Abstract: Computer-implemented methods for identifying or assessing any type of risk and/or opportunity that may arise can include either, alone or in combination, band pass filtering, principal component analysis, random matrix theory analysis, synchronization analysis, and early-warning detection. Each technique can also be viewed as a process that takes a set of inputs and converts it to a set of outputs. These outputs can be used as inputs for a subsequent process or the outputs may be directly actionable for formulating certain economic predictions to make certain decisions.

Abstract: A submersible system and method for measuring the gas volume fraction in an aerated fluid inside a reactor (1) wherein the aerated fluid comprises a gas dispersed in the form of bubbles in a fluid in the form of a solution, suspension, mixture of liquids or a combination thereof. The submersible system comprises: an open and pass-through gas exclusion device (20) of a variable cross section wherein the inlet opening whereby the fluid enters without gas bubbles towards the opened and through gas bubbles exclusion device (20) is greater than the outlet opening whereby the fluid exits without gas bubbles of the opened and through gas bubble exclusion device (20). The outlet opening abuts with an inlet pipe (23).

Abstract: A method and apparatus for extending the period a nuclear steam supply system spent fuel pool can be safely passively cooled by storing the spent fuel offloaded from the reactor, in the containment for one reactor operating cycle. During a refueling the spent fuel that is not to be returned to the reactor and the spent fuel that will be returned to the reactor are stored separately in shielded locations within the containment. After one operating cycle, the spent fuel stored within the containment that was not returned to the reactor just prior to the last operating cycle, is offloaded to the spent fuel pool and replaced by the newly offloaded spent fuel that is being retired.

Abstract: A mobile processing system is disclosed for the removal of radioactive contaminants from nuclear process waste water. The system is fully scalable, modular, and portable allowing the system to be fully customizable according the site-specific remediation requirements. It is designed to be both transported and operated from standard sized intermodal containers or custom designed enclosures for increased mobility between sites and on-site, further increasing the speed and ease with which the system may be deployed. Additionally, the system is completely modular wherein the various different modules perform different forms or stages of waste water remediation and may be connected in parallel and/or in series. Depending on the needs of the particular site, one or more different processes may be used. In some embodiments, one or more of the same modules may be used in the same operation.

Abstract: A light water reactor to safely convert depleted uranium into a fuel source that could be used as a sustainable source of energy for centuries. The reactor is a type of breed-burn reactor uniquely combined with a proliferation-resistant fuel cycle with no uranium enrichment and no plutonium isolation. It is comprised of a compact factory-produced fast region and a thermal region that produces about 95% of the core power and contains the passageways for transports of delayed-neutron emitters to the fast region, where they can provide additional neutrons (source-based mode) or all the necessary excitation without an external neutron source (self-regulating mode). A second embodiment of the invention is a small unit driven by a neutron source with beam recycling for propulsion, electrical power or radioisotope production. It could also serve as a demonstration facility for the transmutation reactor with fission-fusion fuel.

Abstract: This invention relates to the field of coolant flow in a fluoride salt cooled high temperature reactor. In particular, the invention relates to the discovery of use of nitrogen-16 and/or fluoride-20 decay signature to measure coolant flow. The method of the invention comprises detecting gamma radiation emanating from nitrogen-16 activity and/or fluorine-20 activity within the reactor coolant at a first position along the reactor coolant loop; detecting the gamma radiation emanating from the nitrogen-16 activity and/or fluorine-20 activity within the reactor coolant at a second position along the reactor coolant loop downstream of said first position. The gamma radiation detected from the first position and second position are then cross-correlated, thereby determining the transit time of corresponding gamma activity perturbations viewed at the two detector locations.

Abstract: A system and method for predicting acoustic loads expected on a boiling water reactor (BWR) may include a BWR scale model, a test fixture for generating air flow in the scale model, and one or measurement devices for monitoring system behavior to predict how acoustic loads may affect plant operation for the BWR being evaluated.

Abstract: According to an embodiment, a power monitoring system for a nuclear reactor comprises at least a first system and second system. The first system and the second system respectively comprise a plurality of APRM units, a plurality of FLOW units, and a plurality of OPRM units. The APRM units respectively generate an LPRM signal that indicates the local output of neutrons by the reactor core, and generate an APRM signal indicating the average output of the reactor core, based on the LPRM signal. The FLOW units respectively generate a FLOW signal indicating the flow rate of reactor coolant. The OPRM units respectively are supplied with the LPRM signal and the APRM signal from at least two aforementioned APRM units and are supplied with the FLOW signal from at least one aforementioned FLOW unit; and, based on the supplied LPRM signals, APRM signals and FLOW signals, generate a trip signal for shutting down the reactor.

Abstract: A method and apparatus for measuring a liquid level of a Spent Fuel Pool of a Light Water Reactor without using electrical power. The method and apparatus may use a pressurized gas source connected to tubing or piping that may discharge near the bottom of the Spent Fuel Pool. The system may include a flow meter and throttle valve that may be used to determine a required gas pressure to provide a specified flow rate of gas for known Spent Fuel Pool water levels. By obtaining calibration data points of pressure and flow for multiple Spent Fuel Pool water levels, a calibration curve may be obtained that allows for the measurement of Spent Fuel Pool liquid level using the system, without the need for electrical power.

Abstract: Provided is an apparatus for testing a loss-of-coolant accident using a model of a nuclear containment building, including a containment vessel of which an upper surface is opened and side and lower surfaces are transparent; an internal structure which is disposed in the containment vessel; a hose pipe which is disposed at an upper side of the containment vessel; and a measuring device which is disposed at the lower surface of the containment vessel so as to monitor movement of fluid and debris in the containment vessel.

Abstract: A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

Abstract: A method of predicting stresses on a BWR steam dryer that includes creating an analytical acoustic model of a BWR steam system; generating pressure estimations by inputting empirical data into the analytical acoustic model of the BWR steam system; creating an analytical structural model of the BWR steam dryer; and predicting stresses on the BWR steam dryer using the analytical structural model and the pressure estimations.

Abstract: A method and system for entropy coding can comprise, in response to detecting a first symbol combination comprising first run information indicating a first number of contiguous zero coefficients is greater than a cut-off-run value, assigning a first codeword to a first symbol combination, wherein the first codeword comprises an escape code from a first-level VLC table; and in response to a second symbol combination comprising second run information indicating a second number of contiguous zero coefficients is less than or equal to the cut-off-run value, assigning a second codeword to the second symbol combination, wherein the second codeword is from the first-level VLC table. The system and method can further comprise collecting coding statistics for a set of candidate symbol combinations and adjusting a mapping between codewords of the first-level VLC table and a subset of the set of candidate symbol combinations based on the coding statistics.

Abstract: According to an embodiment, a digital process type monitor device includes a plurality of modules and a mother board connected to each of the modules. Each module includes: a base board connected to a connector and having an FPGA for main control and an IPGA for sub board control mounted thereon; and a sub board for a main-machine I/F process, having an FPGA for an I/F process mounted hereon. Each sub board has storage devices for storing man-machine I/F information on the sub board. Each of the FPGA writes transmission data into a predetermined region of a transmission area and has a common transmission protocol to share the transmission data between the respective modules.

Abstract: The present invention provides a nuclear activation apparatus for one or more fluid samples comprising the following modules; a means for introducing one or more fluid samples to a sample conduit, an activation thimble, comprising a section of sample conduit configured for multiple passes adjacent a radiation source, an absorber located adjacent to the activation thimble, and a detector located adjacent the sample conduit, wherein the relative arrangement of the modules can be altered specific to an application and the rate of flow of the fluid sample adjacent the radiation source can be controlled.

Abstract: A technology for accurately measuring with the use of an ultrasonic flowmeter the flow rate through a nuclear reactor water supply piping interposed between a nuclear reactor and a nuclear reactor water supply pump. In a reactor coolant cleaning system (30), piping to be joined with a reactor water supply system (10) is branched, thereby forming an auxiliary path (A) to be similarly joined with the reactor water supply system (10). The auxiliary path (A) is provided with a restriction part (C) having an inside diameter narrower than that of other parts of the auxiliary path (A) to complete a nuclear reactor water supply piping structure. Through this structure, cavitation bubbles are generated from the restriction part (C). An ultrasonic transducer (40) is disposed between the nuclear reactor (1) and the auxiliary path (A) in the piping for composing the reactor water supply system (10), thereby conducting measurement of the flow rate of fluid within the piping.

Abstract: A method of determining the R-factor for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind. The R-factor is a factor that accounts for the weighted local power influence on a fuel rod. A local R-factor (Ri(z)) is determined for each fuel rod (i) in said bundle and for each of a plurality of levels (z) in an axial direction. The individual axial heat generation profile for a certain fuel rod (i) is taken into account when determining the local R-factors (Ri(z)) for said fuel rod (i). A processor is configured for automatically determining the R-factor. A computer program product, a method of determining the critical power for a bundle of fuel rods, a nuclear energy plant, and a method of operating a nuclear energy plant are also described.

Abstract: A method is provided for determining and controlling feed water level in a steam generator of a nuclear power plant, whereby the water level sensor is calibrated in terms of the pressure drop across structural components in the feed water path. The water level differential pressure sensor is calibrated so that maximum water level is indicated as the level corresponding to the upper tap level plus the pressure drop across the foregoing structural components at maximum power plant power.

Abstract: Object of this invention is to provide an axial torsional vibration measurement device of the reactor internal pump capable of measuring the axial torsional vibration easily. The vibration detector 2 that detects vibration in the radial direction, and the sound detector 3 that detects vibration in the circumferential direction are fitted on cylindrical motor casing 6, which contains the motor of a reactor internal pump. The dimension of both signals of a vibration waveform signal detected by the vibration detector 2, and a sound waveform signal detected by the sound detector 3 are matched in vibration calculating device 15, to thereby obtain an axial torsion vibration of a motor from the difference of both signals.

Abstract: Object of this invention is to provide an axial torsional vibration measurement device of the reactor internal pump capable of measuring the axial torsional vibration easily. The vibration detector 2 that detects vibration in the radial direction, and the sound detector 3 that detects vibration in the circumferential direction are fitted on cylindrical motor casing 6, which contains the motor of a reactor internal pump. The dimension of both signals of a vibration waveform signal detected by the vibration detector 2, and a sound waveform signal detected by the sound detector 3 are matched in vibration calculating device 15, to thereby obtain an axial torsion vibration of a motor from the difference of both signals.

Abstract: A computerized method for expanding the operating domain of a boiling water nuclear reactor that permits safe operation of the reactor at core flows lower than normal operating parameters is provided. The operating domain is characterized by a map of the reactor thermal power and core flow. The computerized method includes determining by computer simulation a load line characteristic that is elevated over normal operating parameters and that increases reactor performance over normal operating parameters, performing safety evaluations by computer simulation at the elevated load line to determine compliance with safety design parameters, and performing operational evaluations by computer simulation at the elevated load line.

Abstract: A reactor pressure vessel (RPV) for a nuclear reactor that permits measurement of the flow through each reactor internal pump (RIP) is described. The reactor pressure vessel also includes at least one reactor internal pump that includes an impeller and a pump diffuser. At least two seal rings extend circumferentially around an outer surface of the diffuser housing outer wall and are located in circumferential grooves in the housing outer wall. At least one lateral bore extends through the housing outer wall into a diffuser housing longitudinal flow passage. Each lateral bore is located in an area between two adjacent seal rings, with each inter-seal ring area containing one lateral bore. At least one pressure tap bore extends from the outer surface of the RPV bottom head petal, through the pump deck to an inner surface of a pump deck opening. Each pressure tap bore is aligned with an area in the RIP containing a corresponding lateral bore.

Abstract: A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. A data processing system is used to execute specific program routines that simultaneously simulate the thermal operating characteristics of fuel rods within the reactor during a transient operational condition. The method employs a multi-dimensional approach for the simulation of postulated operational events or an anticipated operational occurrence (AOO) which produces a transient condition in the reactor—such as might be caused by single operator error or equipment malfunction. Based on a generic transient bias and uncertainty in the change in critical power ratio (&Dgr;CPR/ICPR), histograms of fuel rod critical power ratio (CPR) are generated.

Abstract: A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. A data processing system is used to execute specific program routines that simultaneously simulate the thermal operating characteristics of fuel rods within the reactor during a transient operational condition. The method employs a multi-dimensional approach for the simulation of postulated operational events or an anticipated operational occurrence (AOO) which produces a transient condition in the reactor—such as might be caused by single operator error or equipment malfunction. Based on a generic transient bias and uncertainty in the change in critical power ratio (&Dgr;CPR/ICPR), histograms of fuel rod critical power ratio (CPR) are generated.

Abstract: A valve control apparatus for generating an anticipatory output response for controlling a fluid flow in a fluid flow line includes a valve provided inside the fluid flow line, at least one sensor for providing at least one input signal, said at least one input signal responsive to properties of said fluid flow in advance of said valve, and a letdown back-pressure controller receiving said at least one input signal and providing a tunable output signal for selectively opening and closing said valve to vary flow in the flow line responsive to said at least one input signal.

Abstract: A modular differential pressure measuring system for a boiling water nuclear reactor pressure vessel is described. The modular pressure differential system includes a plurality of pressure lines having a plurality of pressure line sections. The system also includes a shroud having at least one replaceable shroud section. Each shroud section includes at least one pressure line section which is configured to connect to and disconnect from corresponding pressure line sections in adjacent shroud sections without welding. Additionally, the system includes a reactor bottom head petal section having a shroud support flange and a plurality of bores defining pressure line sections wherein at least one pressure line section of said bottom head petal is configured to couple with a corresponding pressure line section of an adjacent shroud section. The modular pressure system does not require cutting of the pressure lines or pressure line supports for replacement of the replaceable shroud sections.

Abstract: A safety grade protection system for monitoring parameters indicative of the status of a core of a nuclear reactor, and reducing the criticality of the nuclear reaction when the onset of unsafe conditions are detected. The invention directly detects, in real time, the power at a plurality of axial and radial locations within or about the core and provides a three-dimensional real-time core power output. The system processes those outputs together with a measure of core coolant inlet temperature, pressure and flow rate to calculate, in real time, the fuel centerline temperature, the heat transfer condition at the fuel elements' boundary and the rate of power change within the core.

Abstract: A power range monitor apparatus includes a plurality of average power range monitors for calculating/monitoring local power values, average reactor power values, and core flow rate values, a plurality of rod block monitors for monitoring the average values of local power values, and a plurality of flow rate signal converters for converting the signal form of recirculation rate detection signals. The flow rate signal converters concurrently transmit recirculation rate detection signals as digital signals to the average power range monitors. The average power range monitors obtain core flow rate signals from the respective recirculation rate detection signals, use them to monitor the average reactor power, and transmit them to the rod block monitors. The rod block monitors select optimal values from average reactor power values and core flow rate signals to perform a monitoring operation.

Abstract: Fluid flow rates are determined with a nonradioactive tracer agent in the operation of a water cooled nuclear fission reactor employed for driving electrical power generators. The technique is proposed for use to routinely calibrate other flow measuring means such as restricted flow nozzles.

Abstract: A method for determining the core flow rate from pump differential pressure method equations and a heat balance equation for determining the temperature and density of the reactor water in the downcomer and lower plenum region. The measured reactor water temperature is used only as an initial estimate for this new method. This allows the first estimate of the core flow to be determined based upon the reactor water density that corresponds to this temperature and the pump differential pressure and pump rotational speed measurements. Then, with this intial core flow measurement value, and other available plant measurements and assumed nominal values for the heat balance equation, revised values for the reactor water temperature and density can be calculated. Using a standard numerical method, such as the successive approximation technique, a more accurate estimate for the reactor water density can be obtained.

Abstract: In a steam generator or boiler of the type having a pressure vessel having a zone in which heated water and steam can be separated, an outlet for the flow of pressurized steam and an outlet for the flow of liquid, a riser section in which fluid passes for heating therein and flow into the vessel zone, and a downcomer to receive the recirculated liquid from the vessel zone and feedwater for flow to the inlet of the riser section, the system includes feedwater control apparatus for sensing the mass flow of liquid in the downcomer, determining the liquid mass in the vessel zone, downcomer and riser section and controlling the feedwater rate in relation to such mass and the respective power conditions of the system, thereby providing better stability in the system operation. Trips and other problems caused by shrink and swell are thereby avoided and other benefits achieved.

Abstract: Failed fuel cladding in light water reactors can lead to premature reactor shutdown for removal of the failed fuel bundles from the core. The lost power and associated fuel replacement costs are astronomically high, approaching a million dollars per day in near term costs. Additionally, the long term affect on an operating plant are the release and distribution around the plant of long-lived fission products which increase radiation levels and impose increased radiological protection procedures. A method has been developed which allows for on-line early detection of failed fuel elements while the plant is operating through flowing off-gas through a conduit to a gamma spectrograph with a time delay to allow interfering short lived isotopes to decay and then analyzing for gamma radiation from at least one of the nuclides Kr-85m, Kr-87, Kr-88, Xe-133, Xe-135, Xe-135m and Xe-138.

Abstract: The method and the device are applicable to the penetrations (2) of a pressurised water nuclear reactor vessel head (1) during operation.The device includes a double circuit for picking off two gas samples (100, 16, 17; 12, 13, 14, 15) and a measurement assembly (18) which regularly compares the content, especially of water vapour, in the two samples picked off. The first sample is taken from the volume of a chamber (7) where the possible leak (10) emerges, and the second in proximity to the head (1) outside the region immediately affected by the leak. The assembly (18) makes it possible to generate (19) an alarm in the event of a leak and to distinguish it from operational alarms.Techniques and means are also provided for determining the flow rate of sweeping in the volume of the chamber (7) and for checking the pick-off and detection systems.

Abstract: An instrumentation probe for measuring and recording one or more physical parameters within a fluid duct, such as a nuclear reactor fuel channel, comprises one or more sensors responsive to the parameters to be measured and recorded, each sensor actuating a scribe which cooperates with a recording chart. The recording chart is a drum, disc, tape or like recording device, and is volume driven at a constant speed by an impeller responsive to the fluid flow. No connections to external instrumentation are required.

Abstract: Excore detector measurements are used to generate on-line absolute reactor power in a pressurized water reactor (PWR) by calibrating detector current measurements to the reactor thermal power calculation made at a base time early in reactor cycle while the thermal reactor power measurement is still accurate. Measurements are also made at the base time of the three-dimensional core power distribution and the core inlet temperature. Present core power measurements are then made by measuring the present excore detector current, the most recent three-dimensional core power distribution and the present core inlet temperature. The present core power is then calculated as the ratio of the present detector current to the detector current at the base time multiplied by the reactor thermal power measurement at the base time.

Abstract: A gas injection system of a nuclear power plant including a reactor and a gaseous waste disposal system, comprises a hydrogen injection unit for injecting hydrogen into a reactor of a nuclear power plant, an oxygen injection unit for injecting oxygen into a gaseous waste disposal system in which the oxygen is recombined with the hydrogen passed by way of the reactor, a densitometer for measuring a balance of concentration between the hydrogen and the oxygen in the gaseous waste disposal system, a flowmeter for measuring a flow rate of an exhaust gas of the gaseous waste disposal system, and a controlling unit for controlling the quantities of the hydrogen gas and the oxygen gas.

Abstract: A fluid detector (20) has an enclosure (32) defining an internal chamber (33). The enclosure has a frangible wall portion (34). A sensor (46) is arranged within the chamber. The sensor has a hygroscopic portion (48), the electrical resistance of which changes depending upon whether the sensor portion is wet or dry. The chamber is normally filled with a dry fluid, such as nitrogen. The outer surface of the enclosure is exposed to another fluid containing water. When the pressure differential across the frangible wall portion exceeds a predetermined minimum value, the frangible wall portion ruptures to permit the sensor portion to become wetted. An ammeter (31) is arranged to determine the change in the electrical resistance of the sensor and to indicate the presence of the water-containing fluid within the chamber. The detector is particularly useful in determining the fluid-tight sealed integrity of a closed-end thimble having a portion inserted into a pressurized water reactor.

Abstract: A device for easily installing and removing strain gauge instrumentation and associated wiring on the inside of in-core monitor housings. A vibration sensor is attached to an inside diameter of a cylindrical housing. A signal cable is coupled to the vibration sensor and extends into an exterior of the reactor pressure vessel. A protective assembly minimizes the damage inflicted on the signal cables during handling. The signal cables are helically wound in an annular space between a pair of concentric thin-walled tubes and then passed through cable penetration bores formed in the body of a flange insert. The flange insert supports the concentric tube assembly and is sandwiched between the housing flange and the in-core monitor instrument flange of a conventional in-core monitor assembly.

Abstract: Coolant flow (900) in the core (103) of a natural circulation boiling water reactor (100) is monitored by making use of a detector (208) for nuclear radiation and a turbine device (204) which comprises a rotor (304) which can modulate the flux (206) of the nuclear radiation on the detector. The turbine rotor can be installed in a fuel bundle (104), while the detector can be placed in an adjoining instrumentation tube (320). Included in the turbine rotor is material (308) which modulates the radiation field (neutron field and/or gamma field). The radiation detector detects (602) variations in the radiation field when the turbine rotor device is set into rotation by the coolant flow (900). The coolant flow rate is then calculated (603) from the speed of these variations.

Abstract: A flow measurement probe measures reactor coolant water flow in a cold leg pipe of a nuclear reactor system wherein a lack of elbows precludes the use of elbow tap flow measurements. An elongated probe body disposed in the pipe includes a common dynamic pressure tap in the peripheral wall of the probe body facing about into the direction of coolant water flow. A plurality of static pressure taps are oriented about normal to the direction of coolant water flow. Each of the static pressure taps is operatively connected witch one input of a different differential pressure transmitter. The other input of each of the transmitters is operatively connected with the common dynamic pressure tap. The differential pressure measured by each transmitter is proportional to the square of the magnitude of reactor coolant water flow. The multiple taps in the probe allow multiple redundant flow measurements to be made with a single device, thus avoiding the need for multiple penetrations in the pipe.

Abstract: The device comprises a mobile probe which can be displaced inside thimbles. A plurality of ducts (20) for the passage and support of a thimble are disposed in the partitioning (6) of the reactor, around the core (4) along an axial direction and substantially over the entire height of the core. The ducts (20) are distributed circumferentially around the core (4). A plurality of feed-through sleeves in the cover of the vessel are each disposed along the axial extension of a duct (20) for the passage of a thimble. Guide columns each disposed along the axial extension of a duct and each engaged in a feed-through sleeve in the cover of the vessel form, with the corresponding duct (20) and feed-through sleeve, a channel for the reception of a thimble. An arrangement for the guidance, displacement, switching and parking of the mobile probe is disposed outside the vessel (1) and fixed on to the cover of the vessel.

Abstract: A method and apparatus of sensing obstructions in a tubular nuclear fuel rod or the like is disclosed. A nozzle is positioned to engage one end of the fuel rod in an air tight sealed relationship. A first gas flows through the nozzle and then into and through the fuel rod. The pressure is sensed and compared with a standardized pressure to determine any excess back pressure in the nozzle indicating an obstruction present therein.

Abstract: A flow meter and temperature measuring device comprising a tube with a body centered therein for restricting flow and a sleeve at the upper end of the tube to carry several channels formed longitudinally in the sleeve to the appropriate axial location where they penetrate the tube to allow pressure measurements and temperature measurements with thermocouples. The high pressure measurement is made using a channel penetrating the tube away from the body and the low pressure measurement is made at a location at the widest part of the body. An end plug seals the end of the device and holes at its upper end allow fluid to pass from the interior of the tube into a plenum. The channels are made by cutting grooves in the sleeve, the grooves widened at the surface of the sleeve and then a strip of sleeve material is welded to the grooves closing the channels. Preferably the sleeve is packed with powdered graphite before cutting the grooves and welding the strips.

Abstract: To inspect the integrity of pressurization holes in nuclear fuel rod final end plugs, quality assurance apparatus includes an adapter providing a test chamber into which an end plug is inserted. Instrument gas is introduced into the test chamber through an inlet port for flow to the atmosphere through the pressurization hole. A mass flow meter measures the gas flow through the pressurization hole to determine if it is capable of supporting requisite pressurization of a nuclear fuel rod when the end plug is assembled thereto.

Abstract: Using the method of this invention, the performance of a heat exchanger is determined by measuring the heat transfer capabilities of an individual tube. A relatively small reservoir of service fluid is connected to the inlet and outlet ports of a tube. The reservoir is provided with a heater or chiller and the service fluid is circulated through the tube. When a steady state is reached, the heat transfer characteristics of the tube are measured using known mathematical relationships.

Abstract: In an apparatus including a fluid circulation unit for allowing a fluid in a container to be circulated along a passage in the container, a flow rate measuring apparatus includes a rotational speed transducer for detecting the rotational speed of a motor for driving the fluid circulation unit and a pipe network model calculator for receiving a rotational speed signal from the rotational speed transducer. The model calculator has an initially programmed pipe network model corresponding to flows in the container and calculates the flow rate at a corresponding location of the model with the rotational speed of the circulation unit replaced by constants of the model and hence calculates a flow rate of the fluid at that location in a real passage.

Abstract: A system for measuring the power distribution in a reactor core includes a plurality of gamma detectors disposed outside of the reactor vessel which houses the reactor core. The gamma detectors correspond in number to the number of instrumentation tubes in the fuel assemblies in the reactor core to be sampled. A first plurality of separate flow branches connect the instrumentation tubes with the gamma detectors for routing the flow of coolant from the instrumentation tubes to the gamma detectors so that the gamma detectors can measure the .sup.16 N decay gamma activity in the respective instrumentation tubes of the fuel assemblies. A second plurality of separate flow branches connect a cold leg of the reactor coolant system with the gamma detectors for routing the flow of coolant from the cold leg to the gamma detectors so that the gamma detectors can measure the background gamma activity in the respective instrumentation tubes of the fuel assemblies.

Abstract: A hydroball string sensing system for a nuclear reactor that includes stainless tubes positioned to guide hydroball strings into and out of the nuclear reactor core. A sensor such as an ultrasonic transducer transmitter and receiver is positioned outside of the nuclear reactor core and adjacent to the tube. The presence of an object such a bullet member positioned at an end a hydroball string, or any one of the hydroballs interrupts the transmission of ultrasound from the transmitter to the receiver. Alternatively, if the bullet member and hydroballs include a ferritic material, either a Hall effect sensor or other magnetic field sensors such as a magnetic field rate of change sensor can be used to detect the location and position of a hydroball string. Placing two sensors along the tube with a known distance between the sensors enables the velocity of a hydroball string to be determined.

Abstract: A sensor assembly comprising a rod-shaped thermally conductive element oriented radially with respect to a flow pipe and two or more thermocouples installed on the thermally conductive element. A first end of the thermally conductive element is arranged in contact with or in thermal communication with the outer periphery of the pipe, while a second end of the thermally conductive element is spaced from the pipe. As a result of heat conduction through the pipe to the first end, a temperature gradient is provided across the thermally conductive element. The temperature gradient exhibits a relatively high temperature near the first end and a relatively low temperature near the second end. The thermocouples are installed along the length of the thermally conductive element. One thermocouple is arranged nearer to the pipe than the other thermocouple. Temperature dependent signals obtained from the thermocouples are transmitted to suitable evaluation electronics.

Abstract: A method is provided for determining and controlling steam water level in a steam generator of a nuclear plant, comprising calibrating the water level sensor in terms of velocity head and also adjusting the high level setpoint in terms of a velocity head bias. The water level differential pressure sensor is calibrated so that maximum water level is indicated as that level corresponding to the upper tap level less velocity head at maximum power plant power. The high level setpoint is calculated as corresponding to the riser level less a velocity head bias, the bias being calculated as maximum velocity head at maximum velocity minus riser head percentage of span times velocity head at maximum power.