A method of controlling a nuclear power plant includes obtaining sensor data from one or more sensors of the nuclear power plant, providing the sensor data and a desired plant response to a neural network, wherein the neural network has been previously trained using a simulated nuclear power plant and is structured to determine at least one control system setting to achieve the desired plant response, determining at least one control system setting to achieve the desired plant response with the neural network, and setting or changing at least one control system setting of a control system of the nuclear power plant to the at least one control system setting determined by the neural network.

A sensor assembly for determining an operating characteristic of a nuclear reactor. The sensor assembly includes a solid-state lasing media doped with a fissile species and disposable within a core of the nuclear reactor and an optical fiber operably coupled to the solid-state lasing media and configured to extend out of the core of the nuclear reactor and to control system of reactor. The fissile species include one or more of uranium, plutonium, americium, or californium. A method of determining an operating characteristic of a nuclear reactor includes during operation of the nuclear reactor; receiving from the optical fiber a laser light, analyzing the laser light, and based on the analysis of the laser light, determining the operating characteristic of the nuclear reactor.

A method is for controlling a nuclear power plant comprising pressurized water nuclear reactor (3) having a reactor core producing power, a primary circuit (5) connecting the reactor core to a steam generator (9), one or more of control rods (16), which can be moved into the reactor core for controlling the power of the reactor core, an injecting device (22, 23, 24, 26, 28, 30) for injecting boric acid and/or deionized water into the primary circuit (5) for controlling the reactivity of the reactor core.

A reactor control interface includes a home screen video display unit (VDU) displaying blocks representing functional components of a nuclear power plant and connecting arrows that connect blocks that are providing the current heat sinking path for the nuclear power plant. Directions of the connecting arrows represent the direction of heat flow along the current heat sinking path. If the current heat flow path of the plant changes, the connecting arrows are updated accordingly. Additional VDUs include: a mimic VDU displaying a mimic of a plant component; a procedures VDU displaying a stored procedure executable by the plant; a multi-trend VDU trending various plant data; and an alarms VDU displaying side-by-side alarms registries sorted by time and priority respectively. If a VDU fails, the displays are shifted to free up one VDU to present the display of the failed VDU, and one display is shifted to an additional VDU.

A reactor control interface includes a home screen video display unit (VDU) displaying blocks representing functional components of a nuclear power plant and connecting arrows that connect blocks that are providing the current heat sinking path for the nuclear power plant. Directions of the connecting arrows represent the direction of heat flow along the current heat sinking path. If the current heat flow path of the plant changes, the connecting arrows are updated accordingly. Additional VDUs include: a mimic VDU displaying a mimic of a plant component; a procedures VDU displaying a stored procedure executable by the plant; a multi-trend VDU trending various plant data; and an alarms VDU displaying side-by-side alarms registries sorted by time and priority respectively. If a VDU fails, the displays are shifted to free up one VDU to present the display of the failed VDU, and one display is shifted to an additional VDU.

A self-powered in-core detector arrangement for measuring flux in a nuclear reactor core includes a first in-core detector and a second in-core detector. The first in-core detector includes a first flux detecting material, a first lead wire extending longitudinally from a first axial end of the first flux detecting material, a first insulating material surrounding outer diameters of the first flux detecting material and the first lead wire and a first sheath surrounding the first insulating material. The first sheath includes a first section surrounding the first flux detecting material and a second section surrounding the first lead wire. The first section of the first sheath has a greater outer diameter than the second section of the first sheath. The second in-core detector includes a second flux detecting material, a second lead wire extending longitudinally from a first axial end of the second flux detecting material, a second insulating material surrounding outer diameters of the second flux detecting material and the second lead wire, and a second sheath surrounding the second insulating material. The second sheath includes a first section surrounding the second flux detecting material and a second section surrounding the second lead wire. The first section of the second sheath has a greater outer diameter than the second section of the second sheath. The first section of the first sheath is axially offset from the first section of the second sheath and radially aligned with the second section of second sheath.

A high-precision high-fidelity real-time simulation and behavior prediction method and device for a nuclear power station is provided. The method comprises the following steps: (1) constructing a nuclear power station simulator and a physical nuclear power station based on the same design parameters; (2) operating the nuclear power station simulator and the physical nuclear power station in parallel, and obtaining predicted parameters output by the nuclear power station simulator and operation parameters of the physical nuclear power station in real time; (3) comparing the predicted parameters and the operation parameters representing the same physical quantity one by one, and correcting prediction models in the nuclear power station simulator and input parameters of the prediction models by adopting a large-scale concurrent-parallel parameter search and correction algorithm and an artificial intelligence-based mode recognition and correction algorithm until the predicted parameters reach specified precision; and (4) operating the nuclear power station simulator according to a set operation condition to obtain the predicted parameters, thereby completing a behavior prediction of a physical nuclear power station system.

A dynamic characteristic analysis method of DET and RELAP5 coupling based on a universal instrumental variable method includes steps of: constructing a DET simulation model of a discrete dynamic event tree and modifying TRIP cards of an input file by adding universal instrumental TRIP variables according to state transition types of DET simulation objects, the universal instrumental TRIP variable being variable type or logical type; setting a simulation time of the RELAP5, controlling a simulation step, and analyzing an output result file of each simulation step of the RELAP5; backtracking the RELAP5 according to state transition types of DET simulation objects. The dynamic characteristic analysis method has advantages of simplifying TRIP setting process and method of DET state transition objects in an input file of the RELAP5 required for the coupling of DET and RELAP5, reducing a modeling complexity and improving a modeling efficiency.

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 of determination of a nuclear core loading pattern defining the disposition of fuel assemblies includes definition of at least one potential core loading pattern, calculation of predictive bowing of the fuel assemblies at the end of the operation cycle for each potential core loading pattern, the calculation being carried out by an automatic learning algorithm trained on a training data set comprising a plurality of others loading patterns and, for each of them, the measures of bowing of fuel assemblies at the end of cycle, evaluation of the at least one core loading pattern based on the predictive bowing calculations and at least one predetermined criteria, and selection of one of the potential core loading patterns.