Abstract: Disclosed embodiments include electromagnetic flow regulators for regulating flow of an electrically conductive fluid, systems for regulating flow of an electrically conductive fluid, methods of regulating flow of an electrically conductive fluid, nuclear fission reactors, systems for regulating flow of an electrically conductive reactor coolant, and methods of regulating flow of an electrically conductive reactor coolant in a nuclear fission reactor.

Abstract: The present invention provides a method of operating a Boiling Water Reactor, having the steps of analyzing LPRM signals for oscilliatory behavior indicative of neutron-flux-coupled density wave oscillations, determining if oscilliatory behavior is present in the signals; initiating a reactor protective corrective action if the oscilliatory behavior is determined, and in addition, initiating corrective actions if neutron uncoupled oscillations are possible. Detecting the later is performed through analytically determined exclusion zone on the power flow map or by on-line stability calculations for several high power channels.

Abstract: A recirculation pump motor is supplied with a power from a unit auxiliary middle voltage bus through a power supply system including a first circuit breaker, a voltage source inverter, and a second circuit breaker electrically connected in series to provide a no-load operation by use of the voltage source inverter. The second circuit breaker may be multiplexed with more than one breaker electrically connected in series. An existing nuclear plant using a induction motor driving a hydraulic coupling mechanically coupled to a synchronous generator for driving the recirculation pump, such as an MfG set, may be subjected to a method of replacing the induction motor, the hydraulic coupling, and the synchronous generator with the voltage source inverter and a circuit breaker through electrically disconnection and removal.

Abstract: Disclosed embodiments include nuclear fission reactor cores, nuclear fission reactors, methods of operating a nuclear fission reactor, and methods of managing excess reactivity in a nuclear fission reactor.

Abstract: Disclosed is an offshore fixed, moored, or mobile energy carrier production plant. The plant's energy source for the energy intensive processes of producing an energy carrier is nuclear in nature and the resulting energy carriers of the plant range from hydrogen to hydrocarbons such as methanol and jet fuel. The offshore energy carrier production plant will be able to produce energy carriers at a reduced cost, increased sustainability and scalability, increased safety, and with fewer environmental and social impacts than heretofore possible. The resulting energy carrier products can then be transported by marine vessels, pipelines, and other transportation means or any combination of transportation means thereof to be distributed to end use energy carrier consuming devices and products such as fuel cell applications in a variety of industries as well as internal combustion engines.

Abstract: A nuclear power plant and method of operation for augmenting a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in the previous operation cycle. The plant is equipped, for example, with a reactor; a steam loop comprising high and low pressure turbines; a condenser for condensing steam discharged therefrom the low pressure turbine; a feedwater heater for heating feedwater supplied from the condenser; and a feedwater loop for leading feedwater discharged from the feedwater heater to the reactor. The operation method includes decreasing a ratio of extraction steam which is led to the feedwater heater from a steam loop in the second operation cycle.

Abstract: A control loop includes a pressure vessel, a condenser and a piping provided between the pressure vessel and the condenser. The piping includes a steam pipe for transferring dry air from the pressure vessel into the condenser, a turbine bypass pipe for transferring wet air from the pressure vessel into the condenser; a water-feeding pipe for transferring a first amount of water from the condenser into the pressure vessel and a pump bypass pipe for transferring a second amount of water from the condenser into the pressure vessel. Turbines are provided on the steam pipe and driven by the dry air. A turbine-controlling valve is provided on the steam pipe and operable to control the flow rate of the dry air. A turbine bypass valve is provided on the turbine bypass pipe and operable to control the flow rate of the wet air.

Abstract: In order to provide a reactor power control apparatus which can maintain a stable water level when reactivity control based on control rod operation and water level adjustment is performed, this invention comprises a turbine control which calculates the load set error signal from the error between the set target generator power value and the generator power that was fed back and outputs to the turbine control apparatus; control rod control in the natural circulation boiling water reactor which calculates the control rod operation signal and outputs it to the control rod drive control apparatus; water level control inside the natural circulation boiling water reactor which calculates the water level set signal and outputs it to the feed water control apparatus; and the switch determining device which selectively outputs one of the control rod control and water level control based on the generator power that was fed back, the reactor power and water level, as well as switching rules and determination values.

Abstract: The present invention relates to a method of controlling a nuclear reactor during a transient period. The method includes actuating the steam dump system in response to a temperature error signal and a power mismatch signal.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: A pH adjusting system includes a water pit for refueling 35 that is disposed in a reactor container 10, which stores a reactor, and is capable of storing cooling water and a pH adjusting apparatus 37 that is disposed above the water pit for refueling 35 and houses a pH adjuster. The pH adjusting apparatus 37 causes a pH-adjusted solution generated by dissolving or mixing the pH adjuster to flow out to the water pit for refueling 35 below the pH adjusting apparatus 37. Consequently, even when it is difficult to dispose the pH adjusting apparatus on a floor of the water pit for refueling, it is possible to suitably perform pH adjustment in the reactor container 10.

Abstract: Control rod guide tubes for a nuclear reactor having a body with an axial length that defines a lower end portion and an upper end portion and a cavity within a substantial length of the body. Orifices are included at the upper and lower end portions of the body. A control rod chamber is located within the cavity and is configured for receiving a control rod. A plurality of ports is coupled to the cavity and is positioned at a substantial length from the upper end portion of the body. Also included are at least two flow channels within the cavity that extend a substantial portion of the axial length of the body. Each flow channel is fluidly coupled to one or more of the ports for receiving fluid flow from outside the body and an outlet proximate to the upper end portion of the body for providing the received fluid flow.

Abstract: The present invention is directed to a cooling system for removing heat from a Fischer-Tropsch (F-T) slurry reactor. The cooling system including at least one downcomer having an upper portion, a lower portion, and a diameter; the at least one downcomer disposed within the F-T reactor to deliver a coolant downward through the F-T reactor at a predetermined velocity, the at least one downcomer extending a length within the F-T reactor wherein the coolant is introduced into the upper portion of the at least one downcomer in a substantially liquid phase; the diameter of the at least one downcomer and the pressure of the introduced coolant cooperate to increase the coolant velocity thereby generating backpressure in the at least one downcomer to maintain the coolant in the substantially liquid phase along the length of the at least one downcomer.

Abstract: A method of operating a nuclear reactor is disclosed. The reactor (1) encloses a core having a plurality of fuel rods (9). Each fuel rod (9) includes a cladding and fuel pellets of a nuclear fuel. The fuel pellets are arranged in an inner space of the cladding leaving a free volume comprising an upper plenum, a lower plenum and a pellet-cladding gap. The reactor is operated at a normal power and a normal inlet sub-cooling during a normal state. The reactor is monitored for detecting a defect on the cladding of any of the fuel rods. The operation of the reactor is changed to a particular state after detecting such a defect. The particular state permits an increase of the free volume in the defect fuel rod. The reactor is operated at the particular state during a limited time period, after which the reactor is operated at the normal state.

Abstract: The present invention related to a novel process for accelerating the breeding and conversion of fissile fuel in various types of nuclear reactors. In said process a movable nuclear fuel ball bed filled with a coolant creeps through the reactor core. The said process could provide higher specific power per unit fuel volume inside the reactor core and proceed on-line refueling, thus requires much less initial fissile fuel inventory per unit power output as compared with the traditional breeding or conversion reactors. The said process has full inherent safety characteristics and could follow the external power demand with the inherent negative temperature coefficient of reactivity. The said process, therefore, is a more efficient and economic approach to meeting the enormous demand of fissile fuel for the forthcoming “second nuclear era”.

Abstract: Exemplary embodiments provide automated nuclear fission reactors and methods for their operation. Exemplary embodiments and aspects include, without limitation, re-use of nuclear fission fuel, alternate fuels and fuel geometries, modular fuel cores, fast fluid cooling, variable burn-up, programmable nuclear thermostats, fast flux irradiation, temperature-driven surface area/volume ratio neutron absorption, low coolant temperature cores, refueling, and the like.

Abstract: A radial flow reactor fluid duct distribution apparatus is presented. The apparatus includes a plate of sufficient thickness to impart strength to the fluid flow duct and is milled to have narrow slots allowing fluid to flow through the plate, while preventing the passage of catalyst through the plate.

Abstract: An method for operating a nuclear power generation plant, comprising the steps of: forming a plurality of control rod patterns by operating a plurality of control rods during a first period of one operation cycle of a reactor including said first period before a point of time when all control rods are completely withdrawn from a core of said reactor and a core flow rate reaches firstly a set core flow rate, and a second period after said point of time, controlling stepwise at least once a temperature of feed water supplied to said reactor based on a different set feed water temperature during a period included in said first period for operating said reactor with a formed same control rod pattern, and continuing feed water temperature control based on said set feed water temperature until said core flow rate reaches a set core flow rate set based on said set feed water temperature.

Abstract: The invention relates to a nuclear reactor primary circuit comprising a primary pipeline (30), which defines an internal volume (32) and in which a primary nuclear reactor coolant downwardly runs, an additional pipeline (26) which is branched to the primary pipeline (30) and defines an internal volume communicating with the internal volume (32) of the primary pipeline (30) and a cuff (36) whose first end (50) is connected to the additional pipeline (26) and the second free end (52) is positioned in the internal volume (32) of the primary pipeline (30). According to said invention, the second end (52) is delimited by a free peripherial edge (53) comprising at least one upstream and downstream sections (56, 58) which are oriented towards the upstream, wherein the upstream section (56) penetrates into the internal volume (32) deeper from the primary pipeline (30) than the downstream section (58).

Abstract: The subject of the present invention is a method for producing a gas separation membrane, comprising the deposition of a film from a silica sol onto a porous support followed by heat treatment of the film thus deposited, in which the silica sol deposited is prepared by hydrolysing a silicon alkoxide in the presence of a doping amount of a precursor of an oxide of a trivalent element, especially boron or aluminium. The invention also relates to the membranes as obtained by this method, and also to their uses, especially for the separation of helium or hydrogen at high temperature, and in particular for removing impurities in helium streams.

Abstract: In various embodiments, a system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is provided. In various embodiments, the system includes a heating subsystem for heating feedwater flowing into an annulus of a NCBWR to increase the temperature of recirculation water flowing through the core above a predetermined recirculation water operating temperature. Additionally the system includes a temperature sensor operable to sense the temperature of the feedwater flowing into the annulus. The temperature sensor is communicatively connected to a temperature controller operable to command the heating subsystem to increase the temperature of the feedwater flowing into the annulus to a requested temperature above a predetermined operating temperature of the feedwater flowing into the annulus.

Abstract: The present invention decreases the temperature of feed water supplied to the reactor of a set power when the flow rate of coolant supplied to the core of the reactor increases in the end of an operation cycle. This operating method can increase the thermal power of the nuclear power generation plant and increase the economical efficiency of fuel even when the operation cycle is prolonged. Particularly, even when the core flow rate increases in the end of the operation cycle, this method can suppress the rise of the cooling water temperature at the inlet of the core. Consequently, this invention can make the reactivity gain higher than that when the core flow rate is singly increased. The present invention can increase the thermal power of a nuclear reactor, and can improve the economical efficiency of fuel even when a period of an operation cycle is made longer.

Abstract: A nuclear reactor (1) is provided, in particular a nuclear reactor cooled with a liquid metal (for example, a heavy liquid metal, such as lead or lead-bismuth eutectic), of the type having a cylindrical inner vessel (15) that divides a hot collector (6) over a core (4) from a substantially annular cold collector (7) surrounding the hot collector; housed in the cold collector (7) is a plurality of integrated circulation and heat-exchange assemblies (20), each of which includes a pump (9), two heat exchangers (10) set at the sides of the pump, and a conveying structure (21) through which a primary fluid (8) for cooling the core passes from the pump to the heat exchanger, all of which are fixedly connected to one another to form a unitary structure; each integrated assembly has an inlet (26) connected to the hot collector (6) and two outlet sections (34) in the cold collector (7).

Abstract: A fast reactor having a reflector control system is provided which decreases the change in reactivity of the reactor core with time without performing control of a reflector lifting speed and that of a water flow rate. The above fast reactor has a liquid metal coolant, a reactor core immersed therein, and a neutron reflector which is provided outside the reactor core and which is moved in a vertical direction for adjusting leakage of neutrons therefrom for controlling the reactivity of the reactor core. The neutron reflector described above is gradually moved in an upward direction with the change in reactivity caused by fuel burn-up, and at least a part of a lower region of the neutron reflector is a high reflection region having a high neutron reflection ability as compared to that of the other region. The high reflection region is located from the bottom to a place between one fourth and one half of the height of the neutron reflector from the bottom end thereof.

Abstract: A reflector control type fast reactor 1 comprises: a reactor vessel 7 accommodating therein a primary coolant 5; a reactor core 2 disposed in the reactor vessel 7 and immersed in the primary coolant 5; and a reflector 4 that vertically moves for adjusting leakage of neutrons generated from the reactor core 2 to control a reactivity of the reactor core 2, the reflector 4 including a neutron reflecting part 4a disposed on an outside of the reactor core 2 in a vertically movable manner, the neutron reflecting part 4a having a neutron reflecting ability higher than that of the primary coolant 5, and a cavity part 4b positioned above the neutron reflecting part 4a, the cavity part 4b having a neutron reflecting ability lower than that of the primary coolant 5. The neutron reflecting part 4a is formed of a plurality of metal plates 37 that are stacked on each other. Each of the metal plates 37 has a plurality of coolant channels 36 through which the primary coolant 5 flows.

Abstract: Methods and apparatus for operating a combined-cycle power system are provided. The method includes operating the steam turbine, the combustion turbine, and the steam source at steady state operating conditions. Upon sensing a grid frequency deviation away from the standardized grid frequency value, determining a current thermal energy capacity of the steam source, determining a rate of frequency recovery available using the current thermal energy capacity of the steam source and a predetermined rate of change of the at least one steam turbine control valve, if the determined rate of frequency recovery available is greater than the grid frequency deviation, mitigating the frequency deviation using the current thermal energy capacity, if the determined rate of frequency recovery available is less than the grid frequency deviation then mitigating the frequency deviation using the current thermal energy capacity substantially simultaneously with a power level increase of the combustion turbine.

Abstract: A method of modeling a heat exchanger is disclosed and comprises assigning input temperatures, assumed output temperatures, and a set of flow rates, inputting the parameters into a set of equations arranged to calculate a heat transfer coefficient, inputting parameters into a second set of equations arranged to calculate output temperatures, substituting actual output temperatures for the assumed output temperatures, and again calculating the heat transfer cooefficient. The new heat transfer coefficient is then used to obtain revised actual output temperatures, and the initial actual output temperatures and the revised actual output temperatures are compared to determine whether they differ by less than a desired variance. If not, a new iteration is performed until the output temperatures converge.

Abstract: A variable-voltage variable-frequency power source for an electric motor that drives a re-circulation pump for a boiling water nuclear reactor. The power source has a semiconductor electric power converter and a speed controller for controlling the semiconductor electric power converter. When a part of the semiconductor electric power converter comes into an inoperative state, that part is electrically disconnected, and the variable-voltage variable-frequency power source temporarily stops outputting power, thereby idling the re-circulation pump. Thereafter, the power source re-starts outputting power before the re-circulation pump completely stops. Thus, the re-circulation pump keeps operating, without stopping.

Abstract: A resistance member (e.g., fuel holding portion of the lower tie plate) is provided at the lower end of the fuel assembly. Provision is made of a coolant ascending path in which said water rods have coolant inlet ports that are open in a region lower than the resistance member to upwardly guide the coolant, and a coolant descending path which has a coolant delivery port that is open in a region higher than the resistance member to downwardly guide the coolant. The coolant ascending path and the coolant descending path are communicated with each other at their upper end portions.

Abstract: An apparatus and method for providing cooling to a magnetic circuit element having a magnetic core disposed around a centrally located core support member having at least one core support member wall is disclosed which may comprise a core support coolant inlet; a core support coolant outlet; a plurality of interconnected coolant flow passages contained within the core support member wall and interconnected and arranged to pass coolant from one coolant flow passage to the next within the core support member wall along a coolant flow path within at least a substantial portion of the core support member wall from the core support coolant inlet to the core support coolant outlet.

Abstract: A main steam system around a nuclear reactor which comprises two main steam nozzles in a reactor pressure vessel, each of the main nozzles being disposed at a symmetrical position with respect to a plane parallel to steam outlet faces of steam dryers provided inside the reactor pressure vessel and passing through a center of the reactor pressure vessel; and main steam pipes each connected to the main steam nozzles.

Abstract: An improved method of increasing the power output of an existing nuclear power plant includes increasing the thermal power output of the plant's nuclear island and constructing of an auxiliary BOP to handle the increased thermal power. The thermal power of the nuclear island can be increased such as by increasing the thermal power of the plant's reactor, by replacing the plant's steam generator with one that is more efficient, and by increasing the flow rate and/or change in temperature of a coolant in a secondary cooling loop of the plant. The thermal power of the reactor can be increased such as by replacing existing cylindrical fuel rods with fuel rods having a relatively greater surface area to volume ratio and/or by increasing the flow rate and/or the change in temperature of a coolant of a primary cooling loop. The auxiliary BOP can be constructed while the plant is in operation, and can then be connected with the nuclear island during a maintenance operation on the reactor.

Abstract: The invention relates to a method of controlling an average temperature of a coolant at a reactor core outlet. The method includes the steps of detecting an actual average temperature of the coolant at the reactor core outlet, comparing the actual average temperature of the coolant at the reactor core outlet with a reference temperature thereby to generate an error signal, and adjusting the actual average temperature of the coolant at the reactor core outlet in response to the error signal. The invention extends to a nuclear reactor outlet temperature control system 10, to a cascade controller for a nuclear reactor, and to a nuclear power plant.

Abstract: The invention relates to a method for operating a steam power installation, whereby steam produced in a boiler is condensed in a condenser after passing through at least one turbine, and the condensate obtained is preheated and redirected back to the boiler as boiler feed-water. In order to preheat the condensate, said condensate is split into a first partial current and a second partial current. Only the first partial current is preheated and the second partial current is then mixed with the preheated first partial current. The power of the turbine can thus be increased as required, up to the boiler reserve of the steam power installation.

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 method for expanding the operating domain of a boiling water nuclear reactor that permits safe operation of the reactor at low core flows is described. The operating domain is characterized by a map of the reactor thermal power and core flow. In an exemplary embodiment, the method for expanding the operating domain of a boiling water nuclear reactor permits operation of the reactor between about 120 percent of rated thermal power and about 85 percent of rated core flow to about 100 percent of rated thermal power and about 55 percent of rated core flow. The method includes determining an elevated load line characteristic that improves reactor performance, performing safety evaluations at the elevated load line to determine compliance with safety design parameters, and performing operational evaluations at the elevated load line. The method also includes defining a set of operating conditions for the reactor in an upper operating domain characterized by the elevated load line.

Abstract: This invention relates to methods of minimizing catalyst degradation during the handling of a catalyst used in a slurry phase reactor. The methods include catalyst handling steps such as catalyst loading into a slurry phase reactor, slurry phase reactor start-up, slurry phase reactor shut-down, and slurry phase reactor unloading when catalyst reloading is envisaged. In the method of loading the slurry phase reactor, a slurry of wax and catalyst is formed in a loading vessel. Clean molten wax is formed in the reactor, syngas is pumped through the clean molten wax in the reactor, and the slurry from the loading vessel is transferred to the reactor.

Abstract: A steam turbine control device. A main steam system runs between a nuclear reactor and a steam turbine and includes a main steam isolation valve (MSIV), a main steam control valve, and a turbine by-pass system with a turbine by-pass valve A calculating means generates a main steam pressure control signal and a reactor dome pressure control signals. A pressure control signal changeover means changes a pressure control signal from the reactor dome pressure signal to the main steam pressure signal when the MSIV closes to restrain the abrupt decrease of steam in the main steam system effectively.

Abstract: There is provided a nuclear power plant having a steam turbine controller. The nuclear power plant includes a main steam supply system and a turbine by-pass system. The main steam supply system guides steam from heat generated by the nuclear reactor to a steam turbine. The turbine by-pass system is branched from the main steam supply system at a main steam header. A main steam control valve is equipped with the main steam supply system and adjusts steam pressure in the main steam supply system supplied to the steam turbine. A turbine by-pass valve is used to by-pass steam to the turbine by-pass system. A regulating controller generates first and second opening/closing signals for the main steam control valve and the turbine by-pass valve, and an ON-OFF controller generates a third opening/closing signal for the turbine by-pass valve. The third opening/closing signal has priority over the second opening/closing signal.

Abstract: The present invention is used to reduce thermal load itself, being the cause to generate stress, which develops near liquid surface in a nuclear reactor wall and to contribute to further improvement of safety. A partition member (5) is arranged above a coolant liquid surface (9) in an annulus space (3) between a reactor vessel (1) and a guard vessel (2), a low-temperature gas is circulated through the annulus space above the partition member to cool down, the gas is circulated through the annulus space from under the coolant liquid surface to the partition member, and the high-temperature gas heated under the coolant liquid surface is used to raise the temperature above the coolant liquid surface.

Abstract: There is provided a nuclear power plant having steam turbine controller. The nuclear power plant includes a main steam supply system and a turbine by-pass system. The main steam supply system guides steam from heat generated by the nuclear reactor to a steam turbine. The turbine by-pass system is branched from the main steam supply system at a main steam header. A main steam control valve is equipped with the main steam supply system and adjusts steam pressure in the main steam supply system supplied to the steam turbine. A turbine by-pass valve is used to by-pass steam to the turbine by-pass system. A regulating controller generates first and second opening/closing signals for the main steam control valve and the turbine by-pass valve, and an ON-OFF controller generates a third opening/closing signal for the turbine by-pass valve. The third opening/closing signal has priority over the second opening/closing signal.

Abstract: This invention relates to a nuclear reactor plant which includes a nuclear heat source (12), and a cooling system (10). The cooling system (10) includes at least two cooling circuits (26), each of which includes a plurality of coolant chambers (18) each having an inlet (40) and an outlet (42), the coolant chambers (18) being arranged around the nuclear heat source (12). The cooling system (10) further includes pump means (52) for pumping coolant to and from the coolant chambers (18), the volumetric capacity of the coolant chambers (18) being sufficiently large such that, when in a passive mode, the temperature of a water coolant in the coolant chambers (18) will remain below boiling point for at least eight hours. The invention extends to a cooling system, to a method of cooling a nuclear heat source, to a method of constructing a nuclear reactor plant and to a method of operating a nuclear reactor plant.

Abstract: A method of controlling a nuclear power plant includes determining a first maximum linear heat generation rate and a first minimum critical power ratio by a core monitoring system, and determining a second maximum linear heat generation rate and a second minimum critical power ratio by utilizing the first maximum linear heat generation rate and the first minimum critical power ratio determined by the core monitoring system and plant data. The second maximum linear heat generation rate and the second minimum critical power ratio are compared with predetermined thermal limit values and a control signal outputted from an automatic power regulator system to a re-circulation flow control system and to a control rod control system is held when at least one of the second maximum linear heat generation rate and the second minimum critical power ratio exceeds the predetermined thermal limit values.

Abstract: A fuel bundle assembly for a boiling water nuclear reactor comprising a plurality of fuel rods having respective fuel columns therein, and arranged in an ordered array, extending between upper and lower support structures, the plurality of fuel rods enclosed within a hollow, open-ended channel member at least partially enclosed by the open-ended channel member; at least one water rod supported on the lower tie plate and extending upwardly toward the upper tie plate, the at least one water rod having an upward flow path including at least one inlet at a lower end of the upward flow path, and a downward flow path including at least one outlet at a lower end of the downward flow path, the at least one outlet located about midway along the fuel columns within the fuel rods.

Abstract: A control system, and a control method, of a nuclear power plant capable of easily executing automatic output regulation by an automatic power regulator system even when a lower-accuracy maximum linear heat generation rate and a lower-accuracy minimum critical power ratio are determined in a short cycle by using brief calculation. While a core monitoring system does not calculate a maximum linear heat generation rate and a minimum critical power ratio, they are determined by utilizing the maximum linear heat generation rate and the minimum critical power ratio calculated by the core monitoring system as well as plant data, and the lower-accuracy maximum linear heat generation rate and the lower-accuracy minimum critical power ratio so determined are compared with predetermined thermal limit values.

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, the at least one input signal responsive to properties of the fluid flow in advance of the valve, and a letdown back-pressure controller receiving the at least one input signal and providing a tunable output signal for selectively opening and closing the valve to vary flow in the flow line responsive to the at least one input signal.

Abstract: The object of the present invention is to provide a valve characterized by excellent corrosion and wear resistance and maintainability by bonding corrosion and wear proof alloy containing non-continuously distributed eutectic carbide to the sliding portions of various types of apparatuses and valves by diffusion bonding. Another object of the present invention is to improve the maintainability of a thermal and nuclear power plant and to provide a nuclear power plant comprising the recirculating water which ensures excellent working safety, in particular.