of Human Performance because the licensee did not stop when faced with the unknown or evaluate and manage risk before proceeding. Specifically, the licensee should have evaluated and addressed the FME issue resulting from the pipe spring condition during the initial breach of the feedwater piping before continuing.
05000261/FIN-2014002-03	2014Q1	Robinson	Inadequate Preventive Maintenance on 4 KV Breaker 52/7 Results in an Automatic Reactor Trip	A self-revealing Green finding (FIN) was identified for the licensees failure to perform adequate preventive maintenance (PM) in accordance with, licensee procedure ADM-NGGC-107, Equipment Reliability Process, for 4 KV Breaker 52/7, Unit Auxiliary to 4 KV Bus 1. As a result, while transferring loads from the start-up transformer, a broken operating rod for breaker 52/7 prevented the breaker from closing and caused an automatic reactor trip. The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance, and it adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in breaker 52/7 failing to close and subsequently causing an automatic reactor trip from 19 percent power operations on November 5, 2013. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would not be available. The performance deficiency had a crosscutting aspect of Resolution in the area of Problem Identification and Resolution, because the licensee failed to take effective corrective actions to address a similar failure of an operating rod for the A circulating water (CW) pump breaker in 2011.
05000261/FIN-2014002-04	2014Q1	Robinson	Failure to Provide Adequate Design Control Measures for Diesel Fuel Oil Cloud Point	The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to provide adequate design control measures to ensure appropriate specifications were translated into procedures for diesel fuel oil (DFO) to ensure that the DFO temperatures remained above the DFO cloud point. The licensee entered this into the CAP as action request (AR) 664223 and took immediate corrective actions to change the cloud point acceptance criteria from 23 degrees to 10 degrees Fahrenheit and revise procedure OP-925, Cold Weather, to install temporary heaters if outside temperatures fell below 15 degrees Fahrenheit. The licensees failure to provide design control measures to ensure that the DFO temperature was maintained such that the cloud point was not reached was a performance deficiency. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, during periods of cold weather the DFO temperature could have been allowed to fall below its cloud point and affect operation of the emergency diesel generator (EDG) and/or the dedicated shutdown diesel generator operation due to the DFO transfer system becoming inoperable. The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, dated June 19, 2012, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding is a deficiency affecting the design or qualification of a mitigating SSC; however, the SSC maintained its operability or functionality since the design conditions were not actually reached. The performance deficiency had a cross-cutting aspect of Design Margins in the area of Human Performance because the licensee failed to recognize that additional actions were required to maintain operability of the DFO system when ambient temperatures are below the maximum administrative limit even though samples are reviewed monthly per the DFO Testing Program.
05000261/FIN-2014002-05	2014Q1	Robinson	Defective Motor Operated Potentiometer causes failure of the DSDG during surveillance testing	An URI was identified regarding the trip of the DSDG, on December 31, 2013, during monthly surveillance testing. The URI is being opened to provide for additional inspection of the equipment issues that led to the failure and to review the results of the vendors analysis of a defective motor operated potentiometer to determine if a performance deficiency exists. On December 31, 2013, during monthly testing of the DSDG in accordance with licensee procedure OST-910, Dedicated Shutdown Diesel Generator (Monthly), the output breaker tripped open on overcurrent while the operators were attempting to adjust DSDG output voltage. Operators in the field noted erratic voltage indication prior to the failure. Engineering identified that the likely cause was a failure of the motor operated potentiometer (MOP). The licensee replaced the MOP with a new part from stock and performed post maintenance testing. The MOP that was removed was sent offsite for forensic analysis. During examination, the licensee identified a manufacturing defect for the MOP. The licensees extent of condition investigation found the same manufacturing defect on the MOP installed in the DSDG and in a MOP in storage. The licensee replaced the MOP in the DSDG with a MOP that was verified to be acceptable. Engineering has sent the defective components back to the vendor for additional analysis. This issue will be identified as URI 05000261/2014002-05; Defective Motor Operated Potentiometer causes failure of the DSDG during surveillance testing.
05000269/FIN-2008003-01	2008Q2	Oconee	Required Pressure Tests Not Performed	The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.55a(g)(4) for the failure to perform periodic leakage testing of Class 1 portions of low pressure injection system during the third inspection interval of the Unit 1 Inservice Inspection Program as required by Section XI of the American Society of Mechanical Engineers (ASME) Code for the third 10-year Inservice Inspection interval. The licensee entered this issue into their Corrective Action Program (CAP) for resolution. This finding is more than minor because it affected the Equipment Performance attribute of the Mitigating Systems Cornerstone objective, in that there were no additional measures taken to perform the required pressure testing (or obtain regulatory relief within the time limits of the regulation) to ensure the availability, reliability, and capability of a system that responds to initiating events to prevent undesirable consequences. This finding is of very low safety significance because it was not a design issue resulting in a loss of operability, did not represent an actual loss of a systems safety function, did not result in exceeding a Technical Specification (TS) allowed outage time, and did not affect external event mitigation. This finding has a cross-cutting aspect of Work Control in the area of Human Performance, as identified in NRC Manual Chapter 0305, Section 06.07 (H.3.(b)). (Section 1R08.1
05000269/FIN-2008003-02	2008Q2	Oconee	AVR Maintenance Procedure Resulted in a Loss of Inventory While on Decay Heat Removal	On April 15, 2008, Unit 1 experienced a generator lockout, which caused a loss of normal power through the back charged main transformer to main feeder bus (i.e., N breakers and generator output breakers PCB 21 and 20 opened). This caused a momentary loss of power to the LPI and low pressure service water (LPSW) pumps, and AP-26 (Loss of Decay Heat Removal) was entered by the operators. The main feeder bus (MFB) power was restored in 1.8 seconds via the slow transfer logic from the E breakers (as expected) and power to the LPI and LPSW pumps was restored (non-load shed). MFB re-energization should have re-energized motor control center (MCC) 1XP, but did not, as the alternate feeder breaker to 1XP tripped on high in-rush current. MCC 1XP supplies power to air operated valves 1HP-8 (Unit 1 Purification Demineralizer Inlet Isolation), 1HP-17 (1A Letdown Filter Inlet Isolation) and 1HP-18 (1B Letdown Filter Inlet Isolation). These valves closed due to the loss of power to each valves respective control solenoid valve, which isolated the letdown flow path. The valves should have reopened upon the restoration of power, but the loss of power to MCC 1XP prevented this from occurring. The isolation of the letdown flow path coupled with the LPI pump restart caused 1HP-43 (letdown line relief valve) to lift open as the relief valve now saw LPI pump discharge pressure. This resulted in a loss of RCS inventory to the Mixed Waste Holdup Tank (MWHUT) Operators initiated makeup to the LPI system from the borated water storage tank (BWST) by throttling 1LP-21 (BWST isolation) and 1LP-96 (LP Supply to Purification IX block) was closed to isolate LPI from the purification lineup per AP-26. This allowed 1HP-43 to shut and effectively stop the inventory loss. Specifically, RCS level dropped from 70 to 54 inches on LT-5 (approximately 2000 gallons of inventory was dumped to MWHUT). Level was restored to 79 inches on LT-5 approximately 30 minutes from initiation of the event. The generator lockout was a result of ongoing automatic voltage regulator (AVR) preventive maintenance, IP/0/B/2005/001, Main Generator Automatic Voltage Regulator Maintenance and Channel Transfer, where as part of the AVR procedure, the AVRs measuring unit board (part of each AVR channel) was replaced and powered up. Unrecognized by the procedure, this power up resulted in the actuation of the K31 relay in the AVR, which sends a lockout signal to the main generator. The root cause was determined to be a failure of procedure preparers and reviewers of IP/0/B/2005/001, to recognize the system interaction between the AVR trip circuitry and the backcharge power path. This issue is unresolved pending further NRC review of the licensees procedures, operator actions, and risk management associated with this event. This item is identified as URI 05000269/2008003-02, AVR Maintenance Procedure Resulted in a Loss of Inventory While on Decay Heat Removal. This issue is in the licensees corrective action program as PIP O-08-2056.
05000269/FIN-2008003-03	2008Q2	Oconee	Licensee-Identified Violation	NRC Order EA 03-009 Paragraph C. (5).E requires for each inspection of the reactor vessel head required in the order, the licensee shall submit a report detailing the inspection results within 60 days after returning the plant to operation. Contrary to these requirements, the licensee did not submit a 60 day report for Unit 3 head inspections conducted during the last refueling outage, which ended December 23, 2007. This finding is determined to be of very low safety significance because the deficiency was identified, examinations that met the requirements of NRC Order 03- 009 were performed, and the report was subsequently submitted on April 22, 2008. The licensee entered the finding into their corrective action program as PIP 08- 01635
05000269/FIN-2010005-04	2010Q4	Oconee	Potential Inoperability of the Unit 3 Standby Shutdown Facility Reactor Coolant Makeup Pump	During the performance of a quarterly In-service Testing surveillance test of the Unit 3 SSF RCM pump on August 24, 2010, the licensee observed an increase in the normal reactor building sump level and subsequently found to have been caused by seat leakage in relief valve 3HP-404. It was identified that while leakage during the test was less than 0.5 gpm, a gradual increase in leakage through the relief valve had occurred during tests performed over the preceding 13 months. The SSF RCM Pump was used to provide makeup to the reactor coolant system (RCS) and RCP seal cooling during an SSF-design basis event. Leakage through the relief valve would reduce the amount of water that would reach the RCS and could result in RCP seal failure or the inability of operators to control RCS inventory if the relieve valve leakage continued to increase. The licensee will conduct testing to determine if the leakage rate would degrade enough over the 72-hour mission time of the SSF to prevent the SSF RCM pump from performing its safety function. The licensee has replaced the relief valve. This issue is identified as URI 05000287/2010005-04: Potential Inoperability of the Unit 3 Standby Shutdown Facility Reactor Coolant Makeup Pump.
05000269/FIN-2010007-02	2010Q2	Oconee	Materially Inaccurate Information Provided to NRC Regarding SSF Event Mitigation Capability	A licensee-identified apparent violation of 10 CFR 50.9(a) was identified when the licensee determined that information contained in the Oconee Nuclear Station SSF RC Letdown Action Plan was inaccurate. This information was material to NRC because it was used, in part, as the basis for determining whether the licensees response to the degraded condition was adequate and whether additional compensatory actions or NRC review would be necessary. This apparent violation has been entered into the corrective action program as PIP O-10-0561. The failure to provide complete and accurate information impacted the regulatory process in that the inaccurate information was material to NRCs determination that the licensees response to the degraded condition was adequate. The inspectors determined the severity level of this apparent violation is potentially greater than Severity Level IV. Cross-cutting aspects are not assigned to violations being dispositioned through the traditional enforcement process.
05000269/FIN-2011002-02	2011Q1	Oconee	Inadequate Post Modification Testing to Ensure SSF DG Functionality	An NRC-identified Apparent Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to ensure that a modification installed on the Safe Shutdown Facility (SSF) Diesel Generator (DG) monitoring panel would not affect the ability of the SSF Power subsystem to perform its design function. The finding does not represent an immediate safety concern because the chart recorder was modified so that it did not send an output signal to the SSF control and protection logic circuit. The licensees failure to ensure the post-modification testing was adequate to verify the modification did not affect the SSF Power subsystems ability to perform its design function was a performance deficiency (PD). The PD was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and adversely impacted the cornerstone objective in that the modification would have prevented the SSF DG from starting and supplying power to the SSF. The safety significance of this finding was To Be Determined pending completion of a Phase III risk analysis. The finding was directly related to the cross-cutting area of Human Performance under the Procedural Compliance aspect of the Work Practices component because the licensee failed to ensure station modification program requirements were followed in the development of post-modification testing.
05000269/FIN-2011003-01	2011Q2	Oconee	Inadequate Design Verification of the NPBS-BWST/SSF Trench Foundation	A self-revealing finding was identified for the licensees failure to implement the requirements of the modification program to ensure the natural phenomenon barrier system (NPBS) borated water storage tank (BWST)/standby shutdown facility (SSF) trench foundation modification did not adversely impact the yard drain systems function. The condition was entered into the licensees corrective action program (CAP) as problem investigation program (PIP) O-11-3285. The failure to implement the requirements of the modification program to verify the NPBS BWST/SSF trench foundation modification did not adversely impact flood protection features was a performance deficiency (PD). The PD was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors - Flood Hazard and adversely affected the cornerstone objective in that the design modification bypassed the yard drain system which was credited for external flood protection. The finding was of very low safety significance (Green) because the rainwater intrusion did not result in the loss of operability or functionality of safety-related structures, systems, and components (SSCs). The cause of the finding was directly related to the appropriately planning work activities cross-cutting aspect of the Work Control component in the area of Human Performance because the licensee failed to incorporate environmental conditions which may impact SSCs into the modification. (H.3(a)) (Section 1R01
05000269/FIN-2011003-02	2011Q2	Oconee	Follow-up of NOED 11-2-03	An unresolved item (URI) was identified for NOED 11-2-03. On January 8, 2011, during Unit 1 ES digital channel 2 testing, valve 1HP-5 failed to fully close following an inadvertent close signal. A root cause investigation found that improper material selection for the gland ring in a 2003/2004 modification resulted in a loss of margin for the actuator for the valve, as well as similar valves on ONS Unit 2 and Unit 3. On May 31, 2011, discussions with valve vendor identified that the software used in the licensees operability determination for the Unit 2 and Unit 3 valves may calculate non-conservative torque values resulting in a reduction in the valves closing margin. Ongoing evaluations indicated the required spring closing forces were inadequate resulting in a negative closing margin at normal reactor coolant system pressures. On June 2, at approximately 12:10 p.m., containment isolation valves 2HP-5, 2HP-21, 3HP-5, and 3HP-21 were declared inoperable. The licensee determined that placing Unit 2 or Unit 3 in MODE 5 would result in power delivery challenges during a period of significant grid demand and could result in adverse consequences to the health and safety of the public. On this basis the licensee requested a severe weather NOED, on June 2, 2011, for a period of 14 days or until electrical grid conditions are predicted to return to normal for a period of at least 10 days to ensure the safe and orderly shutdown of one or both ONS units to restore compliance with TS 3.6.3. An independent assessment by NRC staff with consultation with SERC Reliability Corporation confirmed the licensees assertion of the grid stability conditions. The NRC verbally granted the NOED at 6:30 p.m., on June 2. The licensee returned 2HP-5, 2HP-21, 3HP-5, and 3HP-21 to operable status after performing a modification to add additional closing force to regain necessary margin on June 11, 2011. Additional inspection is required to conduct a review of the LER, root cause, and planned corrective actions. This is identified as URI 05000270, 287/2011003-02, Follow-up of NOED 11-2- 03
05000269/FIN-2011005-01	2011Q4	Oconee	Failure to Perform Adequate Surveys to Identify Potential Radiological Hazards	A self-revealing, non-cited violation (NCV) of 10 CFR 20.1501(a) was identified for failure to perform adequate surveys to verify radiological conditions within the Unit 3 Reactor Building (RB). This resulted in a worker unknowingly entering an area with dose rates exceeding Locked High Radiation Area (LHRA) conditions, i.e., dose rates exceeding 1,000 millirem per hour (mrem/hr) at 30 centimeters (cm). Corrective actions included surveying all plant areas for proper posting and control in which no additional problem areas were identified, reviewing jobs that had the potential for dose rate changes, and reviewing electronic dosimeter (ED) trends during each shift. The inspectors determined that the failure to identify the LHRA through adequate surveys that could have revealed changing radiological conditions was a performance deficiency. This performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Program and Process (Monitoring and RP Controls) and adversely affected the cornerstone objective in that failure to identify significant sources of radiation could lead to unintended occupational exposures. The finding was determined to be of very low safety significance (Green) because it was not related to As Low As is Reasonably Achievable (ALARA) Planning and the ability to assess dose was not compromised. The finding was directly related to the cross-cutting aspect of Appropriate Coordination of Work Activities in the Work Control component of the Human Performance area because the licensee failed to identify the change in radiological conditions.
05000269/FIN-2011005-02	2011Q4	Oconee	Licensee-Identified Violation	TS 5.4.1.a required that procedures defined in RG 1.33 shall be established and implemented. RG 1.33, Appendix A, stated, in part, that written procedures for control of radioactivity shall be developed and implemented. Section 4.2.1 to Procedure No. SH/0/B/2000/008, Operational Alpha Program, Rev. 7, stated, in part, that if smears over 20,000 dpm/100 cm2 beta-gamma are identified, an analysis of the smear for alpha be performed. Contrary to the above, on May 12, 2010, a smear conducted on an area of a Reactor Coolant Storage System pipe had 80,000 dpm/100 cm2 beta-gamma and the licensee did not analyze it for alpha. The finding was not greater than very low safety significance (Green) because it did not involve an overexposure and the licensees ability to assess dose was not compromised. The licensee entered the violation into their CAP as PIP O-10-03822.
05000269/FIN-2012002-01	2012Q1	Oconee	Evaluation of Probable Maximum Flood Event	During a walkdown of Manhole 7 on February 1, 2012, inspectors noted that two conduit penetrations used to route PSW cabling into the AB were not sealed and provided a direct flooding pathway into the AB. These penetrations were identified as requiring seals whenever not being used for cable pulls or sealed immediately following cable pulling activities. Flooding from these penetrations would exceed the capacity of the AB sump pumps and fill the high pressure injection (HPI), low pressure injection (LPI) and reactor building spray (RBS) pump rooms rendering the pumps inoperable. The inspectors also identified that a field change rerouted the internal drainage system from the yard drain system to the adjacent radwaste trench. Rainwater accumulating in Manhole 7 would flow through the internal drains to the radwaste trench and into the AB through a non-isolable line which drained into the low activity waste tanks. These tanks would eventually overflow flooding the HPI, LPI and RBS pump rooms rendering the pumps inoperable. The design change in the original design package for Manhole 7 and the field change for rerouting the drain did not evaluate the impact they would have on the AB features to mitigate external floods. Consequently, the currently described Updated Final Safety Analysis Report (UFSAR) described PMF event would result in rendering safety-related/risk significant equipment inoperable. Additionally, the licensee recently completed a site inundation study which projected site water levels to be greater than the maximum flood protection measures for a PMF event as described in UFSAR Section 3.4.1.1 even with a fully functional yard drain system. Changes in site topography and construction of new buildings since initial construction appear to be contributing to the increased water levels. The licensee is currently evaluating the impact of the new site inundation level and has implemented interim actions to provide protection from increased water levels on-site. The NRC will perform additional inspection to ensure the impact to the AB from a PMF event is understood, an accurate timeline on Manhole 7 construction activities is developed, and that the extent of condition is fully defined. This issue is identified as URI 05000269, 270, 287/2012002-01, Evaluation of Probable Maximum Flood Event.
05000269/FIN-2012003-01	2012Q2	Oconee	Failure to Perform a Calculation to Determine Site PMP Ponding Levels in a Timely Manner	An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to implement corrective actions for a condition adverse to quality. The licensee did not develop a calculation to determine the maximum on-site water level resulting from a Probable Maximum Precipitation (PMP) event in a timely manner. Corrective actions included development of a calculation bounding the expected water level resulting from a PMP event. This violation is in the licensees corrective action program (CAP) as PIP O-12-7994. The performance deficiency (PD) was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control and adversely impacted the cornerstone objective because there was reasonable doubt that plant equipment was adequately protected from the increased water level and therefore had the potential to result in a loss of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the licensee subsequently demonstrated that the water entering the plant structures would not have resulted in the loss of safety-related or risk-significant equipment. This finding does not have a cross-cutting aspect because the performance deficiency was not indicative of current plant performance.
05000269/FIN-2012003-02	2012Q2	Oconee	Failure to Follow the Engineering Change Process	An NRC-identified non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to follow EDM 601, Engineering Change Manual, during the design and construction of the Protected Service Water (PSW) ductbank / manhole structure. As a result, rainwater accumulation during a Probable Maximum Precipitation (PMP) event could enter the Auxiliary Building (AB). Corrective actions included sealing penetrations, installation of an isolation valve, revising procedures, and conducting training. This violation is in the licensees CAP as PIPs O-12-1317, O-12-1876, O-12-1906 and O-12-2443. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors - Flooding and adversely affected the cornerstone objective in that water from a PMP event could enter the AB and adversely impact safety-related and / or risk-significant equipment. The licensee was required to perform extensive modeling and calculations to determine what the impact from a PMP event would be on the SSCs located in the lower elevations of the AB. The finding was of very low safety significance due to the high likelihood that the source of the water leaking into the AB would be correctly identified and isolated prior to the loss of safety-related equipment due to the flood. The cause of the finding was directly related to the aspect of ensuring supervisory oversight of work activities such that nuclear safety is supported of the Work Practices component in the cross-cutting area of Human Performance because the licensee failed to ensure that the appropriate level of supervisory and management oversight was applied during design, modification and construction of Manhole 7.
05000269/FIN-2012003-03	2012Q2	Oconee	Licensee-Identified Violation	10 CFR 50, Appendix B, Criterion III, Design Control, stated, in part, that measures shall be established to assure that applicable regulatory requirements and the plants design basis are correctly translated into drawings, procedures, and instructions and that these measures shall provide for verifying or checking the adequacy of the design. Contrary to the above, from initial operation through March 24, 2012, the licensee did not provide measures for verifying or checking the adequacy of design changes to the site topography through grading and erection of structures. These changes were not evaluated to ensure that they did not adversely impact safety related SSCs due to water intrusion into the AB from PMP events. This violation was determined to be of very low safety significance (Green) because the loss of this equipment or function by itself, during the external initiating event it was intended to mitigate would not cause a plant trip or any of the Initiating Events used by Phase 2, would not degrade two or more trains of a multi-train safety system or function, and would not degrade one or more trains of a system that supports a safety system or function. The licensees evaluation demonstrated that water entering the Auxiliary Building could be removed through the use of installed plant equipment and the use of additional equipment, procedural guidance, and directed actions. The licensee entered this violation into their CAP as PIP O-12-4318. This LIV addresses a portion of URI 05000269, 270, 287/2012002-01.
05000269/FIN-2012003-04	2012Q2	Oconee	Licensee-Identified Violation	10 CFR 50, Appendix B, Criterion XVI, Corrective Action, stated, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and defective material are promptly identified and corrected. Contrary to the above, from 2006 to 2012, a condition adverse to quality was not promptly identified and corrected. In 2006 ONS completed a corrective action to review all potential AB flooding sources and failed to identify the Coolant Storage (CS) and RCW system as a flooding source that could impact safety related equipment. In 2012 it was determined that if the non-seismically qualified Coolant Storage (CS) system or RCW System, while cross-connected between Unit 1 / 2 and Unit 3, were to fail as a result of a seismic event, safety related SSCs in the AB could be adversely affected. The violation was determined to be of very low safety significance (Green), based on the licensees determination that even though there were sections of the CS system that were not seismically qualified, the non-seismic CS piping was robust enough to withstand a seismic event. A review of the volume of water that would have been released following a break of the RCW piping if cross connected between the units determined that based on actual tank levels in the AB which would collect water released from a break when the systems had been crosstied in the past that the Unit 3 HPI pumps would not have been affected. In addition, the volume of the Unit 1 / Unit 2 HPI pump room was large enough to preclude any impact following a break even with the RCW systems cross-connected. The licensee entered this issue into their CAP as PIP O-12-1876.
05000269/FIN-2012003-05	2012Q2	Oconee	Licensee-Identified Violation	10 CFR Part 50, Appendix B, Criterion III, Design Control, stated, in part, that measures shall be established to assure that deviations from appropriate quality and design standards are controlled and that the review for suitability of application of equipment essential to safety-related functions of SSCs is maintained. Contrary to the above, from 2005 until August 11, 2011, a review for suitability of application of equipment essential to safety-related functions of structures, systems, and components had not been performed. The licensee failed to ensure that the pressurizer heaters would perform their function following a loss-of-offsite-power (LOOP) event. As a result, the total available emergency-powered heater capacity during a LOOP did not comply with Technical Specification 3.4.9 for Units 1 and 2. This condition existed since March 2005 which resulted in the TS 3.4.9, Action Condition C, 72-hour completion time being exceeded. This violation was determined to be of very low safety significance (Green) because analysis showed that it would take over five days for the east penetration room to reach a temperature that would trip the supply breakers and station procedures dictate that the unit enter a Mode that would not require pressurizer heaters if offsite power could not be recovered prior to five days. The licensee entered this issue into their corrective action program as PIPs O-11-8094, O-11-6700, and O-12-2655.
05000269/FIN-2013005-01	2013Q4	Oconee	Failure to properly maintain a fire barrier penetration seal	An NRC-identified non-cited violation (NCV) of 10 CFR 50.48(c) and National Fire Protection Association Standard 805 (NFPA 805), Section 3.11.4, was identified for the licensees failure to comply with the fire barrier penetration sealing and inspection requirements of the approved fire protection program (FPP). The annular space between the fire barrier opening and the 2 conduit was not properly sealed. The licensee entered the issue in their CAP as PIP O-13-09104, initiated a work order to repair the seal, and implemented an hourly fire watch as required by Oconee Selected Licensee Commitment (SLC) 16.9.5. The licensees failure to comply with the fire barrier penetration sealing and inspection requirements of the approved fire protection program was a performance deficiency. This performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and adversely affected the cornerstone in that the fire barrier could not be relied upon to fully perform its function. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of very low safety significance (Green) because safety significant equipment was located a sufficient distance from the degraded penetration and the reactors ability to reach and maintain a safe shutdown condition was not impacted. The cause of this finding was determined to have a cross-cutting aspect of H.2(c) in the Resources component of the Human Performance area because the licensee did not ensure that complete, accurate, and up-to-date design documentation and procedures were available because adequate guidance was not included in the maintenance inspection procedures to allow personnel to identify a degraded condition.
05000269/FIN-2013005-02	2013Q4	Oconee	Licensee-Identified Violation	10 CFR 50, App. B, Criterion XVI, required in part that conditions adverse to quality, such as non-conformances, are promptly identified and corrected. NSD-203, Operability/Functionality, required entry into the operability determination process (ODP) upon the discovery of circumstances that call into question the operability of any TS SSC including degraded/non-conforming conditions. NSD-203 also requires that actions to confirm if the SSC is degraded or non-conforming should be completed in a timeframe that is commensurate with its safety significance. Contrary to the above, a potential non-conforming condition was identified on December 30, 2012; however, the ODP was not entered until November 26, 2013, and corrective actions generated to correct the non-conforming condition. The finding was not greater than Green because it 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 an actual reduction of hydrogen igniters in containment. This violation was entered into the CAP as PIP O-13-14547.
05000269/FIN-2014005-01	2014Q4	Oconee	Failure to Update FSAR for Mode 4 LOCA	An NRC identified Severity Level IV violation of 10 CFR 50.71(e), "Maintenance of Records, Making of Reports," was identified for the licensees failure to update the final safety analysis report (FSAR) after the licensee adopted the improved technical specifications (ITS). The licensee adoption of ITS introduced the possibility of a Mode 4 loss of cooling accident (LOCA), which was an accident of a different type than previously evaluated in the FSAR. The licensee initiated PIP O-15-00260 in order to determine future corrective actions. Continued non-compliance does not present an immediate safety concern because the inspectors assessed this as a very low safety significant issue. The licensees failure to update the FSAR as required by 10 CFR 50.71(e) was a performance deficiency. The performance deficiency impacted the ability of the NRC to perform its regulatory oversight function and was dispositioned using traditional enforcement. Specifically, a failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating license amendment requests, and in preparation for and conduct of inspection activities. This violation was determined to be a Severity Level IV violation per Section 6.1.d.3 of the NRC Enforcement Policy, revised July 9, 2013, because the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. The NRC Enforcement Policy also requires disposition of findings in the significance determination process, which determined the finding was not more than minor. Since this issue was dispositioned using traditional enforcement, there was no cross-cutting aspect associated with this violation.
05000269/FIN-2014005-02	2014Q4	Oconee	Keowee Hydro Unit 2 Inoperable for Longer Than Allowed TS Outage Time	A self-revealing Green NCV of Oconee Nuclear Station Technical Specification (TS) 3.8.1, AC Sources Operating, was identified for Keowee Hydro Unit 2 being inoperable for longer than allowed TS outage time. The licensee modified Keowee Hydro Unit 2 electrical protection circuitry with a faster response relay which was susceptible to an existing degraded system condition and ultimately caused Keowee Hydro Unit 2 to be inoperable. The licensee implemented engineering change (EC111358) which moved the 86E2X relay to another cabinet which was not susceptible to the vibration from the governor oil system. The licensee entered this issue in their corrective action program (CAP) as PIP-O-13-09152. The licensees failure to properly evaluate a modification to the electrical control circuit of the governor oil system, which resulted in Keowee Hydro Unit 2 being inoperable for longer than allowed TS outage time, was a performance deficiency. The issue is more than minor because it was associated with the equipment performance attribute of the mitigating system cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the modification of the governor oil system, including the addition of the 86E2X governor TXS catastrophic relay, resulted in Keowee Hydro Unit 2 being inoperable for longer than allowed TS outage time. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process (SDP), Attachment 4 and Attachment A and determined to require a detailed risk evaluation. A regional Senior Reactor Analyst performed a risk analysis of the performance deficiency which was found to be Green (CDF < 1E-6/year). The dominant accident sequence was a loss of offsite power where Keowee Unit 1 fails independently and unrelated to the performance deficiency and power is not successfully restored by Oconee operators. The influential factors in the Green result were the limited exposure time (19 days) and the ability to quickly restore power to the unit via the Lee Station gas turbines via the Fant Line. This finding was determined to have a cross-cutting aspect in the problem identification and resolution cross cutting area because the licensees organization failed to take effective corrective actions to address the issue in a timely manner commensurate with its safety significance. Specifically, the licensee failed to take effective corrective actions to address system interactions (i.e. high vibrations) which ultimately had an adverse effect upon modifications to the governor oil system of the Keowee Hydro Unit 2.
05000269/FIN-2014007-05	2014Q2	Oconee	Potential Unanalyzed Condition Associated with Emergency Power System	During a review of Oconees engineered safeguards protection system (ESPS) emergency power start control for the KHUs, the team noted that the 125Vdc control cables for train A of the ESPS and cables for supervisory control of both KHUs were recently modified. The team also noted that these 125Vdc control cables were installed in the same underground concrete raceway systems as the 4160Vac auxiliary power cables, 13.8kVac power cables for both emergency power and protected service water (PSW), and were in close proximity to these power cables. The team was concerned that a short circuit (which the licensee considered outside their design basis) in the 13.8kVac cables could induce voltage and currents in the dc control system which could potentially impact the functionality of the emergency power system which is required to mitigate certain design basis events. A similar issue exists in Manhole 6 of the PSW underground raceway where the new power supply to the PSW (adjacent to the 125Vdc control emergency power system) could short circuit or fault to ground. The licensee had not performed an analysis to determine the effects of such failures on the ability of the emergency power system to perform its safety function, thus the team questioned whether the plant was in an unanalyzed condition. Although the licensee did not agree that these failures were part of their licensing basis, they reported this as an unanalyzed condition to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B) in Licensee Event Report 269/2014-01. In response to the teams concerns, the licensee entered this issue into their corrective action program, and performed immediate and prompt determinations of operability in which they concluded a reasonable expectation of operability exists. The team has requested assistance from subject matter experts in the Office of Nuclear Reactor Regulation via a Task Interface Agreement1 to review the emergency power system licensing basis to determine the acceptability of the licensees design. If the design is found to be noncompliant with the licensing basis, the licensee will be required to implement corrective actions to restore compliance. This issue is being tracked as URI 05000269/2014007-05, 05000270/2014007-05, 05000287/2014007-05, Potential Unanalyzed Condition Associated with Emergency Power System.
05000269/FIN-2016004-01	2016Q4	Oconee	Failure to Perform Appropriate Evaluation of Motor Operated Valve Actuator Output Capability	Green. The NRC identified a non-cited violation (NCV) of Title 10 Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly determine the bounding degraded voltage to be assumed in the determination of motor operated valve (MOV) actuator output capability. Specifically, the licensee did not use appropriate transient voltages as input into the evaluation of the capability of the MOVs that are required to reposition in response to an accident signal. In response, the licensee entered the issue into their corrective action program as nuclear condition report (NCR) 2056895 and planned to formally revise their calculations to reflect the current plant configuration. This performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees programmatic failure to use bounding terminal voltage values in the evaluation of their automatically actuated, safety-related MOVs did not ensure they would be capable of mitigating accidents when powered from sources other than the 230kV switchyard, thus resulting in doubt on their capability to perform their intended safety function. 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 or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the most recent transient analysis that was performed for the sources other than the 230kV switchyard was performed in 2012.
05000269/FIN-2016004-02	2016Q4	Oconee	Inappropriate Voltage Band in Lee Combustion Turbine Unit Operating Procedure	Green. The NRC identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to identify appropriate procedural updates that were needed to ensure the Lee combustion turbine (LCT) procedures were appropriate for the circumstances and maintained current. Specifically, the licensee did not include appropriate operational limitations in procedures associated with the LCTs. In response, the licensee generated NCR 2058763, verified the LCT automatic voltage regulator setpoint was, and had been, 13.8kV, and generated a corrective action to revise the affected procedures limits to 13.78kV, a value bounded by station analyses. This performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees failure to limit the operating voltage band of the LCTs to an amount that was demonstrated as acceptable by analysis resulted in doubt on their capability to provide power to safety-related equipment during an accident. 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 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 update to the procedures occurred in January and October 2007, after replacement of the LCTs.
05000269/FIN-2017004-01	2017Q4	Oconee	Failure to Identify Sensitive Equipment During Modification Results in Loss of Safety Function	A self-revealing Green non-cited violation (NCV) of Oconee Nuclear Station Technical Specification (TS), Section 5.4, Procedures, was identified for the licensees failure to identify sensitive equipment in a work area that warranted implementation of compensatory measures as required by station procedure AD-EG-ALL-1180, Engineering Change (EC) Walkdowns. During the design and planning phase of a station modification, the licensee failed to identify sensitive components located in the subject work area and subsequently failed to implement adequate protective measures as defined in station procedures to prevent plant impacts during modification installation. The licensee entered this issue into their corrective action program (CAP) as nuclear condition report (NCR)02131608 and implemented corrective actions to identify other positionable components required for emergency power source operability that would require the use of protective measures, as defined by AD-OP-ALL-0204, Plant Status Control, in order to prevent inadvertent operation. The licensee created a formal Engineering department communication which included lessons learned from the event and familiarization with the EC walkdown checklist. The signs on the governor actuator cabinets were also revised to emphasize the sensitive nature of the equipment. The licensees failure to properly identify sensitive equipment and implement compensatory measures to prevent plant impacts as required by station procedure AD-EG-ALL-1180 was a performance deficiency. 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the loss of the emergency AC power path function for 11 hours and 31 minutes. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. Inspection Manual Chapter 0609, Appendix A required a detailed risk evaluation because the finding represented a loss of system and/or function. A regional senior reactor analyst (SRA) performed the detailed risk evaluation using SAPHIRE Version 8.1.6 and a modified Version 8.50 of the SPAR Model for Oconee. The SRA developed two change sets to model the total exposure time for the finding. The first simulated a common cause failure of both Keowee units with an exposure time of 7 hours. The second simulated the failure of both Keowee units while the standby buses were energized by the Lee Station for 5 hours. The result was less than 1E-6 for each Oconee unit, which would be a finding of very low significance (Green). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the area of human performance, in that the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5)
05000269/FIN-2018002-02	2018Q2	Oconee	Failure to Coordinate a No-later-than Arrival Time for the Shipment of a Category 2 Quantity of Radioactive Material	The inspectors identified aSeverity Level IV NCV of 10 CFR 37.75(b) when the licensee failed to coordinate a no-later-than arrival time for a Category 2 shipment of radioactive material. Specifically, the licensee failed to recognize that a package of primary resin contained a Category 2 quantity of Cobalt-60 prior to shipment, and therefore failed to arrange a no-later-than arrival time with the receiving licensee.
05000280/FIN-2008006-01	2008Q1	Surry	Failure to Evaluate and Use Limiting Case 4160 VAC Bus Frequency and Voltage in Design Calculations (Section 1R21.2.12)	The inspectors identified two examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate variations of emergency diesel generator (EDG) output frequency in electrical design loading calculations, and failure to consider worst case 4160 VAC bus voltage in safety related motor starting calculations. This finding was entered into the licensees corrective action program as condition reports (CR) 091493 and 091494. Planned corrective actions included revision of the EDG loading calculations to incorporate the most limiting voltages and frequencies. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in any EDG being inoperable based upon additional analysis that showed that the EDGs had sufficient margin to accommodate the increased loading due to worst case acceptably high EDG output frequency; and all safety related motor loads remained operable since they were still capable of starting with the revised worst case low voltage values. (Section 1R21.2.12
05000280/FIN-2008006-02	2008Q1	Surry	Failure to Use Appropriate Acceptance Criteria for Testing Battery Voltage at the One Minute Mark (Section 1R21.2.13)	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect acceptance criteria in test procedure 1-EPT-0106-01, Main Station Battery 1A Service Test. This finding was entered into the licensees corrective action program as condition report 091906. Planned corrective actions included revision of the main station battery test procedures to include the correct voltage at the one minute mark. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the station batteries to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in station batteries being inoperable based upon a recent review of station battery discharge test results. The inspectors determined that the lack of a thorough evaluation of condition report 022112, which addressed deficiencies in station battery test procedures such that resolutions addressed causes, was a significant cause of this performance deficiency. Failure to thoroughly evaluate problems such that resolutions address causes is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution and the aspect of thorough evaluation of problems (P.1(c)). (Section 1R21.2.13
05000280/FIN-2008006-03	2008Q1	Surry	Failure to Use Limiting Case High dP In 2-FW-MOV-260A Design Calculations (Section 1R21.2.19)	The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate the most limiting differential pressure (dP) for opening valve 2-FW-MOV-260A, auxiliary feedwater (AFW) cross-tie motoroperated valve (MOV). This finding was entered into the licensees corrective action program as condition report 091698. Planned corrective actions included internal inspection of the valve and revision of the evaluation that identified the most limiting dP for opening. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the AFW system to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design and plant modifications. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in 2-FW-MOV-260A being inoperable based upon additional analysis which showed that the MOV had sufficient margin to accommodate opening against the worst case high dP. The inspectors determined that the lack of control or understanding of the actual margin to maximum allowable dP to open 2-FW-MOV-260A was a significant cause of this performance deficiency. Failure to maintain design margins is directly related to the Resources component of the cross-cutting area of Human Performance and the aspect of maintenance of plant safety by the maintenance of design margins (H.2(a)). (Section 1R21.2.19)
05000280/FIN-2008006-04	2008Q1	Surry	Licensee-Identified Violation	The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs. 10 CFR 50, Appendix B, Criterion III, Design Control, requires that that measures shall be established to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. UFSAR Table 15.2-1 lists the AFW pumps as being components that will not fail during a tornado since they are protected by tornado resistant structures. Contrary to this, turbine drive AFW pumps 1/2-FW-P-2 were not completely protected in that the steam exhausts from the turbines could have been blocked by tornado missile damage. This was identified in the licensees corrective action program as CR 001132. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because it involved a severe weather initiating event and did not degrade more than one train of a multi-train safety system
05000280/FIN-2009007-01	2009Q2	Surry	Qualification of Fire Barrier Floor/Wall Penetration of Aluminum Conduit Through Sleeve	The team identified an unresolved item (URI) involving the qualification documentation for wall and floor fire barrier penetration seals. While inspecting the wall and floor fire barrier penetration seals, the team requested the licensees documentation for the qualification of a particular penetration seal configuration. That configuration was for one aluminum schedule 40 conduit (of various sizes as applicable) penetrating a 6 in. diameter floor or wall sleeve where the floor or wall was of poured concrete construction and the sleeve void around the conduit was filled with foamed silicon to the thickness of the floor or wall. The documentation package requested should establish a 3-hour fire rating, since the rated fire barrier walls and floors were required to have a 3-hour rating. In response, the licensee presented Impell Corporation Calculation No. 1250-111-C01, Penetration Seal Configuration Documentation Package, 10 in. Dow Corning Q3-6548 Silicone RTV Sealing Foam/North Anna and Surry, Rev.1. The qualification package or calculation was based on a tested configuration similar to that described above, except that the conduit was 3 in. or 4 in. galvanized steel. The team informed the licensee that this calculation was not valid to qualify aluminum conduit due to the lower melting temperature and greater heat conductance of aluminum as compared to steel. The licensee later transmitted supplemental information which included a fire barrier penetration seal fire test report for large diameter aluminum conduits through a sleeve. This new information was not a formal calculation comparing it to any installed penetration seal configuration at Surry. Moreover, certain aspects of the test data such as the temperature rise on the unexposed surfaces may not meet the licensing basis. At the time of issuance of this report, the team did not have sufficient information to determine the design criteria of that penetration seal. The team was aware that the fire barrier penetration seal configurations in question could probably be qualified by existing nuclear industry penetration seal testing data; therefore, there was no immediate safety concern. The licensee Initiated CR 339567 with an action item to establish a valid qualification package for the penetration configuration described above. URI 05000280, 281/2009007-01, Qualification of Fire Barrier Floor/Wall Penetration of Aluminum Conduit Through Sleeve, was established to track this issue until the final qualification package is reviewed
05000280/FIN-2011003-01	2011Q2	Surry	Unplanned Dilution of Unit 2 RCS	On May 28, 2011, while Unit 2 was operating in Intermediate Shutdown (>200 F, 310 psi), a control room operator noticed a decreasing level trend in the primary grade water tank over the past 2.5 hours. Additionally, it was noted that volume control tank and pressurizer level trends were increasing and charging seal injection flow was 101 gpm with letdown flow of 85 gpm. The licensee entered their abnormal procedure for emergency boration and conducted two emergency borations of the RCS while sampling RCS boron concentration and monitoring shutdown margin. Subsequently, it was identified that the cation demineralizer primary grade header isolation valve, 2-CH-19, indicated closed but was allowing primary grade water to leak by. This caused reverse flow through the cation demineralizer and introduced primary grade water into the RCS via the VCT. The licensee estimated that up to 30,000 gallons of PG water could have entered the RCS. Just prior to this event maintenance was conducted on 2-CH-19 and the valve was returned to service in a condition that allowed the primary grade water leakage flow path described above. The licensee entered this issue into their CAP as CR428947, and initiated Root Cause Evaluation (RCE) 001054. The inspectors require additional information, including the licensees completed investigation in RCE001054, to determine if there is a performance deficiency which is more than minor. This issue is identified as URI 05000281/2011003-01, Unplanned Dilution of Unit 2 RCS.
05000280/FIN-2011003-02	2011Q2	Surry	Failure to Classify and Declare a Notification of Unusual Event	A Green non-cited violation was identified by the inspectors for the licensees failure to classify and declare a Notification of Unusual Event when conditions warranted as required by 10 CFR 50.54(q) and 10 CFR 50.47(b)(4). The inspectors reviewed IMC0612, Appendix B, and determined that the finding was more than minor because it adversely affected the Emergency Response Organization performance attribute of the Emergency Preparedness cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Since the finding involved a failure to comply with regulatory requirements during an actual event, the inspectors reviewed IMC0609, Appendix B, Sheet 2, and determined that this was a finding of very low safety significance (Green) because it involved the failure to declare a Notification of Unusual Event. The cause of this finding involved the cross-cutting area of human performance, the component of decision making, and the aspect of conservative assumptions and safe actions, H.1(b), because the licensee failed to use conservative assumptions in the decision to not classify and declare the event as an Unusual Event.
05000280/FIN-2011003-03	2011Q2	Surry	Inadequate Qualification Testing of Fire Barrier Penetration Seals	A Green non-cited violation of Surry Units 1 and 2 Operating License Condition 3.I, Fire Protection, was identified by the inspectors for failure to have adequate qualification testing results, as directed by Appendix A to Branch Technical Position APCSB 9.5-1. Specifically, the licensee did not have sufficient testing results to qualify certain aluminum conduit configurations that penetrate 3-hour fire rated barriers separating fire areas containing redundant equipment required for safe shutdown. As part of the corrective actions, the licensee performed testing to determine the qualification of aluminum conduit penetrations, and performed modifications, as appropriate, to restore compliance. The finding is more than minor because it is associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and it affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events. Specifically, not having qualification testing results for aluminum conduits that penetrate fire rated barriers adversely affected the fire confinement capability defense-in-depth element because subsequent testing revealed some conduit configurations that did not meet the penetration seal criteria established in Branch Technical Position APCSB 9.5-1. The inspectors used the guidance of NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and determined that the performance deficiency represented a finding of very low safety significance (Green). Specifically, the fire areas in question either contained a non degraded automatic gaseous or water-based fire suppression system, or the exposed fire areas did not contain potential damage targets that are unique from those in the exposing fire areas. Inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance.
05000280/FIN-2011003-04	2011Q2	Surry	Licensee-Identified Violation	Licensee Technical Specification, 3.1.B.3, requires, in part, that the spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 deg F. Contrary to this, the licensee identified that this specification was exceeded on Unit 1 on April 17, 2011. The licensee created Engineering Technical Evaluation ETE-SU-2011-0042 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR422769.
05000280/FIN-2011003-05	2011Q2	Surry	Licensee-Identified Violation	Licensee Technical Specification, 3.1.B.3, requires, in part, that the spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 deg F. Contrary to this, the licensee identified that this specification was exceeded on Unit 2 on April 17, 2011. The licensee created Engineering Technical Evaluation ETE-SU-2011-0058 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR422778.
05000280/FIN-2011003-06	2011Q2	Surry	Licensee-Identified Violation	Licensee Technical Specification, 3.1.B.3, requires, in part, that the pressurizer heatup rate shall not exceed 100 degF per hour. Contrary to this, the licensee identified that this specification was exceeded on Unit 1 on April 20, 2011. The licensee created Engineering Technical Evaluation ETE-CEM-2011-0005 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR423197.
05000280/FIN-2011003-07	2011Q2	Surry	Licensee-Identified Violation	Licensee Technical Specification, 3.1.B.3, requires, in part, that the pressurizer heatup rate shall not exceed 100 degF per hour. Contrary to this, the licensee identified that this specification was exceeded on Unit 2 on May 26, 2011. The licensee created Engineering Technical Evaluation ETE-SU-2011-0073 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR428788.
05000280/FIN-2011003-08	2011Q2	Surry	Licensee-Identified Violation	NUHOMS Certificate of Compliance 1030, Amendment 0, Technical Specifications 2.1.c, Functional and Operating Limits, requires, in part, that the spent nuclear fuel stored in each 32PTH DSC/HSM-H at the Independent Spent Fuel Storage Installation (ISFSI) is to be qualified for four (4) heat load zones designated as Zones 1a, 1b, 2 and 3. Contrary to this requirement, the licensee identified that it failed to properly load fuel assemblies into four NUHOMS Dry Shielded Canisters (DSCs) resulting in the fuel assemblies exceeding the decay heat limit for the loading zones in two of the four center zones. Specifically, the Zone 1a and Zone 1b locations were reversed, resulting in the DSC Zone 1b heat load limits being slightly exceeded (less than one per cent in the worst case) at the time of loading. An evaluation performed by the licensee showed that all of the affected DSCs are currently in a safe condition as loaded in the HSMs. This issue is in the licensees CAP as CR419237, NUHOMS DSCs Loaded to Incorrect Heat Load Limits for Specific Orientation. This Severity Level IV violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.b of the NRC Enforcement Policy; specifically, the violation was identified by the licensee, the issue was placed into the licensees CAP, the violation was not repetitive as a result of inadequate corrective action, and the violation was not willful.
05000280/FIN-2011004-01	2011Q3	Surry	Failure to Follow Scaffolding Procedure Requirements	The inspectors identified a NCV of Technical Specifications (TS) 6.4.D for failing to follow the requirements of procedure MA-AA-105, Scaffolding. Specifically, the licensee did not adequately implement scaffold evaluation, screening, and risk requirements for multiple scaffolds constructed in the vicinity of safety-related equipment. The inspectors determined that the failure to follow TS required procedure MA-AA- 105, Scaffolding, by not properly identifying scaffolds for safety-related systems and performing the required engineering evaluations, constitutes a performance deficiency. This finding is considered more than minor because it is similar to IMC 0612, Appendix E, Example 4.a in that the licensee routinely failed to perform the required engineering reviews and evaluations for scaffolding. This finding is also associated with the external factors and equipment performance attributes of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors screened this finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance since it was a deficiency determined not to have resulted in the loss of operability or functionality. The cause of this finding involved the cross-cutting area of human performance, the component of resources and the aspect of training (H.2(b)), because the licensee failed to implement training sufficient to ensure that operators were aware of plant equipment which is designated as safety-related.
05000280/FIN-2011004-02	2011Q3	Surry	Failure to Consider Instrument Uncertainty and Establish Calibration Controls for Rotameters Used to Vent Gas from ECCS Systems	An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, (with two examples) was identified for the failure to establish measures to apply rotameter instrument measurement error and appropriate instrument calibration controls or standards when using instruments of this type to determine the size of voids discovered as a result of ECCS system venting. The issue was entered into the licensees corrective action program (CAP) as CR419024 and CR419243. The failure to establish and implement measures (1) to ensure the application of +/- 5% rotameter instrument error to as-found void measurement, and (2) to ensure that rotameters calibrated to standard pressure conditions were used when utilizing those instruments to evaluate the size of as-found voids were performance deficiencies. The performance deficiencies were greater than minor, because, if left uncorrected, they could result in a more significant safety concern. Specifically, the performance deficiencies represented programmatic issues and if instrument error and/or appropriate calibration standards were not applied to instruments used for future void characterization, then sufficient measurement error could reasonably result such that as-found voids, which challenge or exceed established acceptance criteria, may not be identified as intended by post venting evaluations. The finding was screened for significance using the Mitigating Systems cornerstone column of Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because the finding did not represent a design or qualification deficiency, did not represent the loss of a safety system function, did not represent the loss of a train for greater than the allowed outage time, did not represent the loss of risk significant equipment for greater than 24 hours, and was not potentially risk significant due to external events. Because the licensee had failed to implement complete, accurate, and up-to-date controls necessary to ensure that rotameter error and calibration standards were adequately addressed by procedures used to evaluate the impact of voids on emergency core cooling systems, this finding is assigned a cross-cutting aspect in resources component of the human performance area
05000280/FIN-2011012-01	2010Q4	Surry	Inaccurate Fire Watch Records	The licensee identified a violation of 10 CFR 50.48 Fire Protection requirements when it was determined that a laborer failed to conduct a roving fire watch patrol. The licensee took substantial disciplinary actions and entered the deficiency into the corrective action program for resolution as CR 379888. This issue was dispositioned using traditional enforcement due to the deliberate aspects of the performance deficiency. Furthermore, the failure to provide complete and accurate information has the potential to impact the NRCs ability to perform its regulatory function. An individual assigned as a fire watch deliberately documented the completion of fire watch rounds (Fire Watch Tour Documentation Sheet, Attachment 14) for locations in which he did not conduct the fire watches. This issue was considered more than minor due to the deliberate aspects of the performance deficiency. In accordance with the guidance in Supplement VII of the Enforcement Policy, this issue is considered a Severity Level IV violation because it involved information that the NRC required to be maintained by a licensee that was incomplete or inaccurate and of more than minor safety significance. No cross-cutting aspect was identified because this performance deficiency was dispositioned using traditional enforcement.

Semantic search

Query log

Navigation menu

Search