A method for controlling a nuclear reactor includes acquiring current values of operating parameters of the reactor; and iteratively implementing the sub-steps of generating a sequence of injection of neutron poison and/or water; calculating an evolution in at least one magnitude characteristic of the state of the core of the nuclear reactor during this given time interval using a power program, current values of operating parameters and the injection sequence considered, the evolution being calculated using a predictive model of the core of the reactor; evaluating a cost function, using the calculated evolution; repeating the generating and calculating sub-steps until a convergence criterion of the cost function is met; and repeating the acquiring and the iteratively implementing steps with a time period less than 60 minutes.

The present invention discloses a nuclear reactor coolant pump that does not rely on an electric motor, but is operated by means of driving force generated inside a nuclear power plant, so a to be capable of maintaining the safety of the nuclear reactor when the nuclear reactor is operating normally and also in the event of an accident in the nuclear reactor. The nuclear reactor coolant pump comprises: a pump impeller rotatably installed in a first fluid passage of a nuclear reactor coolant system to circulate a first fluid inside the nuclear reactor coolant system; a drive unit receiving steam from a steam generator to generate driving force to rotate the pump impeller, and rotating about the same rotating shaft as the pump impeller to transfer the generated driving force to the pump impeller; and a steam supplying unit forming a passage between the steam generator and the drive unit to supply at least a portion of the steam released from the steam generator to the drive unit.

The present invention is related to a Thorium Molten Salt Reactor (Th-MSR) using 100% non-radioactive Thorium fuel, composed of LiF+BeF.sub.2+ThF.sub.4 without containing any U-235. The Th-MSR is consisted of a reactor chamber, a fuel injector, a fuel reservoir, an in-line chemical extraction unit, a heat exchanger, a few KW electricity turbine generator and a condenser. A few KW nuclear power generation system is adopting the controlling devices comprised of a neutron flux sensor, a fuel injecting sensor, a thermal sensor and a power output sensor. A neutron generator with a high neutron flux of 10.sup.13 n/s is used. The high flux fast neutrons are slowing down into the thermal neutrons by the graphite moderators in the reactor for initiating the fission.

The present invention is related to a Thorium Molten Salt Reactor (Th-MSR) using 100% non-radioactive Thorium fuel, composed of LiF+BeF.sub.2+ThF.sub.4 without containing any U-235. The Th-MSR is consisted of a reactor chamber, a fuel injector, a fuel reservoir, an in-line chemical extraction unit, a heat exchanger, a few KW electricity turbine generator and a condenser. A few KW nuclear power generation system is adopting the controlling devices comprised of a neutron flux sensor, a fuel injecting sensor, a thermal sensor and a power output sensor. A neutron generator with a high neutron flux of 10.sup.13 n/s is used. The high flux fast neutrons are slowing down into the thermal neutrons by the graphite moderators in the reactor for initiating the fission.

The present invention relates to a method for controlling a pressurized water reactor (100) comprising the steps that involve measuring the effective power (Pe) of the nuclear reactor; acquiring a reference value for the desired power (Pc); acquiring an estimated duration (DURATION) for the increase in power in order to achieve said reference value of the target power (Pc) desired, said estimated duration (DURATION) corresponding to the time taken for the power to increase from said effective power (Pe) to said reference value for the target power (Pc); determining the reference position (Z) of at least one control rod cluster among said plurality of control rod clusters (40) in order to achieve said reference value for said target power (Pc) desired as a function of said estimated duration (DURATION), of said measured effective power (Pe) and of said reference value for said target power (Pc); monitoring the position of said at least one control rod cluster so as to position it in its reference position (Z).

The present invention relates to a method for controlling a pressurized water reactor (100) comprising the steps that involve measuring the effective power (Pe) of the nuclear reactor; acquiring a reference value for the desired power (Pc); acquiring an estimated duration (DURATION) for the increase in power in order to achieve said reference value of the target power (Pc) desired, said estimated duration (DURATION) corresponding to the time taken for the power to increase from said effective power (Pe) to said reference value for the target power (Pc); determining the reference position (Z) of at least one control rod cluster among said plurality of control rod clusters (40) in order to achieve said reference value for said target power (Pc) desired as a function of said estimated duration (DURATION), of said measured effective power (Pe) and of said reference value for said target power (Pc); monitoring the position of said at least one control rod cluster so as to position it in its reference position (Z).

A computer implemented method for simulating an operation of a reactor core includes determining an initial state of the reactor core; calculating a nodal target power distribution and/or the target 3D neutron flux distribution; obtaining an actual power distribution and/or the actual 3D neutron flux distribution of the nuclear reactor core; determining a difference between the target power distribution and the actual power distribution of the nuclear reactor core and/or determining a difference between the target 3D neutron flux distribution and the actual 3D neutron flux distribution of the nuclear reactor core; determining modal expansion coefficients using a Fourier modal decomposition based on the determined difference and applying a Modal Generalized Perturbation Theory to the modal expansion coefficients for determining a 3D cross-section distribution perturbation causing the determined difference; and determining a 3D adaptation distribution for the determined difference based on the determined 3D cross-section distribution perturbation.

In a device for and a method of reconstructing axial measurement values in a nuclear fuel, which is a device that calculates an axial reaction rate distribution by reconstructing a plurality of measurement values measured by a plurality of neutron flux detectors that are disposed at predetermined intervals in a fuel assembly along the axial direction of the fuel assembly, because a reconstruction parameter generator that generates a reconstruction parameter on the basis of core design data, or core analysis data, and a data adjustment factor; and an axial reaction rate distribution generator that calculates an axial reaction rate distribution on the basis of the measurement values that are measured by the neutron flux detectors and the reconstruction parameter that is generated by the reconstruction parameter generator are provided, an accurate axial measurement distribution in the nuclear fuel is obtained by reconstructing the measurement values.

A pressurized water reactor includes a primary reactor coolant circuit flown through by a primary reactor coolant during operation, and a chemical and volume control system for the primary reactor coolant. The chemical and volume control system includes, along the direction of flow of the primary reactor coolant, a letdown line, a high-pressure charging pump with a given discharge pressure, and a charging line leading to the primary reactor coolant circuit. The chemical and volume control system further includes a hydrogenation system with a hydrogen supply and a hydrogen feeding line. In order to achieve efficient and fast hydrogen injection into the primary reactor coolant, a high-pressure feeding pump is arranged in the feeding line to provide a gas pressure higher than the discharge pressure of the charging pump. The feeding line discharges into the charging line.

Systems and methods provide data gathering and execution on the same without human operations. Systems may include controls and sensors that electronically provide data and operations to a processor networked with the same. For a nuclear reactor, the processor may determine reactivity from the sensors and issue commands to actuators to operate the reactor. Reactivity may be determined based on all reactivity factors determined from the plant data, including the use of modelling. The processor may position control elements or moderator feeds to achieve a desired reactivity. The processor may be networked to plant switches and sensors, and multiple processors may be used to independently calculate and decide on plant operations. Human operator input is not required at discreet instances of plant operational change; systems may include displays and input interfaces to permit observation and/or intervention if absolutely necessary.