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During scheduled maintenance, at 1138 CDT, the Fort Calhoun Station Shift Manager was notified via phone call and condition report of an unanalyzed condition which was the result of the maintenance on Shutdown Cooling Heat Exchanger valves. This condition could have led to the inability of the Component Cooling Water (CCW) system to perform its design function of providing a cooling medium for the Containment atmosphere under Loss of Coolant Accident (LOCA) conditions. This was identified by OPPD (Omaha Public Power District) staff engaged in Design Basis Reconstitution. As part of the maintenance, HCV-484, Shutdown Cooling Heat Exchanger AC-4A CCW Outlet Valve, and HCV-481, Shutdown Cooling Heat Exchanger AC-4B CCW Inlet Valve, were opened. Under these conditions, with the assumed single failure loss of DC control power and accident conditions of a LOCA, CCW would be allowed to flow through both shutdown cooling heat exchangers, effectively bypassing flow to the Containment Cooling Units. These conditions are not assumed under plant design basis calculations, and therefore placed the plant in an unanalyzed condition. Following clearance removal at 1535 CDT, both HCV-484 and HCV-481 were returned to service and the condition described above no longer exists. The licensee notified the NRC Resident Inspector.

• * * UPDATE ON 6/15/16 AT 1836 EDT FROM JOHN BLALOCK TO DONG PARK * * *

Discovered 6/15/2016 at 1330 CDT: During the extent of condition review for the above ENS notification, it was discovered that the unanalyzed condition that occurred on 5/10/2016 also occurred five other times during the past 3 years. Details for these additional occurrences will be included in the 60-day Licensee Event Report associated with the original 5/10/2016 ENS notification. The licensee has notified the NRC Resident Inspector. Notified R4DO (Taylor).

On October 14, 2013 a calculation for the containment internal structural analysis was revised and accepted by the station. This calculation limited the Safety injection tank level to 74%. On October 16, 2013 Safety injection tank level was raised to 100% for approximately 13 hours in preparations for plant start-up. While the plant was safely in a cold shutdown condition, this represents a reportable unanalyzed condition. This issue is of a historical nature and does not question the current operability of any plant systems or structures. This was self identified during a Fort Calhoun calculation review. The licensee notified the NRC Resident Inspector.

• * * RETRACTION FROM MICHAEL PEAK TO DANIEL MILLS AT 2335 EDT ON 3/12/15 * * *

Following review of the reported event, attendant calculations and associated documentation, engineering personnel determined that the condition described in event notification EN50800 did not place the plant in an unanalyzed condition. Revision 1 of a calculation for the containment internal structural analysis demonstrated that when the safety injection tanks 'B' and 'D' are 100% filled in an outage condition, approximately a 10% safety margin is maintained. This revision was the calculation of record at the time the safety injection tank levels were raised above 74%, in October, 2013. Revision 2 of the calculation was completed to remove excess conservatism and to provide a closer representation of available margin. In addition, margin was also improved by limiting tank level to 74%. However, improving margin by limiting tank level to 74% does not result in an unanalyzed condition when tank level is 100%, as adequate margin remains. Therefore this event is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R4DO (Okeefe).

It has been determined that some instrument racks in the Containment and Auxiliary buildings do not meet their design basis capacity due to inadequate embedment depth of the seismic anchors. Assumptions made about embedment depth for a previous event were determined to be incorrect; therefore, the design basis capacity cannot be assured. This report is being made under 10 CFR 50.72(b)(3)(ii)(B), 'Unanalyzed condition'. The licensee has notified the NRC Resident Inspector.

• * * RETRACTION ON 6/10/2013 AT 1630 EDT FROM DAVID ORTIZ TO MARK ABRAMOVITZ * * *

Additional evaluation has determined that the instrument racks are adequately anchored. Therefore, this event is not reportable. The licensee notified the NRC Resident Inspector. Notified the R4DO (Gepford).

During a review of plant installed instrumentation racks inside containment, two instrument racks were identified that were over the analyzed weight for the seismic analysis. The instruments on these racks are used for reactor coolant pressure transmitters that are part of the Reactor Coolant System (RCS) pressure boundary, as they are connected via instrument lines to the RCS with no remote closure capabilities. A failure of these racks during a seismic event due to the excessive weight could result in an unisolable leak from the RCS within containment based on engineering judgment. This results in the RCS principal safety boundary being in a degraded nonconforming condition as the Updated Safety Analysis Report (USAR) specified Class 1 requirement is not being met for the current seismic design. Further engineering analysis is in progress to address the weight issue for these racks and mounting requirements. The plant is shutdown and in Mode 5 with the reactor vessel head removed, so RCS is not intact and not required to be for current plant conditions. This report is being made in accordance with 10CFR72(b)(3)(ii)(A) for a degraded condition. The licensee has notified the NRC Resident Inspector.

• * * UPDATE ON 8/3/12 AT 1708 EDT FROM JULIE BISSEN TO DONG PARK * * *

On June 4, 2012, at 2059, Fort Calhoun Station made an 8-hour non-emergency notification for a degraded condition. Subsequent internal review has determined that the initial reporting criterion for degraded condition, 10 CFR 50.72(b)(3)(ii)(A), was incorrect. The instrument racks were identified as being over the analyzed weight for the seismic analysis. This is an unanalyzed condition, not a degraded condition. The report made on June 4, 2012 should have been made under 10 CFR 50.72(b)(3)(ii)(B), unanalyzed condition. The licensee has notified the NRC Resident Inspector. Notified R4DO (Werner).

While investigating operating experience from another station it was determined that Fort Calhoun Station (FCS) is subject to similar conditions. The operating experience involved setpoint drift of safety related pressure switches beyond what had been accounted for in the station's safety analyses. Following investigation and evaluation, it was determined that pressure switches that provide safety related signals for high containment pressure to the reactor protection system (RPS) and engineered safeguards actuation circuitry may be similarly affected at FCS. The impact of the potential drift was evaluated, and it was determined that neither RPS nor the engineered safeguard circuitry may actuate at the required containment pressure of 5 psig. An evaluation determined that the actuation may not occur until slightly higher than the required pressure. Other systems are currently being evaluated to see if this same condition applies. The station is in MODE 5, refueling shutdown condition, and there is no immediate safety concern. The pressure instruments are located in the penetration area which is subject to elevated temperatures. The licensee notified the NRC Resident Inspector.

• * * RETRACTION AT 1616 EDT ON 10/19/2012 FROM LUKE JENSEN TO MARK ABRAMOVITZ * * *

The condition was initially determined to be reportable under 10CFR50.72(b)(3)(ii)(B), plant in unanalyzed condition, based on a conservative assumption that the error introduced violated not only the Technical Specification limit (5.0 psig) but also the safety analysis limit of 5.4 psig, USAR Table 14.1-1. Subsequent evaluation of actual data concluded that the safety analysis limit was not exceeded and therefore not reportable under 10 CFR 50.72(b)(3)(ii)(B). LER 2012-004-1 reported this condition under 10CFR50.73(a)(2)(i)(B), 10CFR50.73(a)(2)(ix)(A), and 10CFR50.73(a)(2)(v)(A,B,C,D). Revision 2 of the LER will correct the reporting criteria. The NRC Resident Inspector was notified by the licensee. Notified the R4DO (Pick).

During a review of environmental qualification records for reactor containment building electrical penetrations, six penetrations were identified that may not provide an adequate seal during worst case (Design Basis Accident (DBA)) conditions as required. These penetrations are through wall from the containment into the auxiliary building. The conditions that could cause degradation of the electrical penetration seals are not applicable to this operating mode. The station is currently in a refueling mode. This event was identified on March 2, 2012. The reportability was confirmed on May 1, 2012 at 1502 CDT. The current penetration configuration has existed since the plant was built. The area of concern is that the Teflon connections may degrade under conditions of high radiation and high temperature during a DBA event. The licensee is investigating the extent of the condition and repair techniques. The licensee notified the NRC Resident Inspector.

* * * UPDATE FROM ROBERT KROS TO PETE SNYDER AT 1523 ON 6/26/12 * * * 

On review of CR 2012-01947 by a new Project Manager, who was brought in as a subject matter expert on HELB/EEQ, and issue was identified with the 530 primary containment electrical penetration feed-throughs used for non-CQE devices. The CR (Condition Report) correctly notes that under the original accident testing, the Teflon seals failed, and water was noted leaking from these penetrations. On further review, the following was noted: Due to the design of the penetration feed-throughs, when the inboard Teflon seal fails (as it is expected to, due to high level of radioactivity in the primary containment, following a Loss of Coolant Accident (LOCA)), the atmosphere of the primary containment will be introduced to the penetration assembly, first through the failed seal or seals, and then through the weep hole between the inboard and outboard seals of the feed-through. This will put the same high level of radioactivity in direct contact with the outboard seals, resulting in the failure of its Teflon Seal. This would result in approximately 530 breaches of the Primary Containment during post LOCA conditions. The existing vendor analysis does not assume any contribution to the outboard seal exposure from the mixing of containment atmosphere with the penetration air after the failure of the inboard seal. This is probable, as each feed-through has a weep hole. Once the inboard seal fails, the penetration will be filled with containment atmosphere to equalize the pressure, which will bring the associated noble gas and Iodine fraction in proportion, into the penetration. The licensee notified the NRC Resident Inspector. Notified R4DO (Clark)

• * * UPDATE AT 1748 EDT ON 7/17/12 FROM MOECK TO HUFFMAN * * *

During the extent of condition review for CR 2012-01655 and 2012-01947 additional penetration feed-through assemblies were identified that are subject to the same failure mechanism. These penetrations are associated with the containment sump recirculation isolation valves, and also associated with the personnel air lock. The licensee notified the NRC Resident Inspector. Notified R4DO (Walker).

The Waste Disposal System (WDS) Class 1 piping requires operable seismic supports downstream of the isolation valve class break. Currently, eight (8) INC (International Nuclear Safety, Corp.) snubbers have been degraded to (Non Nuclear System) NNS Class 4 ridged struts. The snubbers original design function was to allow thermal motion but restrain seismic motion. The snubbers have been identified as potential to create an unanalyzed condition that over stresses the safety class 1 drain pipe upstream of the isolation valve if the snubbers on the drain pipe downstream of the isolation valve were in a locked condition (acting as a strut). Per NRC bulletin 81-01, these snubbers are assumed to be frozen and do not allow movement of the pipe; thus, they have been degraded to rigid struts as they are not in the snubber program and are not tested. They still provide a seismic safety function for (class) II/I issues and act as a strut to provide horizontal restraint to the WDS piping.

The snubbers were removed from the piping system and tested to determine their performance and if they would have moved to allow thermal growth. Six snubbers failed the test and were either in a locked condition or their movement was dimensionally small relative to the required movement. The (Reactor Coolant System) RCS is within acceptable stress values with the snubbers removed. The 8-hour regulatory reporting time has been exceeded. An initial Reportability Evaluation was completed on March 26, 2012 and had determined the supports were operable. A second Reportability Evaluation later determined the supports have been inoperable since October 6, 2011. The WDS is used to drain the RCS. The licensee will notify the NRC Resident Inspector.

• * * RETRACTION ON 9/26/12 AT 1949 EDT FROM ROBERT KROS TO DONG PARK * * *

Additional review and testing demonstrated that (there was) no degradation of the RCS from thermal fatigue. The analysis demonstrates adequate past performance of the snubbers with regard to thermal fatigue. The impact of the snubber has been analyzed and determined to have not resulted in an unanalyzed condition that significantly degraded plant safety. Therefore, this event is being retracted. The failure to retract this notification in a timely fashion has been entered into the corrective action system. The licensee will notify the NRC Resident Inspector. Notified R4DO (Werner).

Operations identified a potential flooding issue in the Intake Structure 1007 ft. 6 in. level. The area of concern is a the hole in the floor at the 1007 ft. 6 in. level where the relief valve from FP-1A discharge pipe goes through the raw pump bay and discharges into the intake cell. There is one penetration of concern. Flooding through this penetration could have impacted the ability of the station's Raw Water (RW) pumps to perform their design accident mitigation functions. Efforts are in progress to seal the penetration. This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v). The licensee notified the NRC Resident Inspector.

• * * RETRACTION ON 9/26/12 AT 1949 EDT FROM ROBERT KROS TO DONG PARK * * *

The penetration in question is not an external penetration and is not within the scope of the CLB (Current Licensing Basis) and therefore the condition is not reportable. The penetration is internal to the intake structure and does not affect internal flooding. The failure to retract this notification in a timely fashion was identified while reviewing flood related station notifications from 2011 and has been entered into the corrective action system. The licensee has notified the NRC Resident Inspector. Notified R4DO (Werner).

At 1622 hours CDT, an electrical ground on 480 Volt Bus 1B3A was determined to be from a supply cable to Motor Control Center (MCC)-3A1. Isolating loads on this MCC required securing power to HCV-1385, Steam Generator RC-2B Inlet Isolation Valve. This condition results in the valve being unable to close on a Steam Generator Isolation Signal (SGIS), which requires entry into Technical Specification 2.0.1(1). This Technical Specification requires the plant to be placed in a Hot Shutdown condition within 6 hours. A plant shutdown to Mode 3 was commenced at 1740 hours CDT. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM DAVID SPARGO TO DONALD NORWOOD AT 0036 EDT ON 4/9/2010 * * *

At 2123 hours CDT, the Reactor was manually tripped from 22% Reactor Power in order to meet the requirements of Technical Specification 2.0.1(1) and have the reactor in a hot shutdown condition. All systems functioned properly. At 2124 CDT, the plant entered Mode 3 Shutdown Condition. AT 2235 CDT, HCV-1385, Steam Generator RC-2B Inlet Isolation Valve has been manually closed. Technical Specification 2.0.1(1) has been exited. The licensee is reporting the manual scram under 10CFR50.72(b)(2)(iv)(B). The licensee notified the NRC Resident Inspector.

An 8 hour notification per 10 CFR 50.72(b)(3)(ii)(B) is being made due to discovery of a previously unanalyzed plant condition. The chemical volume control system (CVCS) charging line is a high energy line susceptible, within design space, to a rupture that could result in pipe whip and an impingement condition. The containment penetration (M-3) for the charging line is directly below the high-pressure safety injection (HPSI) header penetrations (M-5 & M-6). Due to this proximity, a failure of the charging line could impact the SI-1503 class HPSI line (M-6) causing damage to the header and rendering it incapable of fulfilling its design function. The two HPSI headers are different class piping with the M-5 penetration being a 2500 psig (SI - 2501 Class) line and the M-6 penetration being a 1500 psig (SI-1503 Class) line. The 2500 psig line is constructed of robust enough piping to not be susceptible to failure, however both HPSI headers are cross connected prior to location of concern. Although both HPSI and CVCS breaks are isolable, operator action would be required to identify and isolate the break location. The possible result of a high energy line break of charging piping could result in all three pipe headers being inoperable until manual action could be taken to isolate leakage on the 1500 psig HPSI line. Plant is currently shutdown for refueling outage with scheduled startup of November 21, 2006. The licensee notified the NRC Resident Inspector.

• * * RETRACTION PROVIDED BY ERICK MATZKE TO JEFF ROTTON AT 1119 EST ON 11/07/06 * * *

Additional review of the stations design and licensing basis has determined that the charging and high pressure safety injection systems are correctly designed to the stations high energy line break criteria for lines inside containment. This situation is not reportable and therefore event notification 42943 is being withdrawn. The licensee notified the NRC Resident Inspector. Notified R4DO (Clark)