U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation
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U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation
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The organization U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation represents an institution, an association, or corporate body that is associated with resources found in University of Missouri-Kansas City Libraries.
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- U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation
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936 Items by the Organization U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation
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- Generic environmental impact statement for license renewal of nuclear plants : supplement 54, regarding Byron Station, units 1 and 2 : draft report for comment
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 55, Regarding Braidwood Station, Units 1 and 2 | Draft report for comment
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 57, Regarding LaSalle County Station, units 1 and 2; final report
- Standard review plans for environmental reviews of nuclear power plants : supplement 1: operating license renewal, final report
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, unit 2, docket number 50-391, Tennessee Valley Authority
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 38, Regarding Indian Point Nuclear Generating Unit nos. 2 and 3 ; draft report for comment
- Draft final environmental statement, Supplement 2, related to the operation of Watts Bar Nuclear Plant, unit 2, docket number 50-391, Tennessee Valley Authority, draft report for comment
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, docket nos. 50-390 and 50-391 : Tennessee Valley Authority
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, docket nos. 50-390 and 50-391 : Tennessee Valley Authority
- Generic environmental impact statement for license renewal of nuclear plants : regarding Monticello nuclear generating plant, draft report, Supplement 26
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, docket nos. 50-390 and 50-391 : Tennessee Valley Authority
- Final environmental statement, related to the operation of Watts Bar Nuclear Plant, unit 2, supplement 2 : final report
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 54, Regarding Byron Station, Units 1 and 2 | Final report
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 57, Regarding LaSalle County Station, units 1 and 2 | Draft report for comment
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 53, Regarding Sequoyah nuclear plant, Units 1 and 2 | Final report
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, docket nos. 50-390 and 50-391 : Tennessee Valley Authority
- Environmental impact statement for the construction permit for the SHINE medical radioisotope production facility : final report
- Environmental impact statement for the construction permit for the SHINE medical radioisotope production facility : draft report for comment
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 56, Regarding Fermi 2 Nuclear Power Plant | Draft report for comment
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 55, Regarding Braidwood Station, units 1 and 2 | Final report
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 52, Regarding Davis-Besse nuclear power station | Final report
- Generic environmental impact statement for license renewal of nuclear plants, Supplement 46, Regarding Seabrook Station | Final report
- Environmental impact statement for license renewal of the National Bureau of Standards Reactor : final report
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Unit 2, docket no. 50-391 : Tennessee Valley Authority
- Generic environmental impact statement for license renewal of nuclear plants, supplement 50, regarding Grand Gulf Nuclear Powerplant, unit 1 : draft report for comment
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- Cable damage caused by inadequate cable installation procedures and controls
- Capacitor failures in Westinghouse Eagle 21 plant protection systems
- Cavitation erosion of letdown line orifices resulting in fatigue cracking of pipe welds
- Cement erosion from containment subfoundations at nuclear power plants
- Centrifugal charging pump thrust bearing damage not detected due to inadequate assessment of oil analysis results and selection of pump surveillance points
- Challenges to licensees' ability to provide prompt public notification and information duriing an emergency preparedness event
- Changes in the operator licensing program
- Charging pump trip during a loss-of-coolant event caused by low suction pressure
- Circuit breakers left racked out in non-seismically qualified positions
- Circumferential cracking in the stainless steel pressurized heater sleeves of pressurized water reactors
- Circumferential cracking of steam generator tubes
- Cofrentes NPP (BWR/6) ATWS (MSIVC) analysis with TRAC-BF1 : 1D vs. point kinetics and containment response
- Combustibility of epoxy floor coatings at commercial nuclear power plants
- Commission decision on the resolution of generic issue 23 " Reactor coolant pump seal failure"
- Common-cause failures due to inadequate design control and dedication
- Compensatory measures for prolonged periods of security system failures
- Concerns regarding essential chiller reliability during periods of low coolling water temperature
- Configuration control errors
- Consequences of using soluble dams
- Consideration of the stem rejection load in calculation of required valve thrust
- Consideration of thermal conditions in the design and installation of supports for diesel generator exhaust silencers
- Construction experience related to the assurance of quality in the construction of nuclear facilities
- Construction experience with concrete placement
- Construction-related experience with cables, connectors, and junction boxes
- Construction-related experience with flood protection features
- Containment concrete surface condition examination frequency and acceptance criteria
- Containment inspection rule
- Containment liner corrosion
- Contaminated radiography source shipments
- Contamination of breathing air systems
- Continuous communications following emergency notifications
- Control of hot particle contamination at nuclear power plants
- Control rod insertion problems
- Control room habitability
- Control room staffing below minimum regulatory requirements
- Conviction for falsification of security training records
- Core shroud cracking at beltline region welds in boiling-water reactors
- Corrosion of William Powell gate valve disc holders
- Corrosion of steel containment and containment liner
- Counterfeit parts supplied to nuclear power plants
- Counterfeit valves in the commercial grade supply system
- Crack in weld area of reactor coolant system hot leg piping at V.C. Summer
- Cracked insulators in ASL dry type transformers manufactured by Westinghouse Electric Corporation
- Cracking in pressurizer safety and relief nozzles and in surge line nozzle
- Cracking in the lower region of the core shroud in boiling water reactors
- Cracking indications in thermally treated alloy 600 steam generator tubes
- Cracking of reactor vessel internal baffle former bolts in foreign plants
- Cracking of surge ring brackets in large General Electric Company electric motors
- Cracking of valves in the condensate return lines of a BWR emergency condensation system
- Cracking of vertical welds in the core shroud and degraded repair
- Cracks found in jet pump riser assembly elbows at boiling water reactors
- Crediting of operator actions in place of automatic actions and modifications of operator actions, including response times
- Criminal history record information
- Criminal prosecution and conviction of wrongdoing commited by a commercial-grade valve supplier
- Criminal prosecution of nuclear suppliers for wrongdoing
- Criticality monitoring system degradation at BWX Technologies, Inc., Nuclear Products Division, Lynchburg, VA
- Crosby relief valve setpoint drift problems caused by corrosion of the guide ring
- Cross-tied safety injection accumulators
- Current fire endurance test results for thermo-lag fire barrier material
- DOT safety advisory : high pressure aluminum seamless and aluminum composite hoop-wrapped cylinders
- Damage in foreign steam generator internals
- Debris in containment and the residual heat removal system
- Decay heat management practices during refueling outages
- Defacing of labels to comply with 10 CFR 20.1904(b)
- Defect in safety-related pump parts due to inadequate heat treatment
- Defective molded phenolic armature carriers found in Elmwood contactors
- Defective worm shaft clutch gears in Limitorque motor-operated valve actuators
- Recent loss or severe degradation of service water systems
- Additional adverse effect of boric acid leakage : potential impact on post-accident coolant ph
- Age-related constant support degradation
- Air actuator and supply air regulator problems in Copes-Vulcan pressurizer power-operated reliel valves
- Air entrainment in terry turbine lubricating oil system
- Air systems problems at U.S. light water reactors
- Alleged falsification of certifications and alteration of markings on piping, valves, and fittings
- American Power Service falsification of American Society for Nondestructive Testing (ASNT) certificates
- Anomalous behavior of recirculation loop plow in jet pump BWR plants
- Applicability of 10 CFR Part 21 to nonlicensees
- Applying statistics
- Water leakage from yard area through conduits into buildings
- Assessment of spent fuel pool cooling
- Assurance of equipment operability and containment integrity during design-basis accident conditions
- Assurance of sufficient net positive suction head for emergency core cooling and containment heat removal pumps
- Auxiliary feedwater pump recirculation line orifrice fouling-potential common cause failure
- Auxiliary feedwater pump trips resulting from low suction pressure
- Availability of alternate AC power source designed for station blackout event
- Axial outside-diameter cracking affecting thermally treated alloy 600 steam generator tubing
- B&W once-through steam generator tube inspection findings
- BWR operation with indicated flow less than natural circulation
- Backseating valves routinely to prevent packing leakage
- Barriers and seals between mild and harsh environments
- Biodiesel in fuel oil could adversely impact diesel engine performance
- Boiling water reactor licensees use of the BWRVIP-05 report to request relief from augmented examination requirements on reactor pressure vessel circumferential shell welds
- Boraflex degradation in spent fuel pool storage racks
- Boric acid corrosion of charging pump casing caused by cladding cracks
- Braidwood Unit 1 reactor trip due to off-site power fluctuation
- Buildup of deposits in steam generators
- Bursting of high pressure coolant injection steam line rupture discs injures plant personnel
- Problems with the latching mechanism in Potter and Brumfield R10-E3286-2 relays
- Problems with the use of unshielded test leads in reactor protection system circuitry
- Proceedings of the ... NRC/ASME Symposium on Valve and Pump Testing
- Proceedings of the ... NRC/ASME Symposium on Valves, Pumps, and Inservice Testing
- Proceedings of the Tenth NRC/ASME Symposium on Valves, Pumps and Inservice Testing : held at L'Enfant Plaza Hotel Washington, DC, July 14-16, 2008
- Prodiuct alert for fire hydrants
- Production and testing of the VITAMIN-B7 fine-group and BUGLE-B7 broad-group coupled neutron/gamma cross-section libraries derived from ENDF/B-VII.O nuclear data
- Pump shaft damage due to excessive hardness of shaft sleeve generator tubes
- Questionable selection and review to determine suitability of electropneumatic relays for certain applications
- RCP seal replacement with pump on backseat
- RELAP5/MOD3 computer code error associated with the conservation of energy equation
- Radial cracking of shroud support access hole cover welds
- Radiation beams from power reactor biological shields
- Radioactive effluents from nuclear power plants
- Radiography events at operating power reactors
- Rapid flow-induced erosion/corrosion of feedwater piping
- Reactivity insertion transient and accident limits for high burnup fuel
- Reactor coolant pressure boundary leakage attributable to propagation of cracking in reactor vessel nozzle welds
- Reactor coolant pump lube oil fire
- Reactor license renewal : preparing for tomorrow's safety today
- Reactor license renewal : preparing for tomorrow's safety today
- Reactor operation inconsistent with the updated final safety report
- Reactor trip breaker, Westinghouse model DS-416, failed to open on manual initiation from the control room
- Reactor trip breakers and surveillance testing of auxiliary contacts
- Reactor trip followed by unexpected events
- Reactor trips caused by breaker testing with fault protection bypassed
- Reactor vessel top guide and core plate cracking
- Reassessment of NRC's dollar per person-rem conversion factor policy : draft report for comment
- Recall of Star brand fire protection sprinkler heads
- Recent design problems in safety functions of pneumatic systems
- Recent events involving reactor coolant system inventory control during shutdown
- Recent experience with degradation of reactor pressure vessel head
- Recent experience with plugged steam generator tubes
- Recent experience with reactor coolant system leakage and boric acid corrosion
- Recent failures of charging/safety injection pump shafts
- Recent fires at commercial nuclear power plants in the United States
- Recent foreign and domestic esperience with degradation of steam generator tubes and internals
- Recent fuel and core performance problems in operating reactors
- Recent fuel handling events
- Recent human performance issues at nuclear power plants
- Recent issues associated with NRC medical requirements for licensed operators
- Accuracy of motor-operated valve diagnostic equipment manufactured by Liberty Technologies
- Recent operating experience associated with pressurizer and main steam safety/relief valve lift setpoints
- Recent operating experience concerning hydrostatic barriers
- Recent operating experience of service water systems due to external conditions
- Recent operator performance issues at nuclear power plants
- Recent plant events caused by human performance errors
- Recent problems with overhead cranes
- Reconsideration of nuclear power plant security requirements associated with an internal threat
- Recurring events involving emergency diesel generator operability
- Reduced service life of Automatic Switch Company (ASCO) solenoid valves with Buna-N material
- Reliability of ATWS mitigation systems and other NRC-required equipment not controlled by plant technical specifications
- Relocation of selected technical specifications requirements related to instrumentation
- Relocation of the pressure temperature limit curves and low temperature overpressure protection system limits
- Remote trip function failures in General Electric f-frame molded-case circuit breakers
- Removal of a fuel element from a research reactor core while critical
- Removal of accelerated testing and special reporting requirements for emergency diesel generators
- Repeated multiple failures of steam generator hydraulic snubbers due to control valve sensitivity
- Repetitive overspeed tripping of turbine-driven auxiliary feedwater pumps
- Reporting of errors and changes in large-break loss-of-coolant accident evaluation models of fuel vendors and compliance with 10 CFR 50.46(a)(3)
- Requests to dispose of very low-level radioactive waste pursuant to 10 CFR 20.302
- Requirements for steam generator tube inspections
- Requirements in 10 CFR Part 21 for reporting and evaluating software errors
- Respirator coupling nut assembly failures
- Response to indications of potential tampering, vandalism, or malicious michief
- Resultls of chemical effects head loss tests in a simulated PWR sump pool environment
- Results of a special NRC inspection at Dresden Nuclear Power Station Unit 1 following a rupture of service water inside containment
- Results of shift staffing study
- Results of steam generator tube examinations
- Results of thermo-lag 330-1 combustibility testing
- Results of validation testing of motor-operated valve diagnosis equipment
- Review of industry responses to NRC generic letter 97-06 on degradation of steam generator internals
- Review of refueling outage risk
- Revised contents of the monthly operating report
- Revised international nuclear and radiological event scale user's manual
- Revised protection action guidance for nuclear incidents
- Revised seismic hazard estimates
- Risk impact study regarding maintenance during low-power operation and shutdown
- Rupture in extraction steam piping as a result of flow-accelerated corrosion
- Rupture of the shell side of a feedwater heater at the Point Beach Nuclear Plant
- Safe shutdown potentially challenged by unanalyzed internal flooding events and inadequate design
- Safety culture common language
- Safety evaluation report related to the license renewal of Beaver Valley Power Station, Units 1 and 2 : dockets nos. 50-334 and 50-412, FirstEnergy Nuclear Operating Company
- Safety evaluation report related to the license renewal of Byron Station, units 1 and 2, and Braidwood Station, units 1 and 2 : docket nos. 50-454, 50-455, 50-456, and 50-457, Excelon Generation Company, LLC
- Safety evaluation report related to the license renewal of Callaway Plant, Unit 1 : docket number 50-483, Union Electric Company (Ameren Missouri)
- Safety evaluation report related to the license renewal of Davis-Besse Nuclear Power Station : supplement 1, docket nos. 50-346, FirstEnergy Nuclear Operating Company
- Safety evaluation report related to the license renewal of Duane Arnold Energy Center
- Safety evaluation report related to the license renewal of Hope Creek Generating Station : docket no. 50-354 : PSEG Nuclear, LLC
- Safety evaluation report related to the license renewal of Indian Point Nuclear Generating Unit nos. 2 and 3 : docket nos. 50-247 and 50-286, Entergy Nuclear Operations, Inc
- Safety evaluation report related to the license renewal of Indian Point Nuclear Generating Unit nos. 2 and 3 : supplement 1, docket nos. 50-247 and 50-286, Entergy Nuclear Operations, Inc
- Safety evaluation report related to the license renewal of Indian Point Nuclear Generating Unit nos. 2 and 3 : supplement 2, docket nos. 50-247 and 50-286
- Safety evaluation report related to the license renewal of Kewaunee Power Station
- Safety evaluation report related to the license renewal of Limerick Generating Station, Units 1 and 2 : Docket Nos. 50-352 and 50-353
- Safety evaluation report related to the license renewal of Limerick Generating Station, Units 1 and 2 : supplement 1 : docket nos. 50-352 and 50-353
- Safety evaluation report related to the license renewal of Palo Verde Nuclear Generating Station, Units 1, 2 and 3 : docket nos. 50-528 and 50-529, and 50-530, Arizona Public Service Company
- Safety evaluation report related to the license renewal of Prairie Island Nuclear Generating Plant units 1 and 2
- Safety evaluation report related to the license renewal of Salem Nuclear Generating Satation
- Safety evaluation report related to the license renewal of Sequoyah Nuclear Plant Units 1 and 2 : Docket numbers 50-327 and 50-328
- Safety evaluation report related to the license renewal of Sequoyah Nuclear Plant Units 1 and 2 : Docket numbers 50-327 and 50-328
- Safety evaluation report related to the license renewal of Shearon Harris Nuclear Power Plant, Unit 1, docket no. 50-400 : Carolina Power & Light Company
- Safety evaluation report related to the license renewal of Vermont Yankee Nuclear Power Station, docket no. 50-271 : Entergy Nuclear Operations, Inc
- Safety evaluation report related to the license renewal of Vermont Yankee Nuclear Power Station, docket no. 50-271, Entergy Nuclear Operations, Inc
- Safety evaluation report related to the license renewal of Vogtle Electric Generating Plant, Units 1 and 2 : docket nos. 50-424 and 50-425, Southern Nuclear Operating Company, Inc
- Safety evaluation report related to the license renewal of Wolf Creek Generating Station : docket no. 50-482, Wolf Creek Nuclear Operating Corporation
- Safety evaluation report related to the license renewal of the Browns Ferry Nuclear Plant, units 1, 2, and 3 : docket Nos. 50-259, 50-260, and 50-296, Tennessee Valley Authority
- Safety evaluation report related to the operation of Susquehanna Steam Electric Station, Units 1 and 2 : docket nos. 50-387 and 50-388, PPL Susquehanna, LLC
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, unit 2
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, unit 2, docket number 50-391, Tennessee Valley Authority
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, unit 2, docket number 50-391, Tennessee Valley Authority
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, unit 2, docket number 50-391, Tennessee Valley Authority
- inadequate main steam safety valve (MSSV) setpoints and performance issues associated with long MSSV inlet piping
- Safety evaluation report related to the operation of Watts Bar Nuclear Plant, unit 2, docket number 50-391, Tennessee Valley Authority
- Safety evaluation report, related to the license renewal of Columbia Generating Station
- Safety injection system weld flaw at Sequoyah Nuclear Power Plant, Unit 2
- Safety system problems caused by modifications that were not adequately reviewed and tested
- Safety system response to loss of coolant and loss of offsite power
- Safety-related equipment failures caused by faulted indicated lamps
- Secondary piping rupture at the Mihama Power Station in Japan
- Seismic adequacy of Thermo-Lag panels
- Service water system degradation at Brunswick Steam Electric Plant, Unit 1
- Settlement monitoring and inspection of plant structures affected by degradation of porous concrete subfoundations
- Shaft binding in General Electric type SBM control switches
- Shielding deficiency in spent fuel transfer canal at a boling water reactor
- Shutdown order issued because licensed operators asleep while on duty
- Significant loss of safety-related electrical power at Forsmark, Unit 1, in Sweden
- Significant unexpected erosion of feedwater lines
- Single failure vulnerability of engineered safety features activation systems
- Single failures in auxiliary feedwater systems
- Single-failure and fire vunerability of redundant electrical safety buses
- Slow five percent scram insertion times caused by viton diaphragms in scram solenoid pilot valves
- Small arms firing range safety issues
- Snubber lubricant degradation in high-temperature environments
- Software problems involving digital control console systems at non-power reactors
- Solid state protection system card failure results in spurious safety injection actuation and reactor trip
- Spent fuel pool leakage to onsite groundwater
- Spent fuel pool reactivity calculations
- Spent fuel rod accountabilitiy
- Spurious relay actuations result in loss of power to safeguards buses
- Spurious shutdown of emergency diesel generators from design oversight
- Spurious tripping of low-voltage power circuit breakers with GE RMS-9 digital trip units
- Standard review plan for review of license renewal applications for nuclear power plants, final report
- Standard review plan for review of subsequent license renewal applications for nuclear power plants : draft report for comment
- Standard review plans for environmental reviews of nuclear power plants : supplement 1: operating license renewal, draft report for comment
- Standard review plans for environmental reviews of nuclear power plants : supplement 1: operating license renewal, final report
- inadvertent reactor trip and partial safety injection actuation due to tin whisker
- Standard technical specifications, Babcock and Wilcox plants
- Standard technical specifications, General Electric BWR/6 plants
- Status of NRC's staff review of BWRVIP-05
- Status of motor-operated valve performance prediction program by the Electric Power Research Institute
- Steam generator tube and support configuration
- Steam generator tube degradation at Diablo Canyon
- Steam generator tube end cracking
- Steam generator tube failure at Indian Point Unit 2
- Steam generator tube inspection techniques
- Steam generator tube integrity and associated technical specifications
- Stem binding in turbine governor valves in reactor core isolation cooling (RCIC) and auxiliary feedwater (AFW) systems
- Storm-related loss of offsite power events due to salt buildup on switchyard insulators
- Study on post tensioning methods
- Submerged safety-related electrical cables
- Substandard material supplied by Chicago Bullet Proof Systems
- Suceptibility of low-pressure coolant injection and core spray injection valves to pressure locking
- Summary of NRC staff observations compiled during engineering audits or inspections of licensee erosion/corrosion programs
- Summary of fitness for duty program performance reports for calendar year 2008
- Summary of fitness-for-duty program performance reports for calendar year 2000
- Summary of fitness-for-duty program performance reports for calendar year 2006
- Summary of fitness-for-duty program performance reports for calendar year 2007
- Summary of fitness-for-duty program performance reports for calendar years 1996 and 1997
- Summary of fitness-for-duty program performance reports for calendar years 1998 and 1999
- Summary of fitness-for-duty program performance reports for calendar years 2001, 2002 and 2003
- Summary of fitness-for-duty program performance reports for calendar years 2004 and 2005
- Summary of licensed operator requalification inspection program findings
- Supplemental information to Generic Letter 95-03 "Circumferential cracking of steam generator tubes"
- Susceptibility of containment sump recirculation gate valves to pressure locking
- Switchgear fire and partial loss of offsite power at Waterford Generating Station, Unit 3
- Switchover to hot-leg injection following a loss-of-coolant accident in pressurized water reactors
- TMI-2 Lessons Learned Task Force final report
- Tactical communications interoperability between nuclear power reactors licensees and first responders
- Technical bases for changes in the subsequent license renewal guidance documents NUREG-2191 and NUREG-2192
- Technical basis for regulatory guidance on the alternate pressurized thermal schock rule : draft report for comment
- Temporary scaffolding affects operability of safety-related equipment
- Ten-year inservice inspection (ISI) program update for licensees that intend to implement risk-informed ISI of piping
- Testing of safety-related logic circuits
- The United States of America, national report for the Convention on Nuclear Safety
- The effect of the year 2000 issue on medical licensees
- The importance of accurate inventory controls to prevent the unauthorized possession of radioactive material
- Themally induced pressurization of nuclear power facility piping
- Thermal fatigue cracking of feedwater piping to steam generators
- Thermal stratification of water in BWR reactor vessels
- Thermally induced accelerated aging and failure of ITE/Gould A.C. relays used in safety-related applications
- Thermally induced pressure locking of a high pressure coolant injection valve
- Thermo-Lag Fire Barrier Material Special Review Team report findings, current fire endurance tests, and ampacity calculation errors
- Thermo-lag 330-1 flame spread test results
- Thermo-lag 330-660 flexi-blanket ampacity derating concerns
- Thimble tube thinning in Westinghouse reactors
- Three-unit trip and loss of offsite power at Palo Verde Nuclear Generating Station
- Through-wall circumferential cracking of reactor pressure vessel head control rod drive mechanism penetration nozzles at Oconee Nuclear Station, Unit 3
- Thrust limits for Limitorque actuators and potential overstressing of motor-operated valves
- Torus cracking in a BWR mark 1 containment
- Transformer failures--recent operating experiience
- Transient resulting in a reactor trip and multiple safety injection system actuations at Salem
- Traversing in-core probe withdrawn at LaSalle County Station, Unit 1
- Treatment planning system errors result in deaths of overseas radiation therapy patients
- Tripping of Klockner-Moeller molded-case circuit breakers due to support lever failure
- Turbine blade failures caused by torsional excitation from electrical system disturbance
- Turbine overspeed and reactor cooldown during shutdown evaluation
- Turbine-driven auxiliary feedwater pump bearing issues
- U.S. Food and Drug Administration announcement related to certain sleep disorder drugs
- U.S. operating experience with thermally treated alloy 600 steam generator tubes through December 2013
- Unanalyzed condition of reactor coolant pump seal leakoff line during postulated fire scenarios or station blackout
- Unanticipated and unintended movement of fuel assemblies and other components due to improper operation of refueling equipment
- Unanticipated crack in a particular heat of alloy 600 used for Westinghouse mechanical plugs for steam generator tubes
- Unanticipated effect of ventilation system on tank level indications and engineering safety features actuation system setpoint
- Unanticipated reactor water drainage draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, and Fitzpatrick
- Unauthorized forced entry into the protected area at Three Mile Island Unit 1 on February 7, 1993
- Unauthorized us of license to obtain radioactive materials, and its implications under the expanded Title 18 of the U.S. Code
- Uncontrolled rod withdrawal because of a single failure
- Undervoltage protection relay settings out of tolerance due to test equipment harmonics
- Undervoltage relay and thermal overload setpoint problems
- Undetected accumulation of gas in reactor coolant system
- Undetected accumulation of gas in reactor coolant system and inaccurate reactor water level indication during shutdown
- Undetected loss of reactor coolant
- Undetected modification of flow characteristics in the high pressure safety Injection system
- Undocumented changes to non-power reactor safety system wiring
- Unescorted access granted on the basis of incomplete and/or inaccurate information
- Unexpected degradation of lead storage batteries
- Unexpected opening of a safety relief valve and complications involving suppression pool cooling strainer blockage
- Unexpected opening of both doors in an airlock
- Unexpected opening of multiple safety relief valves
- Unexpected plant performance during performance of new surveillance tests
- Unexpected restriction to thermal growth of reactor coolant piping
- Unisolable crack in high-pressure injection piping
- Unisolatable reactor system coolant leak following repeated applications of leak sealant
- Unplanned intakes by workers of transuranic airborne radioactive materials and external exposure due to inadequate control of work
- Unplanned intakes of airborne radioactive material by individuals at nuclear power plants
- Unplanned return to criticality during reactor shutdown
- Unplanned, undetected release of radioactivity from the exhaust ventilation system of a boiling water reactor
- Unqualified butt splice connectors identified in qualified penetrations
- Unrecognized loss of control room annunciators
- Unrecognized reactivity addition during plant shutdown
- Urine specimen adulteration
- Use of NUMARC/EPRI TR-102348, guideline on licensing digital upgrades in determining the acceptability of performing analog-to-digital replacements under 10 CFR 50.59
- Use of blanl ammunition
- Use of galvanized supports and cable trays with Meggitt si 2400 stainless-steel-jacketed electrical cables
- Use of inappropriate guidelines and criteria for nuclear piping and pipe support evaluation and design
- Use of inappropriate lubrication oils in safety-related applications
- Use of less than optimal bounding assumptions in criticality safety analysis at fuel cycle facilities
- Use of nonconservative acceptance criteria in safety-related pump surveillance tests
- Use of sodium hypochlorite for cleaning diesel fuel oil supply tanks
- Validation of SCALE 5 decay heat predictions for LWR spent nuclear fuel
- Valve-stem failure caused by embrittlement
- Vertical deep draft pump shaft and coupling failures
- Vibration caused by increased recirculation flow in a boiling water reactors
- Vibration-induced degradation and failure of safety-related valves
- Vibration-induced degradation of butterfly valves
- Voltage based repair criteria for Westinghouse steam generator tubes affected by outside diameter stress corrosion cracking
- Vulnerability of emergency diesel generators to fuel oil/lubricating oil incompability
- WASH-1400, the reactor safety study : the introduction of risk assessment to the regulation of nuclear reactors
- Water in the vent header/vent line spherical junctions
- 10 CFR part 52 construction inspection program framework document