Abstract: A nuclear fuel rod includes a cladding comprising an interior region. A pressure transmission apparatus is configured to measure pressure in the interior region. Fuel pellets and a resonant electrical circuit are in the interior region. The resonant electrical circuit is configured to generate a pulse that travels wirelessly from the interior region through the cladding. A frequency of the pulse varies based on pressure measured by the pressure transmission apparatus. The frequency of the pulse is indicative of the pressure within the interior region.

Abstract: A system includes a containment vessel configured to prohibit a release of a coolant, and a reactor vessel mounted inside the containment vessel. An outer surface of the reactor vessel is exposed to below atmospheric pressure, wherein substantially all gases are evacuated from the containment vessel.

Abstract: A system that provides a direct indication of peak fuel rod centerline temperature and peak fuel rod clad temperature than conventionally inferred from the power distribution by directly and continuously measuring the fuel temperatures of the fuel pellets in one or more of the hottest fuel elements in the core. The peak fuel rod clad temperature is then obtained from the maximum measured peak fuel rod centerline temperature in combination with the maximum coolant core exit temperature and the minimum coolant flow rate.

Abstract: A fast-neutron nuclear reactor fuel assembly having fuel rods. Each fuel rod has nuclear fuel disposed in a sealed housing in the form of a tubular steel shell and end parts. A steel spacer element is wound in a coil with a specific pitch on the outside surface of the shell and is fastened on the end parts. The spacer element is in the form of a metallic band twisted around its longitudinal axis. The width of said band is approximately equal to the minimum distance between adjacent fuel rods in the fuel assembly. A transverse cross-sectional area of the band is within a range from 0.10 to 0.50 times the area of a circle described around the width of the band.

Abstract: A defective fuel bundle location system for use with a heavy water moderated nuclear fission reactor having a fueling machine, including a test tool defining an internal volume, the test tool being configured to be received within both the fueling machine and a corresponding fuel channel of the reactor, and a test container defining an internal volume, wherein the test container is configured to be received within the internal volume of the test tool and the internal volume of the test container is configured to receive primary fluid from the reactor when the test tool is disposed within the corresponding fuel channel of the reactor.

Abstract: Disclosed are a packaging cover plate, an organic light-emitting diode display and a manufacturing method therefor. The packaging cover plate comprises: a cover plate body, the cover plate body being provided with open slots; cover plugs for covering openings at two ends of the open slots; and water absorption layers for at least covering mouths of the open slots. By means of the arranged open slots, the packaging cover plate can conveniently introduce a dry gas. In addition, the water absorption layers absorb water vapour, the introduced dry gas dries the water absorption layers, and the water vapour and oxygen in the water absorption layers can be taken away by means of the circular flow of the dry gas, so that damage to a device by water vapour and oxygen can be reduced, and the packaging effect is better.

Abstract: A method that provides a more direct indication of peak fuel rod centerline temperature and peak fuel rod clad temperature than conventionally inferred from the power distribution by directly and continuously measuring the fuel temperatures of the fuel pellets in one or more of the hottest fuel elements in the core. The peak fuel rod clad temperature is then obtained from the maximum measured peak fuel rod centerline temperature in combination with the maximum coolant core exit temperature and the minimum coolant flow rate.

Abstract: A debubbling sleeve for a fluid sensor of a sensor system is disclosed. The debubbling sleeve has a receptacle closed circumferentially around a sleeve bottom and a first passageway extending through the sleeve bottom. The receptacle contains the fluid sensor.

Abstract: A valve assembly includes a flange connected to a vessel penetration of a reactor pressure vessel of a nuclear reactor. A valve is disposed inside the flange or protrudes from the flange into the vessel penetration. The valve includes a valve seat and a movable valve member positioned so that pressure inside the reactor pressure vessel urges the movable valve member against the valve seat to close the valve. The valve assembly further includes a plenum having an inlet via which the plenum can be pressurized to apply pressure to the movable valve member that urges the movable valve member away from the valve seat to open the valve. The plenum may be defined in part by a surface of the movable valve member. The valve assembly preferably does not include a valve actuator.

Abstract: A system and method for drying cavities containing spent nuclear fuel is devised. The invention utilizes a non-intrusive procedure that is based on monitoring the dew point temperature of a non-reactive gas that is circulated through the cavity. In one aspect, the invention is a system for drying a cavity loaded with spent nuclear fuel comprising: a canister forming the cavity, the cavity having an inlet and an outlet; a source of non-reactive gas; means for flowing the non-reactive gas from the source of non-reactive gas through the cavity; and means for repetitively measuring the dew point temperature of the non-reactive gas exiting the cavity.

Abstract: A nuclear reactor instrumentation system monitors a nuclear power system that includes a reactor core within a reactor vessel and one or more sensors for monitoring parameters of the nuclear power system. The nuclear reactor instrumentation system includes a computer having a processor and configured to be powered by a normal power source and a backup power source; a wireless transmitter operable under the control of the processor; and a memory coupled to the processor and containing stored programming instructions. The stored programming instructions are executable by the processor to cause the processor to receive data from the sensors; identify a loss of normal power from the normal power source; and in response to identifying the loss of normal power, cause the wireless transmitter to transmit the received data.

Abstract: A nuclear reactor plant including a reactor; a coolant in the reactor; a gas system connected to the reactor and adapted to provide a gas supply to and gas removal from a space above the coolant; and a device for injection of gas into the coolant. The device is installed partially in the coolant and partially in the space above the coolant, and is adapted to supply gas from the space above the coolant to the coolant. The gas system and device are configured to carry out steps including: supplying gas to be injected into the coolant from the gas system to the space above the coolant space; injecting gas into the coolant by maintaining the gas pressure higher than coolant pressure in the device; and injecting gas into the gas system from the space above the coolant.

Abstract: A mobile calorimeter includes a container comprising one or more walls defining a cavity. The container is adapted to hold a concrete mixture within the cavity. The mobile calorimeter also includes one or more heat flow sensors adapted to detect a heat flow generated by the concrete mixture. The heat flow sensors may include a thermoelectric device, a Peltier plate, or a macro fiber composite (MFC) sensor. The one or more heat flow sensors may be attached to the one or more walls, or may be embedded within the one or more walls. Data relating to a heat flow is obtained by the heat flow sensors, and is used to generate a prediction of a characteristic or performance of the concrete mixture.

Abstract: It is provided with a reactor water-level measurement system being possible to measure water-level in a reactor core regardless of a state in the reactor core. The reactor water-level measurement system includes a core bottom water-level measuring device that includes a heating element, a heat insulating element installed by surrounding part of the heating element in a height direction of the heating element, and a temperature difference measuring element that measures a temperature difference between an insulated portion of the heating element surrounded by the heat insulating element and a non-insulated portion not surrounded by the heat insulating element, a water-level evaluation device that evaluates a water level of a reactor based on the temperature difference, wherein the core bottom water-level measuring device measures at least a water level from a lower end of a reactor core contained in a reactor pressure vessel to a bottom of the reactor pressure vessel.

Abstract: A method for monitoring changes in the boron concentration in the coolant of a reactor during a nuclear plant outage that applies temperature compensation to the source range detector output. The method then monitors the compensated output signal to identify changes in the detector count rate above a preselected value.

Abstract: A real time monitoring system of a spent fuel pool includes a detection unit configured to detect condition information using a sensor installed in the spent fuel pool; an input storage unit configured to receive and store configuration information of a spent fuel, history information related to burnup, and a normal value and a limit value of current condition information; an operation and determination unit configured to calculate the current condition information of the spent fuel by using the condition information detected by the detection unit and the configuration information and the history information stored in the input storage unit and configured to determine a risk level by comparing the current condition information with the limit value corresponding to the risk level; and a display unit configured to display the current condition information calculated by the operation and determination unit and configured to display the determined risk level.

Abstract: A system is provided which employs in-core thermocouples for determining the condition of a water cooled nuclear reactor core, especially monitoring the progress of degradation of the nuclear reactor core during various accidents. A water cooled and moderated nuclear reactor core includes tons of zirconium alloy structures. During various accidents these structures become overheated and exothermic chemical reactions between the zirconium alloy structures and the water lead to accelerated destruction of the nuclear reactor core. The very severe accidents at Three Mile Island Unit-2 during April 1979 and the Fukushima units in Japan during March 2011 were unforeseen and instrumentation was not in place to monitor the course of those accidents. Timely data on the initiation and progress of the degradation of a nuclear reactor core is provided with the inventor's apparatus and his methods of using of the apparatus regardless of the path of an accident.

Abstract: A method for determining a leaking nuclear fuel rod that remotely measures the internal pressure in the plenum region of the fuel rod by remotely measuring the thermal response of the rod when a localized heating is imposed on the outside of the rod in the plenum region. The temperature of the cladding at two symmetrically spaced points on either side of the location where the heating is imposed is remotely monitored as a function of time and compared to a standard to determine the internal pressure of the plenum.

Abstract: A nuclear reactor module includes a reactor vessel containing coolant, a reactor core submerged in the coolant, and a heat exchanger configured to remove heat from the coolant. The nuclear reactor module further includes one or more heaters configured to add heat to the coolant during a startup operation and prior to the reactor core going critical.

Abstract: Systems and methods for the forming of user device groups are presented. First, logical relationship information can be recorded, the logical relationship information describing logical relationships between a first user and other users, the logical relationship information associated with a social networking service. Communications between the first user and the other users, the communications being of at least two different types of communications, may be monitored.

Abstract: A method of detecting defects in nuclear fuel within a fuel rod that first heats the fuel rod to a temperature substantially above the ambient temperature. The surface temperature of the fuel rod cladding is then monitored as the fuel rod is allowed to cool. Variations in the temperature measured over the surface is then noted as an indication of defects.

Abstract: A reactor water level measurement system including: a condensation tank connected with a steam region of a pressure vessel; a reference pipe with one end connected with the condensation tank; fluctuation pipes with one end connected with the pressure vessel; pressure difference measurement type water gauges that detect the water head difference of the reference pipe and the fluctuation pipes and that are connected with the other ends of the reference pipe and the fluctuation pipes; non-pressure difference water gauges that detect the water level of the pressure vessel; and a computation device that selects a pressure difference type water gauge or a non-pressure difference type water gauge, in accordance with a condition of the containment vessel or of the pressure vessel and that indicates/records the water level of the pressure vessel based on the detection result of the selected water gauge.

Abstract: A reactor core coolant temperature measuring apparatus for measuring the temperature of the coolant flowing in the reactor core of a nuclear reactor comprises a ?-ray thermometer having a temperature measuring section arranged below the bottom of the reactor core and a ?-ray heat generation detecting section arranged between the bottom and the top of the reactor core, a cable for transmitting the signal output from the temperature measuring section and an output signal processing means for receiving the signal and computationally determining the local temperature of the coolant at the position of placement of the temperature measuring section.

Abstract: A fuel bundle temperature sensing device may include an indicating rod made of a first material; an outer housing having an upper opening, the outer housing being made of a second material and surrounding at least a portion of the indicating rod; and a rod holder attached to an inner surface of the outer housing, the rod holder being made of a third material and being configured to support the indicating rod such that a top surface of the rod extends out of the upper opening of the outer housing, the third material having a lower melting point than the first and second materials.

Abstract: An improved apparatus for determining the position of a drive rod within the interior of a drive rod housing includes a transmission antenna at one location on the housing and a receiving antenna at another location on the housing. An electromagnetic excitation signal sent to the transmission antenna is detected, at least in part, by the receiving antenna, and the received signal is processed with a vector network analyzer routine to model the drive rod housing as a wave guide having a filter response. A group delay is detected and is compared with a calibration data set which provides a current position of the drive rod that corresponds with the group delay.

Abstract: A method and system for measuring a temperature and liquid level of water within a Spent Fuel Pool (SFP) of a Light Water Reactor, without using external electrical power. The method and system may use a string of thermocouples attached to a cable and mounted in the spent fuel pool. The thermocouples supply their own power when heated by water that is exposed to the thermocouples within the spent fuel pool. The thermocouples may measure the temperature in the SFP. By measuring the temperature, the liquid water level of the SFP may be inferred by comparing a difference in temperature measurements of the thermocouples, as thermocouples exposed to ambient air will experience a noticeably different temperature from the thermocouples covered by water. The cable (with the thermocouples on one end) may terminate in a location that is remote from the SFP.

Abstract: A pair of linear arrays of gamma thermometer (GT) sensors arranged in a nuclear reactor core including: a first linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along a length of the first linear array; a second linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along the second linear array and wherein the second linear array of GT sensors is asymmetrical with respect to the first linear array of GT sensors, and the first linear array positioned in the reactor core at a first core location and the second instrument housing positioned at a second core location, wherein a line of symmetry of the core extends through a center of the core and the first core location is the same horizontal distance from the line of symmetry as the second core location.

Abstract: A method of monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment is provided. The method includes the steps of sensing a condition of the reactor pressure vessel with an instrument, transmitting a signal indicative of the condition of the reactor pressure vessel from the instrument to a powered wireless transmitting modem disposed in the radioactive environment, wirelessly transmitting a signal indicative of the condition of the reactor pressure vessel from the transmitting modem to a receiving modem in the line of sight of the transmitting modem, transmitting a signal indicative of the condition of the reactor pressure vessel from the receiving modem to a signal processing unit, and determining the condition of the reactor pressure vessel from the wirelessly transmitted signal.

Abstract: A method for monitoring and acoustic damping a boiling water reactor which includes: a reactor pressure vessel; a steam pipe for transporting steam out from a steam dome of the reactor pressure vessel; a high pressure turbine driven by the steam; a feedwater heater which heats feedwater supplied to the reactor pressure vessel using bleed steam from the high pressure turbine; a bleeding valve which adjusts a flow rate of the bleed steam; and a pressure sensor provided in a main steam line including the steam dome and the steam pipe. The method includes monitoring a fluctuating pressure in the main steam line and controlling an opening degree of the bleeding valve based on a magnitude of the fluctuating pressure.

Abstract: A system for monitoring a state of a reactor core in a nuclear reactor may include an internal monitoring device located inside the reactor core, the internal monitoring device including one or more internal sensor arrays configured to take measurements of conditions of the reactor core at different vertical regions within the reactor core to generate internal measurement data; an external monitoring device located in the reactor structure outside the reactor core, the external monitoring device including one or more external sensor arrays configured to take measurements of conditions of the reactor core at positions outside the reactor core corresponding the plurality of different vertical regions within the reactor core to generate external measurement data, and a transmitter configured to wirelessly transmit the external measurement data; and a receiver station configured to determine a state of the reactor core based on the external and internal measurement data.

Abstract: The present invention discloses a water level determining method for a boiling water reactor. Thereby, for a boiling water reactor under anticipated transient without scram, risk of sudden power increase due to the uncertain water level raised to a main steam tube can be reduced by installing a thermometer on the top of the reactor or the main steam tube. By controlling flow rate of water and keeping the steam temperature higher than the saturated temperature, the core of the reactor can be sure to cool down properly.

Abstract: An assembly comprising a sample and a device for generating a high temperature gradient in said sample, comprises: a chamber inside which said sample is placed; a resistor passing through said sample; first induction means at the periphery of the chamber to create an electromagnetic field; second induction means connected to said resistor and capable of picking up said electromagnetic field so as to create an induced current circulating in said resistor.

Abstract: In implementing an induction heating stress improvement (IHSI) method in a nuclear plant, cooling characteristic with respect to the inner surface of piping is improved by appropriate structure/layout of nozzles, and further, the cooling effect when applying IHSI to a real machine is verified by installing a thermometer, as well as air in the portion subjected to IHSI is removed by heating the piping prior to the execution of IHSI.

Abstract: An optical gamma thermometer includes a metal mass having a temperature proportional to a gamma flux within a core of a nuclear reactor, and an optical fiber cable for measuring the temperature of the heated metal mass. The temperature of the heated mass may be measured by using one or more fiber grating structures and/or by using scattering techniques, such as Raman, Brillouin, and the like. The optical gamma thermometer may be used in conjunction with a conventional reactor heat balance to calibrate the local power range monitors over their useful in-service life. The optical gamma thermometer occupies much less space within the in-core instrument tube and costs much less than the conventional gamma thermometer.

Abstract: A reactor water-level/temperature measurement apparatus is disclosed that includes an in-core instrumentation tube inserted into an in-core instrumentation housing and an in-core instrumentation guide tube, a plurality of water-level/temperature detection sensors provided in the tube, a temperature measurement device measuring the temperature of a thermocouple included in each of the sensors, a heater control device for controlling a current flowing to a heater wire included in each of the sensors, a storage device used for storing a threshold-value table, and a water atmosphere and a sensor failure and a water-level/temperature/failure determination device.

Abstract: A temperature sensor array includes several temperature sensors at different positions for installation within an instrumentation tube of a nuclear reactor. The temperature sensors measure temperature at multiple axial positions of the nuclear reactor, and plant operators are able to access and interpret this measurement data. Temperatures associated with vessel coolant boiling or loss and/or fuel damage can be detected by the temperature sensors to permit more direct determinations of core fluid levels. Multiple temperature sensor arrays permit vessel fluid levels and conditions to be measured at multiple core locations.

Abstract: A critical heat flux prediction device, a critical heat flux prediction method, a safety evaluation system, and a core monitoring system using the safety evaluation system can predict critical heat flux in a core of a reactor with a high degree of accuracy by obtaining a correlation plot distribution representing a relation of critical heat flux on a thermal equilibrium quality based on experimental data, approximating a correlation plot distribution through a logistic function that is a model function in which critical heat flux is expressed by a function of a thermal equilibrium quality, and obtaining a critical heat flux correlation of critical heat flux and a thermal equilibrium quality.

Abstract: A method of detecting defects in nuclear fuel within a fuel rod that first heats the fuel rod to a temperature substantially above the ambient temperature. The surface temperature of the fuel rod cladding is then monitored as the fuel rod is allowed to cool. Variations in the temperature measured over the surface is then noted as an indication of defects.

Abstract: An apparatus for detecting the location of at least one annulus spacer between concentric interior and exterior tubes when a temperature gradient is present therebetween. A probe head assembly is movable within the interior tube. At least one temperature sensor is coupled to the probe head assembly and configured to detect a temperature of an interior surface of the interior tube. A drive assembly is operable to move the probe head assembly relative to the interior tube. A data acquisition system is coupled to the at least one temperature sensor and configured to receive a plurality of temperature measurements in order to identify at least one position along the interior surface having a temperature abnormality corresponding to a reduced temperature gradient.

Abstract: A liquid level indication system that employs a plurality of heated thermocouples staggered at discrete elevations along a height of a liquid pool, whose outputs are respectively compared to the output of an unheated thermocouple positioned at one of the lower discrete elevations. A significant difference in the outputs of the heated and unheated thermocouples provides an indication of the liquid level.

Abstract: An improved apparatus for determining the position of a drive rod within the interior of a drive rod housing includes a transmission antenna at one location on the housing and a receiving antenna at another location on the housing. An electromagnetic excitation signal sent to the transmission antenna is detected, at least in part, by the receiving antenna, and the received signal is processed with a vector network analyzer routine to model the drive rod housing as a wave guide having a filter response. A group delay is detected and is compared with a calibration data set which provides a current position of the drive rod that corresponds with the group delay.

Abstract: A pair of linear arrays of gamma thermometer (GT) sensors arranged in a nuclear reactor core including: a first linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along a length of the first linear array; a second linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along the second linear array and wherein the second linear array of GT sensors is asymmetrical with respect to the first linear array of GT sensors, and the first linear array positioned in the reactor core at a first core location and the second instrument housing positioned at a second core location, wherein a line of symmetry of the core extends through a center of the core and the first core location is the same horizontal distance from the line of symmetry as the second core location.

Abstract: A system and method of passively monitoring nuclear fuel coolant conditions is disclosed, and includes a passive (non-electrically powered) gamma thermometer(s) located in nuclear fuel coolant area(s) (e.g., in the reactor vessel and/or in the facility spent fuel pool) and in electrical communication with millivolt meter(s) to facilitate passive monitoring of nuclear fuel coolant levels, nuclear fuel coolant temperatures, and/or heat transfer conditions in the nuclear fuel coolant area(s) during a station blackout (SBO) or prolonged station blackout (PSBO) event.

Abstract: The present invention is related to a system and method of inspecting a weld. The weld to be inspected is subjected to a transient sonic excitation while the weld area is monitored using a remote infrared camera. The sonic excitation induces mechanical vibration, which causes heat generation at any cracks or other anomalies in the weld. The infrared camera detects any temperature differentials in the weld, indicating the presence of an anomaly. The system and method is particularly useful for welds in environments that are unsafe for extended human presence or that are crowded or otherwise inaccessible, such as with piping or other structural components.

Abstract: According to an embodiment, core thermal limit monitoring device is provided with calculating units, and signal input processing units, synchronization processing units and signal output processing units, corresponding to the calculating units. The calculating unit determines if it is necessary to output a signal to the control unit by calculating the thermal state values of the monitoring regions based on a signal representing the state of the core. The synchronization processing unit, if it is necessary to output a signal to the control unit, transmits a signal-output stop signal to the other synchronization processing units, and otherwise, the synchronization processing unit transmits a signal-output stop cancellation signal to the other synchronization processing units. The signal output processing unit, when the synchronization processing unit does not receive a signal stop signal, outputs a signal representing the calculation results of the calculating unit to the control unit.

Abstract: The invention provides at least two electrochemical sensors positioned in a nuclear reactor or in a system adjacent to the nuclear reactor, wherein at least one of the at least two electrochemical sensors has a heated zirconium electrode, and the at least two electrochemical sensors produce voltages proportional to an electrochemical corrosion potential for a surface that each of the at least two electrochemical sensors are installed upon. The invention also provides an arrangement configured to accept the voltages produced by the at least two electrochemical sensors, wherein the arrangement is configured to determine an electrochemical corrosion potential of a zirconium fuel rod in the nuclear reactor based upon the voltages of the at least two electrochemical sensors.

Abstract: Disclosed herein are apparatus and methods for detecting and quantifying contaminants in aqueous solutions. In one embodiment, an apparatus for the detection of contaminants within a liquid comprises a sensor and a controller. The sensor comprises a film and a transducer configured such that, during use, a first surface of the transducer is in fluid communication with a liquid, and the film is disposed between the first surface and the liquid. The film can be a polymeric material capable of hindering the transport of a non-desired species therethrough. The controller is in operational communication to the transducer, and is configured to determine the concentration of contaminants within the liquid.

Abstract: A moderator temperature coefficient measuring device include a control rod driving device; a reactivity measurement processing unit capable of measuring the reactivity of the reactor; an inflow temperature sensor for sensing the inflow temperature of a reactor coolant flowing into the reactor; an outflow temperature sensor for sensing the outflow temperature of the reactor coolant flowing out of the reactor; an average temperature measurement processing unit for measuring the average temperature from the sensed inflow and outflow temperatures; an power control unit for providing control so that the reactor delivers constant thermal power; and a temperature coefficient calculating unit for calculating the temperature coefficient of the reactor coolant based on the amount of change in reactivity between before and after the insertion of the control rod and the amount of change in average temperature between before and after the insertion of the control rod under the constant thermal power.

Abstract: A method for monitoring and acoustic damping a boiling water reactor which includes: a reactor pressure vessel; a steam pipe for transporting steam out from a steam dome of the reactor pressure vessel; a high pressure turbine driven by the steam; a feedwater heater which heats feedwater supplied to the reactor pressure vessel using bleed steam from the high pressure turbine; a bleeding valve which adjusts a flow rate of the bleed steam; and a pressure sensor provided in a main steam line including the steam dome and the steam pipe. The method includes monitoring a fluctuating pressure in the main steam line and controlling an opening degree of the bleeding valve based on a magnitude of the fluctuating pressure.

Abstract: A method of monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment is provided. The method includes the steps of sensing a condition of the reactor pressure vessel with an instrument, transmitting a signal indicative of the condition of the reactor pressure vessel from the instrument to a powered wireless transmitting modem disposed in the radioactive environment, wirelessly transmitting a signal indicative of the condition of the reactor pressure vessel from the transmitting modem to a receiving modem in the line of sight of the transmitting modem, transmitting a signal indicative of the condition of the reactor pressure vessel from the receiving modem to a signal processing unit, and determining the condition of the reactor pressure vessel from the wirelessly transmitted signal.