A method of controlling a plasma in a nuclear fusion reactor. The nuclear fusion reactor comprises sensors and plasma control inputs. An initial control model is provided, relating readings of at least a subset of the sensors to control of the plasma control inputs. A control loop is performed, comprising: operating the plasma control inputs in dependence upon the sensors according to the control model; determining correlations between readings of each of the sensors, and/or between readings of the sensors and states of the plasma control inputs; and adjusting the control model based on the determined correlations.

A method for determining at least one threshold value of at least one operating parameter of a nuclear reactor is implemented by an electronic determination system and includes the steps of determining a first threshold value of a respective operating parameter for an operation of the reactor at a first power; and determining a second threshold value of said parameter for an operation of the reactor at a second power. The operation at the lower power of the first and second powers is an operation continued for a duration of at least 8 hours over a 24-hour sliding window. The method also includes determining a third threshold value of said parameter for an operation of the reactor at a third power between the first power and the second power.

The invention relates to a device for controlling a plurality of nuclear reactors in clusters, comprising, for each reactor, a plurality of sensors for measuring operating parameters as well as a system for controlling the nuclear reactor, the nuclear reactors being grouped into clusters having a cluster head.

Heat can be stably extracted with easy criticality control. A nuclear reactor includes: a fuel portion being a reactor core having a nuclear fuel body; a shielding portion covering all over outer sides of the fuel portion to shield against radiations generated from the reactor core; and a thermal conduction part that conducts heat generated in the reactor core to exterior of the shield part. The nuclear fuel body contains a fissile material with an enrichment not less than 5% by weight throughout an operation period.

A nuclear-power-plant computer-based procedure display device is disposed in a main control room of a nuclear power plant, and includes a operating procedure storage unit that stores a computer-based procedure in which plant operation procedures of the nuclear power plant are divided into procedure steps and listed, a operating procedure display unit that displays the computer-based procedure, and a operating procedure display control unit that controls display of the computer-based procedure. In a case where the procedure step displayed on the operating procedure display unit is selected by an operator, the operating procedure display control unit displays an indication that the procedure step is selected, on the operating procedure display unit.

A computer-implemented simulation method of predicting local concentrations of constituents in coolant water anywhere along fuel rods within any fuel assembly mechanical design of a Boiling Water Reactor (BWR) potentially resulting in crud deposits on said fuel rods. The method is based on a sub-channel approach of predicting local mass fluxes of vapor and liquid in coolant water anywhere along fuel rods within any fuel assembly mechanical design of a Boiling Water Reactor (BWR) for given steady-state or transient boundary conditions. The sub-channel approach is based on the solution of mass, momentum and energy conservation equations for the vapor phase and the liquid phase, the liquid phase is represented by more than one field variable, and is specifically represented by three fields, with the vapor phase as a fourth field, consisting of droplets, a liquid base film, and disturbance waves. The method comprises:

simulating steady-state or transient boundary conditions, such as inlet coolant water flow into said sub-channels, the coolant water flow may have a predetermined flow velocity variation,

analyzing predefined parameters of said disturbance waves and base film, including wave velocity, wave frequency and base film thickness, and

analyzing liquid base film thickness between consecutive passing disturbance waves, to calculate local instantaneous impurity concentrations based on said simulated boundary conditions, the calculation is made for each fuel rod of the fuel assembly, wherein, for each fuel rod, the method further comprises comparing said calculated local instantaneous impurity concentration to a crud compound precipitation limit, and during the time said concentration is higher than said precipitation limit, crud is considered to have occurred. In a related simulation method also base film dryout, clad temperature increase, and drop entrainment from waves, may be determined.

A method of controlling a plasma in a nuclear fusion reactor. The nuclear fusion reactor comprises sensors and plasma control inputs. An initial control model is provided, relating readings of at least a subset of the sensors to control of the plasma control inputs. A control loop is performed, comprising: operating the plasma control inputs in dependence upon the sensors according to the control model; determining correlations between readings of each of the sensors, and/or between readings of the sensors and states of the plasma control inputs; and adjusting the control model based on the determined correlations.

A verification method for an application logic provided with one or more macro logics configured to perform a predetermined operation, a macro operation control unit configured to instruct the one or more macro logics to start the operation to cause the one or more macro logics to perform the operation, and an operation data storage area configured to store data. In the application logic, static verification by property description of a formal verification language is performed for each of the one or more macro logic, the macro operation control unit, and the operation data storage area, and dynamic verification by simulation is further performed for at least one of the one or more macro logics.

A core performance calculation apparatus includes: a nuclear constant storage device that stores nuclear constants that have been evaluated in advance in analysis of a fuel assembly; and a three-dimensional core nuclear thermal-hydraulic characteristics analysis device that obtains core characteristics including a power of the fuel assembly. The nuclear constant storage device stores, as the nuclear constants, response relationships between a neutron that flows into a fuel assembly cell and fuel assembly nuclear characteristics, and response relationships between a neutron that is produced from a fuel rod and the fuel assembly nuclear characteristics. The three-dimensional core nuclear thermal-hydraulic characteristics analysis device obtains a neutron effective multiplication factor by using the response relationships that have been stored in the nuclear constant storage device, and obtains the power of the fuel assembly by using the neutron effective multiplication factor.

A core monitoring system including: a TIP measuring a neutron amount in a nuclear reactor; a TIP drive device; a TIP panel; a neutron monitoring panel; and a process computer. The TIP panel includes: a TIP level processor and a TIP position processor that process a TIP level signal and a TIP position signal input from the TIP drive device, respectively; a time setting section synchronizing the TIP level signal and the TIP position signal; and a TIP level data storage section storing synchronized TIP level data. The neutron monitoring panel includes a time setting section setting collecting time of a LPRM level signal and an APRM level signal. The process computer compares the time and stores the TIP level data from the TIP panel and the LPRM and APRM level signals from the neutron monitoring panel corresponding in time, and calculates core performance based on the TIP level data.