A method for estimating a quantity of a system includes the following steps: for each variable of a plurality of variables, obtaining a sequence of successive measurements, determining a sequence of successive values of a stability parameter, the parameter being a weighted sum of rates of variation of the variables, identifying a time interval in which the stability parameter is less than or equal to a predetermined threshold for a duration greater than or equal to a predetermined duration, estimating a sequence of successive estimates of a particular variable, a start time of the sequence being included in the time interval, comparing the sequence of estimates and the sequence of measurements, in order to determine an adjustment parameter, and estimating the physical quantity of the system using the adjustment parameter.

The supply amount of reactive power can be expanded while the soundness of a nuclear reactor and a BOP. A control apparatus of a power generation plant connected to a power system including a power system stability degree previous evaluation unit that evaluates a stability degree at the time of the predicted failure of the power system, a nuclear power safety evaluation unit, and a current day power generation control instruction unit that corrects a required power supply amount given from the outside according to the evaluation result of the power system stability degree previous evaluation unit and the evaluation result of the nuclear power safety evaluation unit, in which the generated power of the power generation plant is adjusted by a signal from the current day power generation control instruction unit.

The present invention relates to an apparatus for efficiently and economically generating a cooling medium by using high-temperature and high-pressure steam generated in a nuclear power plant, and cooling method therefor. According to one embodiment of the present invention, the cooling medium generating apparatus provided in a containment vessel of a nuclear power generation facility so as to generate the cooling medium can comprise: a nuclear reactor for heating a coolant by using heat included in the heated coolant; a cooling module for generating the cooling medium by using the steam generated in the steam generator; and a cooling medium supplying pipe of which the end portion is connected to the outside of the containment vessel so as to supply the cooling medium, having been generated in the cooling module, to the outside of the containment vessel.

A high temperature gas cooled reactor steam generation system (1) includes a nuclear reactor (2) that has helium gas as a primary coolant and heats the primary coolant by heat generated by a nuclear reaction that decelerates neutrons by a graphite block, a steam generator (3) that has water as a secondary coolant and heats the secondary coolant by the primary coolant via the nuclear reactor (2) to generate steam, a steam turbine (4) that is operated by the steam from the steam generator (3), and a generator (5) that generates electricity according to an operation of the steam turbine (4). Moreover, the system (1) includes pressure adjustment means for setting a pressure of the secondary coolant in the steam generator (3) to be lower than a pressure of the primary coolant in the nuclear reactor (2).

A high temperature gas cooled reactor steam generation system (1) includes a nuclear reactor (2) that has helium gas as a primary coolant and heats the primary coolant by heat generated by a nuclear reaction that decelerates neutrons by a graphite block, a steam generator (3) that has water as a secondary coolant and heats the secondary coolant by the primary coolant via the nuclear reactor (2) to generate steam, a steam turbine (4) that is operated by the steam from the steam generator (3), and a generator (5) that generates electricity according to an operation of the steam turbine (4). Moreover, the system (1) includes pressure adjustment means for setting a pressure of the secondary coolant in the steam generator (3) to be lower than a pressure of the primary coolant in the nuclear reactor (2).

A condition determination system includes: an operation condition data obtaining unit that obtains operation condition data indicating an operation condition of a facility; and a determination unit that determines, based on the operation condition data, a level of a phenomenon that occurs due to the operation condition of the facility.

A condition determination system includes: an operation condition data obtaining unit that obtains operation condition data indicating an operation condition of a facility; and a determination unit that determines, based on the operation condition data, a level of a phenomenon that occurs due to the operation condition of the facility.

This invention relates to a control method for a pressurized water nuclear reactor, which comprises a core generating thermal power and means of acquiring magnitudes representative of core operating conditions. The method comprises a step to regulate the temperature of the primary coolant, if the temperature of the primary coolant for a given thermal power is outside a predefined set temperature interval (TREF) depending on the reactor power. The set temperature interval (TREF) is characterized by variable amplitude (T) on a thermal power range between N % and 100% nominal power, where N is between 0 and 100 and comprises a zero amplitude at 100% nominal power, a zero amplitude at N % nominal power.

This invention relates to a control method for a pressurized water nuclear reactor, which comprises a core generating thermal power and means of acquiring magnitudes representative of core operating conditions. The method comprises a step to regulate the temperature of the primary coolant, if the temperature of the primary coolant for a given thermal power is outside a predefined set temperature interval (TREF) depending on the reactor power. The set temperature interval (TREF) is characterized by variable amplitude (T) on a thermal power range between N % and 100% nominal power, where N is between 0 and 100 and comprises a zero amplitude at 100% nominal power, a zero amplitude at N % nominal power.

A nuclear reactor power system includes: a reactor core comprising a plurality of nuclear fuel elements, each nuclear fuel element comprising: a first cooling channel passing through the nuclear fuel element; and a second cooling channel passing through the nuclear fuel element and fluidly isolated from the first cooling channel; a first cooling system configured to transport a first fluid coolant through the reactor core, the first cooling system fluidly connected to the first cooling channel of each nuclear fuel element; and a second cooling system configured to transport a second fluid coolant through the reactor core, the second cooling system fluidly connected to the second cooling channel of each nuclear fuel element. A direction of first fluid coolant flow through the first cooling channel is the same as a direction of second fluid coolant flow through the second cooling channel.