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On April 10, 2010, the Monticello Nuclear Generating Plant (MNGP) will be performing planned maintenance to upgrade the site phone system. The maintenance activity will result in the majority of commercial phone lines to the site and the onsite phone system being taken out-of-service for approximately two hours starting at approximately 0800 CDT. The site NRC Emergency Telephone System (ETS), phone lines (which includes ENS, HPN, and ERDS), auto ring lines between the site and the State of MN Emergency Operations Center, onsite radio communication systems, offsite radio communications systems, and the plant Gaitronics system will be unaffected by the maintenance activity. In addition, a commercial phone line in the Control Room and dedicated cellular phone for the Shift Manager will be available for off-site communications. The licensee notified the NRC Resident Inspector and will notify state and local agencies.

• * * UPDATE FROM NICK KRAMER TO CHARLES TEAL AT 1604 EDT ON 4/10/2010 * * *

On April 10, 2010 at 1030 hours use of the commercial phone lines and the on-site phone system was restored to the Monticello Nuclear Generating Plant (MNGP). Testing activities are in progress/completed to verify the functionality of affected phones. The licensee has notified the NRC Resident Inspector. Notified R3DO (Peterson).

On October 11, 2010 at 1305 CDT it was identified that the analysis of record for the Technical Specification 3.3.5.1, Table 3.3.5.1-1 function 1e and 2e, Reactor Steam Dome Pressure Permissive-Bypass timer (Pump Permissive) did not reflect current plant conditions. Specifically, the analysis was not updated to account for any increase in plant licensed power and a change to the RWCU (Reactor Water Cleanup System) isolation for enhanced ability to isolate RWCU on a line break on critical crack. The allowable value for these function is greater than or equal to 18 minutes and less than or equal to 22 minutes. All equipment associated with emergency core cooling function are unaffected. Discussion with General Electric indicates that a margin exists to accommodate the higher power level. Additionally, the changes to the RWCU isolation logic added leak detection instruments that will isolate RWCU earlier for the majority of pipe leaks. This discovery is being reported as an unanalyzed condition solely due to the lack of a formal analysis of current plant conditions. The NRC Resident Inspector has been notified.

• * * RETRACTION FROM MARTIN RAJKOWSKI TO JOHN KNOKE AT 1451 ON 10/29/10 * * *

Under NUREG-0737, Item II.K.3.18 is a regulatory requirement to implement a modification to extend the ADS (Automatic Depressurization System) to a unique event sequence that involves multiple failures including HPCI (High Pressure Coolant Injection) plus no operator action after 10 minutes. According to Item II.K.3.18, the bypass timer logic complements, but does not replace, the existing ADS actuation logic. This requirement is not associated with any design basis accident mitigation sequence at MNGP (Monticello Nuclear Generating Plant). The plant has performed an evaluation that addresses the changes in plant thermal power and RWCU (Reactor Water Cleanup System) enhanced isolation capabilities to the analysis of record. This evaluation concluded that Peak Clad Temperature (PCT) will remain under the 2200 ?F acceptance limit with the current Technical Specification allowable value, current plant configuration and current licensed thermal power. GEH (General Electric) did an independent evaluation that concluded that based on use of limiting scenarios analyzed for MNGP, this condition is only a lack of formal analysis of current plant conditions and that no Substantial Safety Hazard exists. It is judged that the maximum Reactor Steam Dome Pressure Permissive-Bypass timer (Pump Permissive) setting of 22 minutes will not result in a predicted PCT higher than 2200?F with consideration of the RWCU pipe break isolation instrumentation modification. Since the timers with their current setpoint will protect the fuel cladding, this event does not significantly degrade safety. Therefore, the event notification made on October 11, 2010 is being retracted. The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski)

Following a power reduction to 49 percent reactor thermal power at approximately 2312 (CDT) on November 4, 2010, 'B' high pressure feedwater heaters were isolated at approximately 2320 in preparation for repair of the 15B feedwater heater. Calculated core thermal power (CTP) rose by approximately 50 MWth. This resulted in all four average power range monitors (APRMs) failing to satisfy Technical Specification (TS) Surveillance Requirement SR 3.3.1.1.2 in that the absolute difference between the APRMs and the calculated power from the heat balance was greater than 2.0 percent rated thermal power (RTP). All four APRM channels were between approximately 3.5 and 4.0 percent lower than CTP. Since all four APRM channels were affected, the functions of the APRMs were inoperable and RPS trip capability had not been maintained. TS Conditions 3.3.1.1.A and 3.3.1.1.C were entered at 2325. All four APRM gains were adjusted and the TS Conditions were exited at 2349. This event is reportable under 10CFR50.72(b)(3)(v) as an event that could have prevented the fulfillment of the safety function of a system needed to: (A) shutdown the reactor and (D) mitigate the consequences of an accident. This event notification is being submitted outside the 8 hour reporting requirement. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM RANDY SCHULTZ TO HOWIE CROUCH AT 1652 EST ON 11/7/10 * * *

The following is an update to Event Notification 46397 made on 5 November 2010 concerning Average Power Range Monitors (APRMs) inoperability due to >2% deviation from calculated Core Thermal Power (CTP). During review of inputs to the CTP calculation, Operations staff determined that indicated power levels were not consistent with the indicated electrical output of the turbine generator. Subsequent investigation revealed an erroneous feedwater flow input from the isolated 'B' feedwater heater string. Actual CTP was less than indicated CTP by approximately 50 MWth. This was validated and verified through manual calculation. 50 MWth constitutes approximately 3% of rated thermal power. With the adjustments made to APRM on 4 November 2010, this resulted in all four APRMs being outside of the acceptable tech spec value. Following Operations review, at 1108 on November 7, 2010, all four APRMs were declared inoperable and tech. spec. 3.3.1.1.A and 3.3.1.1.C actions were initiated. Values consistent with the actual plant configuration ('B' feedwater heater string isolated) were input to the CTP calculation with subsequent indicated CTP returning to actual CTP. All four APRM gains were adjusted and APRMs returned to operable status. Tech Spec 3,3.1.1.A and 3.3.1.1.C conditions were exited at 1207 on November 7, 2010. The licensee has notified the NRC Resident Inspector. Notified R3DO (Giessner).

Dose assessment cannot be performed onsite. Pursuant to 10CFR50.72 (b)(3)(xiii), Monticello Nuclear Generating Plant is reporting a planned loss of emergency assessment capability scheduled for 0915 hours on May 10, 2011 for system maintenance. Dose assessment in the control room, TSC, and EOF are non-functional. Main meteorological tower data is unavailable. Projected outage duration is approximately 7 hours. Control room and local instrumentation/annunciation remain available for accident assessment. Telephone lines remain available between site emergency response facilities and offsite agencies for communication of plant status information. Dose assessment capabilities will exist at the backup EOF. Backup meteorological data will be available in the control room, TSC, and EOF. The licensee will not be in any Technical Specification action statements as a result of this evolution. The NRC Resident Inspector has been notified.

* * * UPDATE FROM JEFF OLSON TO PETE SNYDER ON 5/10/11 AT 1650 EDT * * * 

The licensee has restored their dose assessment capability. Notified R3DO (Passehl).

On 8/29/11, Monticello Nuclear Generating Plant's Technical Support Center (TSC) power supply will be isolated to perform a planned maintenance activity. The maintenance activity requires implementation of compensatory measures to maintain TSC functions during the activity. The compensatory measures include having the Emergency Director report to the Control Room and co-locating the remaining TSC staff at the EOF should an event be declared requiring ERO activation. The ERO has successfully previously demonstrated the ability to implement these compensatory measures. The maintenance activity is scheduled to be completed with the TSC returned to full functionality by the end of the dayshift on 8/29/11. The site Emergency Response Organization has been notified of the maintenance activity and instructed on the planned compensatory measures to be implemented during the activity. The licensee has notified the NRC Resident Inspector.

• * * UPDATE AT 1735 EDT ON 8/29/11 FROM ANDERSON TO HUFFMAN * * *

The licensee reports that TSC maintenance activities were completed at 1308 CDT and that the TSC was declared fully operational at 1545 CDT on 8/29/11. The licensee has notified the NRC Resident Inspector. R3DO (Kunowski) notified.

The station experienced a lockout of the 2R Auxiliary Power Transformer. The resulting transient caused an automatic actuation of the RPS system. All control rods fully inserted. A Group 2 Primary Containment isolation occurred. Both 11 and 12 Emergency Diesel Generators started on a loss of voltage signal. Equipment response was that the 11 ESW (Emergency Service Water) pump (cooling for the #11 Emergency Diesel) failed to develop required pressure. The #13-4160V non-safety related bus failed to restore after the transient (and feed the Division 1 Essential Bus). Additionally, the #15 bus transferred to the 1AR transformer (and is feeding the Essential Bus). The #11 Emergency Diesel Generator is currently tagged out of service. Electrical supply is being provided by offsite power. Reactor heat is being removed through the main steam line to the main condenser and reactor water inventory is being provided by the feedwater system. The SRVs lifted and reseated. The HPCI system was manually place into a pressure control mode. The Minnesota Pollution Control Agency is being notified due to the licensee violating the site discharge canal temperature rate of change limit. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM RYAN RICHARDS TO JOHN KNOKE AT 1730 EDT ON 11/01/2011 * * *

Prior to this event the 'B' Control Room Emergency Filtration (CREF) and 'B' Control Room Ventilation (CRV) Systems were inoperable for planned maintenance. On 10-21-11 at 1325 CDT, the #11 EDG ESW Pump was declared inoperable due to low cooling water pump flow, resulting in the #11 EDG being inoperable, which in turn resulted in the 'A' CREF and 'A' CRV being inoperable. Contrary to reporting requirements this condition was not identified and reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as required within 8-hours in the previous event notification. This condition resulted in a loss of safety function for both divisions of CREF and CRV. This update amends the 10-21-11 event notification to include this as an 8-hour non-emergency event pursuant to 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified the R3DO (Nick Valos)

At 1258 CDT on 07/05/2012, Operations was notified that both panels of Door 45 (south doors for the reactor building railroad bay airlock) were blocked by a man lift. Blocking both doors represents an unanalyzed condition as a flow path through the door is assumed for pressure relief during postulated HELB events. The man lift was immediately removed correcting the situation and all work related to Door 45 was stopped. Door 45 was blocked for approximately 20 minutes. The NRC Resident Inspector has been notified.

• * * RETRACTION FROM BART BLAKESLEY TO DONALD NORWOOD AT 1605 EDT ON 8/31/2012 * * *

The purpose of this notification is to retract the previous Event Notification Report (#48072) made by the Monticello Nuclear Generating Plant on 7/5/2012. The initial report indicated that blocking both panels of the railroad bay doors by a man lift represented an unanalyzed condition, as a flow path through the door is assumed for pressure relief during postulated HELB events, and was reported in accordance with 10CFR50.72(b)(3)(ii)(B), Unanalyzed Condition. Since the initial report, Engineering has completed an evaluation that demonstrates equipment supporting safe shutdown would have been capable of performing their specified design function during postulated HELB events. Based on this analysis, the condition initially reported in EN #48072 did not result in an unanalyzed condition that significantly degraded plant safety and is therefore retracted. The NRC Resident Inspector has been informed of this retraction. Notified R3DO (Cameron).

At 2045 (CDT) on 8/14/12, MNGP (Monticello Nuclear Generating Plant) Operations determined that valves RHR-82 and RHR-84 had been inappropriately closed as part of an isolation clearance order for work on shutdown cooling suction piping. These valves are required to be open to provide overpressure protection for RHR piping passing through primary containment penetration X-12. Upon discovery of the condition, Primary Containment was declared Inoperable and the Required Actions of Tech Spec 3.6.1.1 were entered. Following discovery, the isolation was restored and the valves opened. At 0001 (CDT) on 8/15/12, Primary Containment was declared Operable. This issue is being reported in accordance with 10CFR50.72(b)(3)(v)(C) and 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety functions of a system needed to control the release of radioactive material or to mitigate the consequences of an accident. The MNGP Senior NRC Resident Inspector has been notified of this issue. The licensee will contact the Minnesota State Duty Officer.

• * * RETRACTION FROM RANDY SAND TO CHARLES TEAL ON 08/23/12 AT 1545 EDT * * *

This notification is a retraction of ENS 48190 based on further engineering evaluation. Monticello had previously evaluated penetration X-12 for thermally induced over pressurization. The evaluation qualified the piping components in the penetration for a maximum pressure of 3,306 psig using ASME Section III Appendix F operability criteria. The peak pressure calculated for the penetration was 2,743 psig based on Reactor pressure of 1000 psig with Reactor in Mode 1, and at worse case LOCA conditions for the Drywell. These assumptions and parameters envelop those that were present when valves RHR-82 and RHR-84 were closed on August 14, 2012. Therefore, this event would not have prevented the fulfillment of the safety function reported. The NRC Resident Inspector has been notified. Notified R3DO (Duncan).

During a walkdown on December 20, 2012 at 1600 CST, two degraded Appendix R fire barriers (walls) were identified. These barriers separate the Torus Room (Fire Area IV)/ 'A' RHR Room (Fire Area I) and the Torus Room (Fire Area IV)/ 'B' RHR Room (Fire Area II). The walls separate Appendix R fire safe shutdown divisional equipment. A fire watch was established as a compensatory measure immediately following identification of the issue on December 20, 2012. The barrier affecting the 'B' RHR Room has been repaired on both sides. The barrier affecting the 'A' RHR Room has been repaired on the Torus Room side. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The fire watch remains in place until verification of the completed repair is performed. The licensee notified the NRC Resident Inspector.

• * * RETRACTION AT 1746 EST ON 2/07/13 FROM JACK EARSLEY TO HUFFMAN * * *

An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on December 21, 2012 for degraded fire barriers between the Torus Room (Fire Area IV) and 'A' RHR Room (Fire Area I), and the Torus Room (Fire Area IV) and 'B' RHR Room (Fire Area II). Subsequent engineering analysis determined that the degraded fire barriers maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted. The licensee will notify the NRC Resident Inspector. Notified R3DO (Kunowski).