Abstract: In view of above problems, an object of the invention is to provide a primary containment vessel venting system having a structure capable of continuously discharging vapor in a primary containment vessel out of the system and continuously reducing pressure of the primary containment vessel without discharging radioactive noble gases to the outside of the containment vessel and without using an enclosing vessel or a power source. In order to achieve the above object, an nuclear power plant of the invention includes a primary containment vessel which includes a reactor pressure vessel, a radioactive substance separation apparatus which is disposed inside the primary containment vessel and through which the radioactive noble gases do not permeate but vapor permeates, a vent pipe which is connected to the radioactive substance separation apparatus, and an exhaust tower which is connected to the vent pipe and discharges a gas, from which a radioactive substance is removed, to the outside.

Abstract: In view of above problems, an object of the invention is to provide a primary containment vessel venting system having a structure capable of continuously discharging vapor in a primary containment vessel out of the system and continuously reducing pressure of the primary containment vessel without discharging radioactive noble gases to the outside of the containment vessel and without using an enclosing vessel or a power source. In order to achieve the above object, an nuclear power plant of the invention includes a primary containment vessel which includes a reactor pressure vessel, a radioactive substance separation apparatus which is disposed inside the primary containment vessel and through which the radioactive noble gases do not permeate but vapor permeates, a vent pipe which is connected to the radioactive substance separation apparatus, and an exhaust tower which is connected to the vent pipe and discharges a gas, from which a radioactive substance is removed, to the outside.

Abstract: An operator of a nuclear plant is provided with an action margin time up to the execution of a next action to be performed in response to a detected accident based on sensor signals received from sensors for the plant. The operator is provided with a displayed graph showing the received sensor signals, estimated sensor signals for the future, the action margin time and an action threshold value corresponding to the next action. Accordingly, the operator can readily grasp how much time is available to perform the next action before the received sensor signals are predicted to exceed the action threshold value.

Abstract: An operator of a nuclear plan is provided with a margin time up to the execution of the action at the time of an accident. The system includes: a sensor signal input device; an accident event estimation device; a procedure manual database; a procedure manual selection device; a procedure automatic check device that checks items in the procedure manual and determines a next action using the sensor signal; a plant state analysis device that analyzes a state of the plant in a case where the actions are executed at a plurality of action timings based on the accident event and the next action, and calculates predicted values of the sensor signals; an action threshold value database in which an action threshold values of the sensor signals corresponding to the action procedure at the time of the accident are stored; and an action margin time calculation device.

Abstract: An object of the invention is to provide a system for assisting operation at the time of a plant accident, which, even if a device malfunctions at the time of the plant accident, can determine normality of a sensor without being influenced by the malfunction. A system for assisting operation at the time of a plant accident, includes: a phenomenon identification apparatus for identifying a state of a plant; a plant behavior analysis apparatus for predicting progress of a process state after a phenomenon occurs by using at least a phenomenon identification result as an initial condition of plant behavior analysis; and a sensor normality determination apparatus for determining normality of a sensor by comparing a result of prediction of the progress of the process state which is output from the plant behavior analysis apparatus with a sensor signal.

Abstract: To eliminate or reduce stress for an operator who operates an operation panel. A touch panel-type operation panel includes a display with touch panel including an image display surface for displaying an operation part and a control device. The display with touch panel detects approach of an operation finger to the operation part and the position of the operation part and outputs signals of the detected approach of the operation finger and the detected position. The control device predicts and analyzes, in response to an input of the signals of the approach of the operation finger and the position, a state of a control target by operation corresponding to the operation part with reference to state data of the control target, determines appropriateness of the predicted and analyzed operation, and outputs a result of the determination.

Abstract: An object of the invention is to provide an operation support system for plant accidents that estimates events reflecting plant states changing momentarily at plant accidents. The invention includes an event narrow-down device that narrows down occurring event candidates based on at least one of a sensor signal, a device state signal, and an alarm signal and a discriminant rule, an event analysis device that analyzes a plant behavior based on a plurality of event narrow-down results as output of the event narrow-down device, the sensor signal, the device state signal, and the alarm signal, and an event estimation device that estimates an occurring event by comparing an analysis result from an analysis of a process state as output of the event analysis device and the sensor signal, wherein the event estimation device outputs the analysis result of the process state and the sensor signal corresponding to the occurring event as an event estimation result.

Abstract: An evaluation chart has test patterns, a first code mark, and a second code mark. The evaluation chart is electronically read at one time to obtain first code information that specifies measuring conditions for the evaluation chart and second code information that specifies image forming conditions for the evaluation chart. Quantized information of the test patterns that is measured under the measuring conditions is associated with the image forming conditions.

Abstract: In order to stably control a nuclear reactor in a short time, so as not to enter an unstable region that is determined by the relationship between the reactor pressure, the reactor power and the subcooling of the core inlet coolant at start-up time, the nuclear reactor system comprises: an power control apparatus for generating a control rod operation signal for operating a control rod, based on the reactor water temperature change rate; a feed water control apparatus for generating a feed water flow rate signal and a discharge water flow rate signals based on the reactor water level signal; and a process computer for performing overall control of the power control apparatus and the feed water control apparatus, wherein the feed water control apparatus has the reactor water temperature change rate setting section for adjusting the reactor water temperature change rate set value based on the variation of the reactor water level signal.

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: In order to stably control a nuclear reactor in a short time, so as not to enter an unstable region that is determined by the relationship between the reactor pressure, the reactor power and the subcooling of the core inlet coolant at start-up time, the nuclear reactor system comprises: an power control apparatus for generating a control rod operation signal for operating a control rod, based on the reactor water temperature change rate; a feed water control apparatus for generating a feed water flow rate signal and a discharge water flow rate signals based on the reactor water level signal; and a process computer for performing overall control of the power control apparatus and the feed water control apparatus, wherein the feed water control apparatus has the reactor water temperature change rate setting section for adjusting the reactor water temperature change rate set value based on the variation of the reactor water level signal.

Abstract: A natural circulation boiling water reactor includes a reactor pressure vessel, a chimney including a lattice member and arranged above a core in said reactor pressure vessel, and a plurality of thermocouple extension wire pulling conduits into which at least one temperature detection thermocouple and at least one cable connected to the temperature detection thermocouple are inserted. At least one of thermocouple extension wire pulling conduits is disposed on an upper end surface of the lattice member and is directly mounted to the upper end surface of the lattice member.

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: A natural circulation boiling water reactor includes a reactor pressure vessel, a chimney including a lattice member and arranged above a core in the reactor pressure vessel, and at least one thermocouple extension wire pulling conduit into which a temperature detection thermocouple and a cable connected to said temperature detection thermocouple are inserted. The thermocouple extension wire pulling conduit is disposed on an upper end surface of the lattice member and mounted to the upper end surface of the lattice member.

Abstract: A reactor start-up monitoring system, comprising: a determination apparatus for determining that moderator temperature reactivity coefficient is positive based on neutron flux measured by a neutron detector and a reactor water temperature measured by a temperature detection apparatus; an output information creating apparatus for creating first output information indicating positive moderator temperature reactivity coefficient when determination information inputted from the determination apparatus indicates the positive moderator temperature reactivity coefficient; and at least one of a display apparatus and an audio output apparatus for inputting the first output information.

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: A chimney lattice is arranged in a reactor pressure vessel. Thermocouple extension wire pulling conduits inserting a temperature detection thermocouple and a cable connected to the thermocouple are mounted to the upper end of the chimney lattice. By mounting the thermocouple extension wire pulling conduit on the upper end of the chimney lattice, the thermocouple extension wire pulling conduit does not become an obstacle when the fuel assembly is taken out from the chimney. Because the thermocouple extension wire pulling conduit is installed at the upper end of the chimney, the replacement of the thermocouple damaged is easy.

Abstract: A fuel assembly comprises a plurality of fuel rods which contain nuclear fuel material inside, a lower tie plate which holds the lower end of the fuel rods and has a path inside to lead coolant between the fuel rods, and a channel box which encloses a bundle of the fuel rods. An orifice, in which a plurality of round rods are arranged to cross the coolant flow path, is installed in a through hole at a side wall of the lower tie plate by connecting to the side wall. Orifice coefficient of the orifice becomes large at small flow rate of coolant which supplied to the fuel assembly, and becomes small at large flow rate of coolant. By using the fuel assembly described above, void fraction in a gap region between fuel assemblies can be altered during beginning and end of an operation cycle of the nuclear reactor.

Abstract: A high conversion nuclear reactor core has fuel assemblies made up from large numbers of axially-extending uranium-plutonium mixed oxide fuel rods. The fuel rods are densely packed so as to give a high conversion ratio of fissile substances, preferably approaching unity. The average plutonium enrichment in the assemblies is higher in their bottom, upstream halves, than in their top downstream halves. This has the effect of reducing a potentially dangerously high void coefficient in the core.

Abstract: A sequence valve is interposed between the manual valve and the reverse clutch and arranged to control the communication therebetween. The valve is responsive to the pressure which develops in the low and reverse brake and establishes communication between the manual valve and the reverse clutch when the pressure prevailing in the low and reverse brake is such as to indicate that the brake is engaged to a predetermined degree.