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.

There is provided a method of tracking a characteristic line, a method and apparatus of calculating core neutron physics, and a computer-readable storage medium. The method of tracking the characteristic line includes: determining, sequentially from the uppermost level to the lowest level, whether an intersection of the sub-model meeting a first preset condition in each of the plurality of levels and the characteristic line segment meets a second preset condition or not; and obtaining an intersection, that meets the second preset condition, of the sub-model meeting the first preset condition in the lowest level and the characteristic line, so as to obtain a segment correspondence between the characteristic line and the material of the core according to the intersection obtained.

An exposure dose measurement simulation device according to an embodiment of the present invention includes: a virtual work area generating portion that generates a virtual work area corresponding to a work area of a radiation facility; and an exposure dose prediction portion that predicts an exposure dose of a worker positioned in the virtual work area, wherein the exposure dose prediction portion includes a database portion that includes first data on a spatial dose rate of the radiation facility and second data on a position of a radiation source in the radiation facility, and a first exposure dose calculation portion that calculates the exposure dose of the worker positioned in the virtual work area using the first data and second data.

The present relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into an integral molten salt nuclear reactor (IMSR). Once the design life of the IMSR is reached, for example, in the range of 3 to 10 years, it is disconnected, removed and replaced as a unit. The spent IMSR functions as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt contained in the vessel of the IMSR. The present also relates to a nuclear reactor that has a buffer salt surrounding the nuclear vessel. During normal operation of the nuclear reactor, the nuclear reactor operates at a temperature that is lower than the melting point of the buffer salt and the buffer salt acts as a thermal insulator. Upon loss of external cooling, the temperature of the nuclear reactor increases and melts the buffer salt, which can then transfer heat from the nuclear core to a cooled containment vessel.

The present relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into an integral molten salt nuclear reactor (IMSR). Once the design life of the IMSR is reached, for example, in the range of 3 to 10 years, it is disconnected, removed and replaced as a unit. The spent IMSR functions as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt contained in the vessel of the IMSR. The present also relates to a nuclear reactor that has a buffer salt surrounding the nuclear vessel. During normal operation of the nuclear reactor, the nuclear reactor operates at a temperature that is lower than the melting point of the buffer salt and the buffer salt acts as a thermal insulator. Upon loss of external cooling, the temperature of the nuclear reactor increases and melts the buffer salt, which can then transfer heat from the nuclear core to a cooled containment vessel.

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.

The present relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into an integral molten salt nuclear reactor (IMSR). Once the design life of the IMSR is reached, for example, in the range of 3 to 10 years, it is disconnected, removed and replaced as a unit. The spent IMSR functions as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt contained in the vessel of the IMSR. The present also relates to a nuclear reactor that has a buffer salt surrounding the nuclear vessel. During normal operation of the nuclear reactor, the nuclear reactor operates at a temperature that is lower than the melting point of the buffer salt and the buffer salt acts as a thermal insulator. Upon loss of external cooling, the temperature of the nuclear reactor increases and melts the buffer salt, which can then transfer heat from the nuclear core to a cooled containment vessel.

The present relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into an integral molten salt nuclear reactor (IMSR). Once the design life of the IMSR is reached, for example, in the range of 3 to 10 years, it is disconnected, removed and replaced as a unit. The spent IMSR functions as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt contained in the vessel of the IMSR. The present also relates to a nuclear reactor that has a buffer salt surrounding the nuclear vessel. During normal operation of the nuclear reactor, the nuclear reactor operates at a temperature that is lower than the melting point of the buffer salt and the buffer salt acts as a thermal insulator. Upon loss of external cooling, the temperature of the nuclear reactor increases and melts the buffer salt, which can then transfer heat from the nuclear core to a cooled containment vessel.

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.