Provided is a method of the simulation construction for measurement of the control rod insertion time including a three-dimensional modeling operation of an inside wall of the nuclear reactor, a control rod, etc; a flow field configuration operation wherein the flow field is differentially configured by a variable grid system comprising variable cells which change the configuration and by an aligned grid system comprising fixed cells which maintains the configuration; a calculation operation of simulation estimated value for the insertion time by analyzing the thermal-hydraulic phenomenon using the three-dimensional CFD; and a cell change operation, wherein an error between the estimated value and the actual value is verified whether the error lies within the reference range, and, when the error exceeds the reference range, the size of the variable cell and/or of the size of the fixed cell is changed.

A nuclear fuel simulation method is provided that includes: step (a) for receiving an input of data on the order in which nuclear fuels are moved; step (b) for extracting, from the data, information on nuclear fuels, the coordinates of locations from which nuclear fuels are unloaded, and the coordinates of locations into which nuclear fuels are loaded; and step (c) for simulating the information extracted in step (b) according to a flowchart of the data. The present invention has an advantage in that it is possible to accurately and quickly verify all fuel movement works requiring the unloading and loading of nuclear fuels by receiving an input of a huge amount of data on the order in which nuclear fuels are moved and systematically verify an error that may occur during a simulation according to a flowchart, which enables the workload of about three man-days, required per cycle for each reactor, to be done in three man-hours, thereby achieving a significant reduction in working time.

A machine-learning tool learns from sensors' data of a nuclear reactor at steady state and maps them to controls of the nuclear reactor. The tool learns all given ranges of normal operation and responses for corrective measures. The tool may train another learning tool (or the same tool) that forecasts the behavior of the reactor based on real-time changes (e.g., every 10 seconds). The tool implements an optimization technique for differing half-life materials to be placed in the reactor. The tool maximizes isotope production based on optimal controls of the reactor.

The present invention relates to a nuclear cross section Doppler broadening method and device. The method includes: discretizing a product F(x,?) of an average reaction cross section function ?(E,T) and an energy E on grids equally divided on a square roll N of the energy as F.sub.k.sup.c(?), where incident particles have mass m and energy E, target particles have mass M and Maxwellian energy distribution under a temperature T, and E(x,?)=E?(E,T), F.sub.k.sup.c(?)=F(x.sub.k,?), k=0,1, . . . N?1, x=?{square root over (E)}, and c are discrete superscript symbols; expanding the product F(x,?) of the average reaction cross section function and the energy on a group of orthogonal function sets, an expansion coefficient is {circumflex over (f)}.sub.j(?), and j is an index of the orthogonal function sets, where for the discretized product F.sub.k.sup.c(?) of the average reaction cross section function and the energy, an orthogonal function expansion coefficient thereof is {circumflex over (f)}.sub.j.sup.c(?)?{circumflex over (f)}.sub.j(?), based on the product F(x,0) of the average reaction cross section function and the energy under a 0 K temperature, obtaining a group of coefficient weights {circumflex over (f)}.sub.j.sup.c(0), where {circumflex over (f)}.sub.j.sup.c(?) is a function of {circumflex over (f)}.sub.j.sup.c(0); and representing F(x,?) as a sum of an orthogonal function of the group of coefficient weights, using the group of coefficient weights {circumflex over (f)}.sub.j.sup.c(?), calculating F(x,?), and obtaining an average reaction cross section ?(E,T).

Described herein are methods for analyzing an operating envelope of a nuclear reactor. An example method includes obtaining operating envelope parameters associated with a first reactor core by performing a plurality of thermo-hydraulic and thermo-mechanical calculations for the first reactor core, where the first reactor core includes a first fuel-cladding material and has a first fuel pin geometry; obtaining operating envelope parameters associated with a second reactor core by performing a plurality of thermo-hydraulic and thermo-mechanical calculations for the second reactor core, where the second reactor core includes a second fuel-cladding material and has a second fuel pin geometry; and assessing an expandable operating envelope by comparing the respective operating envelope parameters associated with the first reactor core and the second reactor core, where the first and second fuel-cladding materials are different materials. The example method can include iteratively performing the steps described herein.

A plant operation assistance system includes: a data obtaining unit-configured to obtain monitoring data indicating state quantity of a plant, the state quantity being detected by a sensor; an identifying unit configured to identify, based on the state quantity, a probability distribution of the monitoring data; a model generation unit configured to generate, based on a plant parameter composed from a database including design information of the plant, a stochastic model of the plant; a data processing unit configured to assign the probability distribution to the monitoring data obtained by the data obtaining unit; and a prediction unit configured to input the monitoring data assigned with the probability distribution, into the stochastic model, and predicts a state of the plant.

According to an embodiment, a design method for a light-water reactor fuel assembly comprises: accumulating a determined fuel data, showing that each of a combination of p.Math.n/N and e is feasible as the core or not, wherein N is a number of the fuel rods in the fuel assembly, n is a number of the fuel rods containing the burnable poison, p is a ratio wt % of the burnable poison in the fuel, and e is an enrichment wt % of the uranium 235 contained in the fuel assembly; formulating a criterion formula which determines whether a combination of p.Math.n/N and e is feasible as a core or not and is formulated based on the determined fuel data; and determining whether a temporarily set composition of the fuel assembly is approved as a core or not based on the criterion formula.

A system for simulating maintenance of a reactor core protection system that has at least two or more channels, includes: a simulation signal generation unit for generating a simulation state signal including a normal state or an abnormal state, a communication unit connected to each of the channels of the reactor core protection system to transmit the state signal to the channel, and a control unit for receiving a result signal output from the channel in response to the input simulation state signal and confirming whether the reactor core protection system normally determines a reactor core state by analyzing the result signal.

A computer-assisted method for determining an optimal core-loading pattern for a nuclear reactor core. Positions of nuclear fuel assemblies are tested to assign optimal positions and to load the reactor. The reactor core includes cells positioned symmetrically to axes of symmetry and a standard assembly for insertion into each cell. Standard assemblies are distributed by the number of previous production cycles. Groups of cell positions symmetric to the axes of symmetry are identified, and symmetric positions are counted. Families of standard assemblies having similar burnups are formed, wherein the standard assemblies correspond to positions in a group. The loading pattern of standard assemblies in initial positions is tested by numerical simulation, then the positions are swapped while maintaining the previously formed families of assemblies. The swapped positions loading pattern is tested by numerical simulation. This is repeated until at least one candidate pattern for loading the reactor is obtained.

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.