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05000029/LER-2001-001 +Yankee Nuclear Power Station ceased power operation in February 1992 and is being decommissioned. On 06/26/01 during the conduct of a Nuclear Safety (Quality Assurance) Audit a discrepancy regarding the alarm setpoints for the Spent Fuel Pit (SFP) Area Radiation Monitor (ARM) was identified. The SFP ARM is an instrument required by Technical Specification 3.3 to ensure early detection of inadvertent criticality during fuel handling activities. The Technical Specification requires the alarm setpoints for the ARM be set at less than 5 mr/hr or two times the background radiation level, whichever is greater, while moving irradiated fuel, control rods or sources. The discrepancy identified was that the background radiation level annotated on procedure OP-4816, "Functional Test and Alarm Setting of the Area Radiation Monitoring System" was 2 mr/hr while the alarm setpoint for both the alert and high alarms was set at 7 mr/hr, thus greater than the Technical Specification requirement. As such, this LER is submitted in accordance with 10CFR50.73(a)(2)(i)(B) as a condition of non-compliance with a Technical Specification. No fuel handling evolutions were in progress at the time of discovery of this issue.  +
05000133/LER-2010-001 +On June 24, 2010, while conducting the quarterly inventory of radioactive sources in accordance with Humboldt Bay Power Plant (HBPP) Unit 3 Radiation Control Procedure (RCP) -6D, "Inventory and Controls for Radioactive Sources," it was discovered that source number HBS-498 was missing from the count room. HBS-498 is a mixed gamma source with an activity of 0.35 micro-curies as of August 9, 2010, and is used for the calibration of gamma detectors. The calibration source radionuclide composition yields an aggregate quantity of missing licensed material of 53 times the quantity specified in 10 CFR 20 Appendix C, which exceeds the reporting criterion of 10 times the quantity specified in Appendix C. At the time Revision 0 to this LER was submitted, it was reported that an extensive search had not been successful in locating the source to date (i.e., as of August 20, 2010); however, the missing source was located in Unit 3 on February 1, 2011. The root cause for this event has been determined to be inadequate Radiation Protection procedures to ensure control of radioactive sources. Procedure RCP-6Lt was revised to strengthen the source control process.  +
05000219/LER-2001-001 +On November 11, 2001, a 4160 VAC cable failure de-energized the 1B2 Unit Substation of the 480 VAC system. Due to the equipment which was declared inoperable, it was determined that a reactor shutdown would be required. On November 12, 2001, at 3:33 am, the reactor was placed in the COLD SHUTDOWN CONDITION. The cause of the cable failure was determined to be a localized insulation weakness aggravated by water intrusion into the cable conduit. The safety significance of this occurrence was determined to be minimal. Although the 1B2 Unit Substation of the 480 VAC system was lost, the redundant electrical division remained fully operable at all times. At no time during this event did a functional failure of any safety system occur. The plant remained within Technical Specifications limits at all times, and achieved a COLD SHUDOWN CONDITION within the allowed time limits. The failed portion of the cable was replaced, the plant was restarted and resumed POWER OPERATION. Long term actions include evaluating a new design cable and possible rerouting of the cable run.  +
05000219/LER-2002-001 +At 7:00 AM on March 23, 2002, a sufficient volume of water had condensed in the Offgas Radiation Monitor sample chamber to cause the monitor to read unacceptably low. This resulted in the monitor being incapable of performing its intended function of isolating the Off Gas Line if Technical Specification release limits were exceeded. The sample chamber heat trace had failed allowing moisture to accumulate in the sample chamber. The system engineer, in response to a maintenance request, developed a troubleshooting plan which included a procedure to remove moisture from the sample chamber. On April 9, 2002, at 5:49 PM, draining was commenced on the sample chamber. Upon the draining of the chamber, indications returned to normal. The system was declared operable at 7:25 PM. The cause of the Technical Specification violation was a failed sample chamber heat trace. The heat trace was restored to an operable condition. The safety significance of this occurrence was minimal. Although the Air Ejector Offgas Radiation Detection System was inoperable, the Stack Radioactive Gaseous Effluent Monitoring System was in operation. No unexpected activity was noted. Weekly samples performed by the chemistry department of air ejector offgas did not show an increase in offgas activity.  +
05000219/LER-2002-002 +On October 11, 2002, Local Leak Rate Test results indicated that Main Steam Isolation Valve V-1-0007 exceeded the Technical Specification leak rate limit of .05(.75)4 at 35 psig (equivalent to 15.98 SCFH). The leak was unable to be quantified and was greater than 50 SCFH at 35 psig. The cause of this occurrence was attributed to component degradation. The safety significance of this occurrence is considered minimal as the total penetration leakage would have been limited by Main Steam Isolation Valve V-1-0009 in the same header. The leakage past V-1-009 was quantified at 2.214 SCFH. V-1-0007 was refurbished and successfully leak rate tested prior to restart from the 1R19 Refueling Outage.  +
05000219/LER-2002-003 + * A void was discovered in an area expected to be filled with sand beneath a portion of the two 480 VAC switchgear rooms. This void created an open area between two 4160 VAC feeder conduits. Because of the void, Appendix R electrical separation criteria were no longer met. Apparently, sand may not have completely filled the area and/or had settled over time creating this void. The safety significance of this discovery is minimal, as there Is no combustible material in the void. Both cables are contained in conduit and have sufficient Class 1 E electrical separation. Immediately upon discovery, a continuous fire watch was stationed. Additional actions were subsequently taken to open communication between the void and adjacent smoke detectors. This would provide early warning early warning of a degraded condition. Adequate separation will be provided by re-filling the void between the two 4160 VAC feeder conduits to meet Appendix R requirements. The sand fill will be monitored. No similar events have occurred. 1   +
05000219/LER-2003-001 +On April 15, 2003, at approximately 4:50 A.M. the Oyster Creek Operations Director arrived at the station and was appropriately processed by the Outer Owner Controlled Area (OCA) checkpoint. He proceeded to the Inner OCA checkpoint where he noticed that the security guards did not immediately exit the guardhouse to verify his identification and check his vehicle. He proceeded to the Guardhouse where he observed the guards, apparently asleep. An immediate search of the Inner OCA was conducted to ensure that no unauthorized vehicles had entered. The Inner OCA checkpoint is monitored by a surveillance camera. A review of the videotape revealed that no vehicles entered the Inner OCA during this period. This information was corroborated by an interview with the Outer OCA guard. Both guards were relieved from all duties. Two new guards manned the Inner OCA post. Both guards were "For Cause" Fitness for Duty tested. All readings were negative. All required NRC and Exelon notifications were made. Security reinforced management expectations to all Security Personnel on attentiveness to duty.  +
05000219/LER-2003-002 +On May 20, 2003, at 0030 hours, 4160 VAC bus IC Iccked out due to a ground fault. The plant continued operating at full power. Due to the equipment Made inoperable by the loss el power to bus 10, Technical Specification requited the reactor to be placed In the COLD SHUTDOWN CONDITION. The reactor was manually scrammed at 0943 hours. llie COLD SHUTDOWN CONDITION was reached at 1913 Tom .. _. _ The safety significance of this event is considered minimal. The redundant 4160 VAC bus remained to service and redundant safety-related equipment remained operable. The plant remained witNn Technical Specification limits and achieved SHUTDOWN and COLD SHUTDOWN within the allowed time limits. Previous experience has shown this type d cable Is subject to accelerated degradation from water at the site of any defects in the Insulation. Ni cables in this run were replaced with cables of a afferent rnaradacture. Corrective action included confirming that all burled cable powering safety-related equipment was net of the typo that failed. NRC FORM ass (1-2M1)  +
05000219/LER-2003-003 +Nszt = below or In 60-73(aX2)(Y)(C) Form 202203(a) (2) (tv) 50.73(8X2)(1XA) 50-73(aX2)(Y)(13) 20.2203 a u 50.73 a „(), ; 50/3 a; u.) ' - PtI:e1, 202203(a)(2)(y1) 50.73(R)(2)(1)(G1 50 73(a)(2)(vIllAA) - " *■ 2 1 .1.,,.. _50.13(a)(2)(01i)(B) - - :. . , Mi(ie411, .._ 202203(a)(3)10 , 50.73(a)(2)(9)(A) _ LICENSEE CONTACT FOR THIS LER (12)  +
05000219/LER-2003-004 +On August 22, 2003, a Turbine Trip was caused by a spurious actuation of Moisture Separator Hi-Hi Level switch, LS-4-691. This resulted in a Reactor Scram from 100% power. The reactor shut down as designed. Plant cooldown to cold shutdown was required due to the trip of all five recirculation pumps. The safety significance of this event is considered minimal. The plant responded as designed for this type of event. Technical Specification limits were maintained. There was no radioactive release. All safety systems were fully operable. Off-site power was available. Operator performance was satisfactory. All four level switches were replaced and a root cause is in progress.  +
05000219/LER-2004-001 +On May 17, 2004 at 04:40, at the end of the bi-weekly #1 Emergency Diesel Generator (EDG) (EK - DG) load test an operator reported that the EDG was making an unusual noise and the cooling fan belts and pulleys were observed to be loose. It was also observed that the pillow-block bearing supporting the cooling fan drive shaft and sheave assembly had one bolt missing and one bolt loose. A 7-day Technical Specification LCO was entered at 05:00. The previous maintenance on the pillow block bearing support occurred about 17 days earlier during the two year overhaul completed in late April 2004. A review of the procedures and documentation that were used to perform the cooling fan maintenance and interviews with the individuals that performed the fan work during the overhaul were conducted. The investigation determined that the pillow block bearing hold down bolts had not been torqued as required by the vendor manual. All repairs were completed on May 17, 2004 at 17:12. Immediately following the repairs, #1 EDG Load Test was again performed as the cooling fan post-maintenance test and to validate operability. The diesel was declared operable at 20:25. No previous similar events of improper maintenance activities on the diesel generators leading to loose and missing bolts and fasteners due to improper torquing were identified.  +
05000219/LER-2004-002 + * On May 14, 2004, at approximately 0813 hours with the plant operating in the RUN Mode at 100% power, General Electric (GE) informed Oyster Creek Generating Station (OCGS) of a change in the calculation of Peak Cladding Temperature (PCT) and maximum local cladding oxidation. A new heat source has been postulated during the Loss of Coolant Accident (LOCA) event that involves the recombination of hydrogen and oxygen within the fuel bundles during core heatup. This event was initially reported to the NRC on May 14, 2004 in a voluntary notification as a result of the 10 CFR 50.46(a)(3)(ii) requirement to report this issue in accordance with 10 CFR50.72 and 10 CFR50.73. . . . The cause of this event is that the potential oxygen source and subsequent heating effects, of the hydrogen-oxygen recombination phenomenon were not properly considered during the original development of the LOCA,evaluation methodology. . An evaluation performed by GE based on a corrective action from Revision 0 of this LER has determined that there is sufficient conservatism in the Appendix K analysis such that it bounds the nominal Upper Bound PCT and maximum oxidation values (including the hydrogen-oxygen recombination phenomenon) at all exposures. Therefore, the original.SAFER/CORCL application methodology for conformance of the Appendix K analysis 10 CFR50.46 limits remains applicable. The hydrogen-oxygen recombination phenomenon does not need to be considered in the AppendikK analysis. This evaluation applies to the normal inerted containment condition. * As an Interim Corrective Action, an 8% Peak Linear Heat Generation Rate (P.LHGR) reduction is applied when the containment is allowed to be deinerted above 25% power. This constraint is currently being maintained, however, based on the above GE analysis, this reduction may be reduced or, if further evaluations relative to the basis for deinerted 'operation allow, may be eliminated.   +
05000219/LER-2004-004 +On September 1, 2004, at 13:50 (EDT), upon installation of a replacement Plant Computer System (PCS) (EllS-ID) Isolator Power Supply (EIIS-RJX), the PCS Heat Balance indicated 1936 MWth, which exceeded the licensed maximum power limit of 1930 MWth. Initial results of the prompt investigation indicated that the power supply began to degrade on August 29, 2004. This degradation resulted in a slow lowering of indicated Control Rod Drive (CRD) (EllS-AA) Flow, which provides input to the PCS Heat Balance calculation. Based on evaluations of the impact of the degradation, the 1930 MWth limit was exceeded on August 30, 2004 at 18:41. Power peaked at 1937 MWth, approximately 100.4%, on September 1, 2004. As a result, the thermal power limit was exceeded for greater than eight hours prior to recognition and actions taken to restore power below the license limit. The cause of this event was the slow degradation of the PCS isolator power supply, which resulted in a non-conservative heat balance calculation. All safety systems were fully operable and the safety significance of this event is considered minimal based on the small amount of deviation from the thermal power limit. There were no previous similar events of slowly degrading heat balance input parameters causing calculated core thermal power to lower resulting in operator actions to raise reactivity and consequently thermal power inadvertently above the license limit.  +
05000219/LER-2004-005 +'...0ri 'Satiiidil;Septerriber 11; 2004, with the plant in the Run Mode at 100% pdwer;_the oUtboard * Main - i Steam Isolation Valve (MSIV) NSO4 failed the10% MSIV closure teet.A 24 hOur TS LCO *waS entered at :; 0930:. To verify operability of the valVe,' plant power was reduced to-40% power and the MSIV fUll Closure' .' test was performed at1747. This test alsofailed.':At 1823 the inboard MSIVeNSO3A Was'isolated and :.: electroniCally ((laced in place due to the NSO4A being inoperable: With power maintained at apprOximately 40%,-a coMpleic troubleshooting plan was performed that concluded the valve was stuck open due : -:-, ' . performed_ . .. . internal problem. A plant shutdown was h Septembei' 14,' 2004 to repair the MSIV:- . _ Internal inSpeCtion of the vaiVe deterrniried the.failure of the poppet to close was caused by excessive rib :', * and Poppet guide wear resulting from poppet vibratiortinduced by steam flow * ' -. _ ,, , - „ , - . All safety systems were fully operable and the -safety significance of this event is considered minimal based on the demonstrated capability of the inbOard MSIV NSO3A to fully cloae:. ... . ., . : , , , No other MSIV failures caused by rib wear were identified. LER 88-013; occurred on July'9;1988 when ; inboard MSIV (NSO3A) failed the then quarterly full cloSure test due to stem failure caused by poppet - vibration.  +
05000219/LER-2004-006 +On Friday,' NOvernber 5;2004; withthe plant in :bild thutdOwn for refueling OUtage11320,:ihe as-found ' LoCal Leak Rate Test (LLRT) of Main Steam Isolation ValVe (MSIV) NSO4A failed to meet the ' acceptance criteria of Technical SpeCificatiOn 4.5.112. The aCceptance 'criteria was revised just prior to ' _ '1R20 and is now less than or equal to 11.9 SCFH and the actual measured leakage was 24.3 SCFH at : 20.6 psig. The valve was laSt refUrbished during forced outage 1F07 in Septernber 2004 and found to , be acceptable after maintenance that replaced the valve poppet:The valve was sucCessfUlly repaired in - : 1R20 and subsequentlypaesed the as-left LLRT.1:.„ . . .. _ The. apparent cause of the as-found LLRT:failure in 1 was irregulariti of the mating surfaces between: ' e the poppet seating surface and the valve bOdy seating surfaces. - . ' -...:. * ' Previous similar events of LLRT failures have Occurred in each of thelast two refueling outages:. . .. -, -  +
05000219/LER-2004-007 +On November 22, 2004 at 10:41 a.m. during startup operations following 1R20 Refueling Outage, a reactor operator (RO) bypassed the automatic Containment Ventilation and Purge Isolation (EIIS-JM) function due to an incorrect step in the plant start-up procedure. The incorrect step had the wrong nomenclature and directed the RO to a key-locked bypass control switch (CFI-HS) only intended to be used in implementing Emergency Operating Procedure (EOP) actions. The procedure guidance was reviewed and determined to be incorrect and the bypass switch realigned to normal after 1 Hour 41 Minutes. During the time that the switch was placed in bypass, the automatic isolation function of both V-27-1 and V-27-2 was bypassed and consequently the safety system function of these Primary Containment Isolation Valves to close and isolate in response to a LOCA signal was prevented. This condition is, by definition, a Safety System Functional Failure (SSFF) in accordance with NEI 99-02. The apparent cause of this event was an incorrectly worded step in the plant startup procedure. All other safety systems were fully operable and there were no circumstances that would have required automatic containment isolation while the function was bypassed. There were no previous similar events involving inadvertent bypassing of the containment isolation function due to inadequate procedures.  +
05000219/LER-2005-001 +During a functional test of the 1A2 Local Shutdown Panel (LSP) at 1254 on February 17, 2005, the "A" Control Rod Drive (CRD) Pump failed to start from the Main Control Room (MCR). The cause of this failure was attributed to contact high resistance on relay TR-2 in the 1A2 LSP. It was concluded that the high resistance cleared itself during subsequent relay operation. The relay was then scheduled for future replacement. "A" CRD Pump was successfully surveilled and returned to service. On March 16, 2005 at 1834, during a regular monthly surveillance, "A" CRD Pump did not start. Subsequent investigation found the 480 VAC breaker-closing spring was not charged. The closing spring is recharged by an electric motor immediately after the breaker opens. On March 18, 2005, investigation revealed a loose terminal wire connection to relay TR-2 within the LSP, which prevented charging of the closing spring. It is believed this condition existed on February 17, 2005 and should have been corrected. Technical Specifications (TS) only allow a 7-day out of service time for CRD Pumps resulting in a violation of TS 3.4.D. Corrective actions included replacing the relay, sending the relay out for failure analysis and performing verification of closing spring condition on all safety related 480 VAC breakers. This Supplemental LER is based on determination of the root causes to be a manufacturing deficiency consisting of a loose connection in the LSP, which caused the closing spring to not recharge, and a failure to detect that the closing spring was discharged prior to exceeding the Tech Spec out of service allotment. . There were no previous similar events at Oyster Creek Generating Station involving a breaker failing to close on demand due to the closing spring being discharged.  +
05000219/LER-2005-002 +On June 1, 2005, at 21:09, with Oyster Creek at 100% power, an Anticipatory Generator Load Reject scram occurred. During restoration of a transformer by the transmission utility at their substation, a failure of lightning arrestors resulted in a phase-to-phase-to-ground short circuit. This resulted in a grid transient of sufficient magnitude that the Oyster Creek Turbine-Generator sensed a load rejection condition, which resulted in a reactor scram signal. The reactor scrammed and the turbine-generator tripped as expected for this condition. All safety systems performed as expected. The plant was stabilized in the hot shutdown mode. Corrective actions included completing restart required evaluations, testing and confirmation from the involved transmission utility that conditions in the substation would not result in recurrence of the grid disturbance. The apparent cause of this event was equipment failure of the lightning arrestors in the transmission utility substation that created a grid disturbance. There have been several grid disturbances over the life of the plant, including two LERs: LER 2003-003, Actuation of Reactor Protection System Due to a Grid Transient (August 14, 2003). LER 1994-007 was a reactor scram caused by a 230 KV bus section differential relay trip while a switchyard worker was installing a Digital Fault Recorder.  +
05000219/LER-2005-005 +Oyster Creek Generating Station was in RUN at 100% power on 10/13/05 when a condition was discovered during routine laboratory as-found testing for Safety Valves (SVs) (EIIC: RV) removed during the 1R20 refueling Outage in November 2004. There were no structures, systems or components out of service that contributed to this event. In accordance with ASME Boiler & Pressure Vessel Code Criteria the nine SVs removed in refueling outage 1R20 were sent for as-found setpoint testing within the one-year time frame. Based on information received from the laboratory performing SV as-found testing, Site Engineering personnel determined that SV setpoint deficiencies existed with three SVs that were installed during the 1R19 refueling outage. Three (3) of the nine (9) valves exceeded the setpoint tolerance of +/-1% (+1-12 psig) as specified in the Technical Specifications paragraph 2.3F. All three SVs were within the ASME Code allowable +/-3% tolerance for as found values and there were no actual safety consequences associated with this event.  +
05000219/LER-2006-001 +On May 6, 2006, Oyster Creek was shutting down to start a Forced Outage (1F1 0) to repair a leak from the Steam Packing Exhauster Cooling Condenser (EIIS: SE). This leak was adding excess volume to the radiological liquid waste processing systems (Radwaste). During the shutdown, it was decided to perform a manual reactor scram to improve plant cool down and thereby minimize the volume of component leakage being sent to Radwaste (EIIS: WD). This manual SCRAM was not a scheduled item during the pre-outage planning. The decision to insert the manual scram was made during the shutdown process. It was not performed to avoid reaching an automatic scram setpoint. The 1F10 reactor shutdown was planned to be accomplished using full manual control rod insertion and manual scram was not a planned activity. The manual scram was performed after reducing reactor power level to the Source Range and the reactor was subcritical at the time. All Control Rods fully inserted as a result of the manual scram, and all systems performed as designed.  +
05000219/LER-2006-002 +Oyster Creek performed a full closure test of main steam isolation valve NSO4A (outboard MSIV on 'A' main steam line) during the 1F09 forced outage in February 2006. During this test the valve did not stroke closed in the allowable time specified in the surveillance procedure. When measured using the plant process computer (PPC), the valve stroked full closed in 2.0 seconds. The acceptance criteria for this timing test is 3.0 seconds. The failure of this surveillance test was not recognized at the time of performance and as a result the plant was started up from 1F09 with an inoperable MSIV. During the later 1F10 forced outage in May 2006 the valve was tested again under the surveillance procedure and failed to meet the acceptance criteria for the PPC timing. The valve therefore failed the surveillance test and was adjusted and retested acceptably prior to startup in May 2006. The discovery date of the failed surveillance tests was June 20, 2006.  +
05000219/LER-2006-003 +On Monday, October 16, 2006, `with the plant in'Cold Shutdown for refueling outage 1R21., the as-found Local Leak Rate Test (LLRT) of inboard Main Steam Isolation Valve (MSIV) NSO3B,failed to meet the acceptance criteria ofTechriical,SpecificatiOn 4.5.b.2. , No as-found data could be obtained on this valve - due to failure to m6intaitv'pressure. The valve was last refurbished duringrefueling outage 1R18 in October 2000, when- a pilot poppet modification and new stem were installed:,Following maintenanCe,'in. ,1R18 the valve passed the as-left LLRT. , The apparent cause of the as-found LLRT failure in 1R21 was attributed to an oxide layer build up on the valve poppet seat and in-body seat ring. The safety significance of this event is considered minimal. The leakage oast,NSO3B would have been limited by the leak rate of the outboard MSIV (NSO4B) in the same header which met the LLRT acceptance criteria of, Technical Specification 4.5.D.2 when tested in 1R21,. This, leakage provideS adequate margin betWeen projected potential offsite doe and 10 CFR 100 guidelines. , '‘ ,  +
05000219/LER-2006-004 +. .�„ An, Uhplanned,reactor power increase to 162:46'% of rated thermal power, occurred due: to opening and,,,, re-closing of one of the five Electromatic Relief Valves (EMRVs).. After being open for approximately, ,, 57. seconds, the EMRV re-closed without operator action. Subsequent to the EMRV closure, the'" reactor powerincreased to 1:02.46%o and Was immediatelY'4dUcectto less than 100%-as'a result of action to degrease the redrdulatiOn fiOw rate. The most PrObable'caus‘e‘ofthit event was the malfunctibm'of the Preasure switch connectecttO the subject EMRV. The -preSsure switch was . subsequently replaced  +
05000219/LER-2007-001 +On July 17, 2007 at 05:21 while operating at 100% power, an automatic reactor scram occurred due to low reactor water level following a trip of the "C" Reactor Feed Pump (RFP). The cause of the "C" RFP trip is attributed to an electrical fault internal to the motor. This transient led to an automatic scram on * low reactor water level and subsequent reactor isolation on a low-low reactor level. There were no safety consequences impacting plant or public safety as a result of this event. . This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to the automatic reactor protection system and subsequent ECCS actuations.  +
05000219/LER-2007-002 +' On July 20, 2007, a plant startup from. forced outage 1F12 commenced with InterMediate Range Monitor (IRM) 1.7 in the bypaSsed mode. At 14:40, operators noted that IRM'1 p was not responding to an increase in reactor power.: Both. IRM 16 and IRM 17 are part of Reactor Protection System (RPS) Trip System 2: Three of four IRMs in each of two Trip Systems aresequired * to be operable by. Technical Specifications for the Startup Mode. , .. * , * The apparent cause *for IRM-16 failure is a failureof the detector. Corrective actions included replacing. the ...„.. failed detector,... notifying the industry of the failure, and developing a plan to restore all nuclear instrumentation to an (a)(2) Maintenance Rule. status. ... .. There' were no safety consequences impacting plant or public.safety as a result of this event: .. * *This event is .being *reported pursuant to 10 CFR 50.73(a)(2)(i)(B) 011. e to the station being in a condition prOhibited by *Technical Specifications Table 3.1.1; Protective Instrumentation Requirements, Function 9; * , . * Intermediate .Rarige Monitor...  +