Abstract: In a boiling water reactor controlling each control rod hydraulically by driving a solenoid valve, the control system comprises an operation control means 41 having a duplicated data processing unit for generating sequence patterns based on the timing of the driving sequence based on the control information provided manually, a transmission control means 42 for creating a command word corresponding to each control rod being controlled based on said sequence pattern, and mutually communicating each command word between duplicated data processing units and computing the AND logic of said data within a predetermined time difference, and when the computed result coincide, transmitting the selected command word serially; a transmission unit 32 for receiving said command word and performing protocol conversion thereto, and transmitting the same to a plurality of transmission branch units positioned downstream as the control command; and a solenoid valve drive circuit 31 for driving the control rod drive unit corres

Abstract: According to the method, a reactor core is shut down during the operation cycle. One or more fuel bundles of the reactor core are then moved to new positions within the reactor core to increase a total energy, output of the reactor core as compared to continuing operation of the reactor core without the shutting down and moving steps.

Abstract: In order to provide an operation and maintenance planning aiding system for a power generation installation which prepares an operation plan for a plurality of power generation units by making use of actual plant data and based on a total judgement including a variety of circumstances of the machine and apparatus or the parts thereof in the power generation units, in the system the plurality of power generation units 41, 42, 51 and 52, a power supply command center 3 and a service center 1 are arranged and connected via a communication network 6, the service center 1 obtains the plant data via the communication network 6 from the plurality of power generation units 41, 42, 51 and 52, calculates in real time a power generation efficiency of a concerned power generation unit for every plurality of power generation units 41, 42 51 and 52 by making use of the obtained plant data and design data of the concerned power generation unit and prepares an operation and maintenance plan for each of the power generation u

Abstract: The disclosed method, a technique for analyzing anisotropic particle scattering, includes the step of establishing a three-dimensional geometric representation of an object. A set of rays is then projected through the three-dimensional geometric representation to produce simulated anisotropic particle scattering. An integral particle transport analysis is then executed on the simulated anisotropic particle scattering to produce particle transport data. The disclosed apparatus, a computer readable memory to direct a computer to function in a specified manner, includes a computer aided design system to establish a three-dimensional geometric representation of an object. The computer aided design system includes a line-of-sight analysis feature to project a set of rays through the three-dimensional geometric representation to produce simulated anisotropic particle scattering.

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: In order to provide an operation and maintenance planning aiding system for a power generation installation which prepares an operation plan for a plurality of power generation units by making use of actual plant data and based on a total judgement including a variety of circumstances of the machine and apparatus or the parts thereof in the power generation units, in the system the plurality of power generation units 41, 42, 51 and 52, a power supply command center 3 and a service center 1 are arranged and connected via a communication network 6, the service center 1 obtains the plant data via the communication network 6 from the plurality of power generation units 41, 42, 51 and 52, calculates in real time a power generation efficiency of a concerned power generation unit for every plurality of power generation units 41, 42 51 and 52 by making use of the obtained plant data and design data of the concerned power generation unit and prepares an operation and maintenance plan for each of the power generation u

Abstract: A fuel assembly includes a bottom nozzle set on a lower core plate of a nuclear reactor, a top nozzle with a hold down spring to urge the bottom nozzle against the lower core plate, guide thimbles which guide control rods, having passed through the top nozzle, toward the lower core plate, a dashpot formed on each of the guide thimbles to reduce the fall velocity of a corresponding one of the control rods, a thimble screw which connects each of the guide thimbles to the bottom nozzle, and a drain hole formed to extend through each of the thimble screw. The dashpot has a large-diameter portion with substantially the same diameter as that of each of the guide thimbles. The diameter d of the drain hole falls within a range of 0.04D<d<0.08D where D is an inner diameter of the large-diameter portion.

Abstract: A fuel assembly for a pressurized water reactor having control rod guide thimbles (5) each having a dashpot (12) for protecting against flexural deformation which may impair insatiability of a control rod. The fuel assembly includes, a plurality of control rod guide thimbles (5) having bottom and top end portions fixedly secured to a lower nozzle (2) and an upper nozzle (4), respectively, disposed in opposition to each other. The dashpot (12) of each control rod guide thimble (5) includes a small diameter section (13b) having an outer diameter smaller than that of the control rod guide thimble (5) formed at an upper portion of the dashpot (12), and a large diameter section (13a) having an outer diameter substantially equal to that of the control rod guide thimble (5) formed at a lower portion of the dashpot (12). With the length of the control rod guide thimble (5) represented by L, the effective length (S) of the small diameter section (13b) is selected to lie within a range of from 0.03 L to 0.1 L.

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 fuel assembly for a pressurized water reactor having control rod guide thimbles (5) each having a dashpot (12) for protecting against flexural deformation which may impair insertability of a control rod The fuel assembly includes, a plurality of control rod guide thimbles (5) having bottom and top end portions fixedly secured to a lower nozzle (2) and an upper nozzle (4), respectively, disposed in opposition to each other. The dashpot (12) of each control rod guide thimble (5) includes a small diameter section (13b) having an outer diameter smaller than that of the control rod guide thimble (5) formed at an upper portion of the dashpot (12), and a large diameter section (13a) having an outer diameter substantially equal to that of the control rod guide thimble (5) formed at a lower portion of the dashpot (12). With the length of the control rod guide thimble (5) represented by L, the effective length (S) of the small diameter section (13b) is selected to lie within a range of from 0.03 L to 0.1 L.

Abstract: A method of controlling the system pressure in a power generating system, having a turbine-generator and a BWR, that modulates the core thermal power of the reactor while maintaining the main turbine control valves in a constant steady position is described. The core thermal power may be adjusted by adjusting the control rod density within the reactor core or by adjusting the flow rate through the reactor which may be accomplished by modulating the speed of variable frequency recirculation pumps or by modulating recirculation flow control valves. The method includes transferring the power generation system from normal turbine control valve modulation pressure control to core thermal power modulation pressure control. Additionally the method includes modifying the bypass valve closure bias and the power control bias to accommodate the variances from core power modulation pressure control over normal pressure control.

Abstract: High temperature reactor with residual-heat transfer system comprises a cooling gas intake at the bottom and cooling gas outlet at the top so that a cooling gas can flow from the bottom to the top through the reactor core. In order to assure reliable heat transfer a bypass duct is provided with a lower end communicating with the cooling gas intake and the upper end communicating with the cooling gas outlet. The bypass duct is arranged parallel to the reactor core and passing a partial flow of cooling gas from the bottom to the top. This partial flow of cooling gas heats up only trivially. This partial flow of cooling gas is further cooled by the cooler. The upward flow of the comparatively cold cooling gas in the bypass duct stops and by itself reverses because the cooling gas in the bypass duct is drawn toward the reactor core on account of the natural convection.

Abstract: A core of a light water boiling reactor comprises a plurality of vertical fuel assemblies (10) and a plurality of control rods, each control rod comprising four vertical blades arranged in a cruciform. The control rods are arranged with each one of their blades between two fuel assemblies located in the same row, such that each control rod together with four fuel assemblies arranged around the blades of the control rod form a unit, the control rod unit (30, 30-o), having an at least substantially square cross-section. The control rod units are arranged in a symmetrical lattice with each control rod unit included in two rows of control rod units perpendicular to each other.

Abstract: A gas cooled nuclear reactor is permanently deactivated on occurrence of an emergency by pyrolytic deposition of boron or refractory or stable compounds of boron in the fluid channels of the fuel elements of the reactor. The boron is enriched in boron 10. The deposition takes place in so short a time interval as to preclude a major catastrophe by reason of penetration of water into the reactor. Carbon and/or nitrogen-containing boron compounds are injected directly into the reactor or compounds generated by reaction in the working fluid of the reactor between diborane and other boron hydrides with unsaturated compounds, such as acetylene and ammonia flow through the reactor. The compounds are carried by the working fluid through the core and are pyrolized in the heat of the core to produce more stable boron, or boron-carbon and boron-nitrogen compounds or metal borides which adhere to the walls of the channels in the fuel elements, deactivating the core.

Abstract: An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18, 20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24) for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18, 20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.