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 operating a nuclear reactor is provided. The method includes defining a layer increment of a deposit layer modeling a deposit on a heat transfer surface of the nuclear reactor; periodically updating a thickness of the deposit layer by adding the layer increment to the deposit layer; recalculating properties of the deposit layer after each layer increment is added to the deposit layer; determining a temperature related variable of the heat transfer surface as a function of the recalculated properties of the deposit layer; and altering operation of the nuclear reactor when the temperature related variable of the heat transfer surface reaches a predetermined value. A method of modeling a deposit on a heat transfer surface of a nuclear reactor is also provided.

A method for analyzing a sever accident in a nuclear reactor based on an advanced particle method includes steps of: 1) performing geometric modeling, setting initial conditions and boundary conditions; 2) updating material physical properties and key parameters; 3) performing mechanical structure module calculation, updating solid particle stress, strain, internal energy, displacement and velocity; 4) performing thermal hydraulic module calculation, updating fluid particle internal energy, position and velocity; 5) performing chemical reaction module calculation, updating particle matter composition and internal energy; 6) performing neutron physics module calculation, updating particle neutron flux density; and 7) outputting data. The method of the present invention is based on the discrete form of the advanced particle method, which is capable of accurately capturing cross-sectional changes, matter changes, and phase changes. Compared with grid method, the present invention can effectively avoid a mesh distortion problem existing in a large deformation.

The invention relates to methods for protecting a nuclear reactor core, such as a boiling water reactor core, from fuel and cladding damage due to thermal hydraulic instability in extended operating power flow conditions and, in particular, when an extended power uprate is implemented. The methods employ existing licensed stability methodologies and incorporated minor changes, e.g., to the Average Power Range Monitor (APRM)-based trip system to preclude operation inside the stability vulnerable region of the power/flow map. The APRM-based trip system is modified to set down the APRM flow-biased scram line when core flow is less than a predetermined core flow to prevent the core from entering an unstable region of operation.

Illustrative embodiments provide for the operation and simulation of the operation of fission reactors, including the movement of materials within reactors. Illustrative embodiments and aspects include, without limitation, nuclear fission reactors and reactor modules, including modular nuclear fission reactors and reactor modules, nuclear fission deflagration wave reactors and reactor modules, modular nuclear fission deflagration wave reactors and modules, methods of operating nuclear reactors and modules including the aforementioned, methods of simulating operating nuclear reactors and modules including the aforementioned, and the like.

The invention relates to methods for protecting a nuclear reactor core, such as a boiling water reactor core, from fuel and cladding damage due to thermal hydraulic instability in extended operating power flow conditions and, in particular, when an extended power uprate is implemented. The methods employ existing licensed stability methodologies and incorporated minor changes, e.g., to the Average Power Range Monitor (APRM)-based trip system to preclude operation inside the stability vulnerable region of the power/flow map. The APRM-based trip system is modified to set down the APRM flow-biased scram line when core flow is less than a predetermined core flow to prevent the core from entering an unstable region of operation.

A method of corrosion rate control of nuclear power plants process circuits equipment. The electrochemical potential of the structural material of heat exchanging tubes and the specific electrical conductivity of blowdown water in steam generators are measured, the polarization resistance of the structural material of the pipelines of the condensate-feeding path and the specific electrical conductivity of feed water in steam generators are measured, and these parameters are automatically averaged and compared with the rated values, which determine various degrees of corrosion activity in relation to the material of pipelines of the feed water circuit in steam generators. Depending on the data comparison, no actions are taken, coolant parameters are adjusted, or the power unit is shut down.

Various examples are related to prediction of neutronics parameters using deep learning. In one embodiment, a method includes generating a training data set based upon one or more principled approaches that provide a gradient of values; generating a neural network using structured or unstructured sampling of a hyperparameter space augmented by probabilistic machine learning; training the generated neural network based on the training data set to produce one or more neutronics parameters; and generating at least one neutronics parameter utilizing the trained neural network.

This invention relates to the monitoring and diagnosing of nuclear power plants for its thermal performance using the NCV Method. Its applicability comprises any nuclear reactor such as used for research, gas-cooled and liquid metal cooled systems, fast neutron systems, and the like; all producing a useful output. Its greatest applicability lies with conventional Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plants generating an electric power. Its teachings of treating fission as an inertial process, a phenomena which is self-contained following incident neutron capture, allows the determination of an absolute neutron flux. This process is best treated by Second Law principles producing a total fission exergy. This invention also applies to the design of a fusion thermal system regards the determination of its Second Law viability and absolute plasma flux.

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.