Calculation Of Power Distribution Research Articles

Effect of Gamma Heating on Heat Pipe–Cooled, Calcium Hydride–Moderated Core

A novel microreactor, called MoveluXTM, was previously proposed that utilizes heat pipes as the primary heat transfer device and calcium hydride as the moderator. In this core design, the moderator temperature is the critical core operation limit because at high temperatures above 800°C, the hydrogen dissociates from the calcium hydride. The core temperature distribution, therefore, was previously evaluated. However, this evaluation did not consider gamma heating in the core and assumed that power was produced only in the fuel region. By contrast, the moderator region has a power density under realistic conditions due to gamma heating. Thus, the present work considers gamma heating in the core power distribution calculation and evaluates the impact on the moderator temperature. The power density of gamma heating was 1/10th that of the fuel region and around 1/100th that of the core thermal power. This increased the temperature of the moderator by 10 K from the case without considering gamma heating. In addition, this temperature distribution difference did not have an impact on the core criticality. In conclusion, considering the gamma heating, concerns regarding the core design are not suggested.

Модель реактора ВВЕР-1000 у прецизійному коді SERPENT 2 для розрахунку розподілу енерговиділення в активній зоні

A VVER-1000 reactor model using the Monte Carlo Serpent 2 code for core power distribution calculation is presented. The core and zones located near to the core were modeled in detail, without simplification. The assembly power distribution and axial power profiles were calculated for fresh core of the X2 VVER-1000 benchmark, namely the core of the KhNPP2 first loading for the hot zero power. The results were compared with the data obtained by specialists from Helmholtz-Zentrum Dresden-Rossendorf.

Multi-physics coupling analysis of test heat pipe reactor KRUSTY based on MOOSE framework

The heat pipe reactor utilizes heat pipes for passive heat transfer, eliminating the need for any moving parts. This feature significantly enhances its safety characteristics. Given the intricate nature of the heat pipe reactor system, a comprehensive multi-physics coupling simulation is necessary to study its safety features. In this study, a high-fidelity multi-physics coupling analysis incorporating a neutronic-thermal–mechanical-heat pipe coupling model is conducted. The work is based on the multi-physics coupling framework MOOSE, which integrates the Monte Carlo code OpenMC for neutronic analysis and built-in modules for 3-dimensional thermal–mechanical analysis. Furthermore, a heat pipe simulation method called the Thermal Resistance Grid Method is implemented within the MOOSE framework and employed in the multi-physics coupling analysis. To facilitate transient analysis, the Point Kinetics method is also integrated into the MOOSE framework to account for the time-varying power factor during reactor transients. The various modules and physical fields are coupled together using either loose or tight coupling strategies. These coupling strategies are applied to assess a test heat pipe reactor known as KRUSTY. Steady-state simulations are conducted for both normal operation and single heat pipe failure conditions, enabling the calculation of fission power distribution, temperature distribution, stress and strain profiles, as well as power and temperature redistribution following a heat pipe failure. Moreover, a load-following transient of KRUSTY is simulated, demonstrating its capability for load following without reactivity control. The aforementioned calculations validate the feasibility of the coupling framework utilized in this study and highlight its ability to accurately capture local phenomena within the reactor core.

MGGC2.0: A preprocessing code for the multi-group cross section of the fast reactor with ultrafine group library

How to generate the precise broad group cross section is important for the fast reactor design. In this study, a fast reactor multi-group cross-section generation code MGGC2.0 are developed in-house for processing ultrafine group MATXS format library. Validation and verification are performed for MGGC2.0 code by applying the benchmarks of ICSBEP handbook, and the results of MGGC2.0 agree well with that of MCNP. The consistent PN method with critical buckling search is in good agreement that condensed with TWODANT flux and flux moment for the inner core and outer core region. For the radial blanket and reflector, two region approximation method has been applied in MGGC2.0 by using collision Probability Method neutron flux solver. The RBEC-M benchmark was used to verify the power distribution calculation, and the relative error of power distribution comparison with the reference are less than 0.8% in the fuel region and the maximum relative error is 5.58% in the reflector region. Therefore, the precise broad cross section can be generated by MGGC2.0 for fast reactor.

Polarization-selective narrow band dual-toroidal-dipole resonances in a symmetry-broken dielectric tetramer metamaterial.

Here we propose a metasurface consisting of symmetry-broken dielectric tetramer arrays, which can generate polarization-selective dual-band toroidal dipole resonances (TDR) with ultra-narrow linewidth in the near-infrared region. We found, by breaking the C4v symmetry of the tetramer arrays, two narrow-band TDRs can be created with the linewidth reaching ∼ 1.5 nm. Multipolar decomposition of scattering power and electromagnetic field distribution calculations confirm the nature of TDRs. A 100% modulation depth in light absorption and selective field confinement has been demonstrated theoretically by simply changing the polarization orientation of the exciting light. Intriguingly, it is also found that absorption responses of TDRs on polarization angle follow the equation of Malus' law in this metasurface. Furthermore, the dual-band toroidal resonances are proposed to sense the birefringence of an anisotropic medium. Such polarization-tunable dual toroidal dipole resonances with ultra-narrow bandwidth offered by this structure may find potential applications in optical switching, storage, polarization detection, and light emitting devices.

Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code

Power distribution plays a significant role in preventing the fuel temperature from exceeding the safety limit of 1600 °C in high-temperature gas-cooled reactors. Experiments for measuring the power distribution in the graphite-moderated system were performed at the Very High-Temperature Reactor Critical Assembly (VHTRC) facility. The power distribution was determined from the measured Cu activation rate for both the radial and axial distributions. In this study, the pin-wise power distribution of the VHTRC was evaluated with the Monte Carlo MVP3 code. The difference between the calculated and measured results was less than 1 % for the axial and radial distributions. The significant results were concerned with the area around the fuel and reflector regions in the axial direction, where the average discrepancy between the calculated and measured values was 0.8 %. This result showed improved agreement compared to the diffusion calculation that was conducted in the previous study.

Neutronics for the GEMINI+ HTGR

Literally at the heart of the Euratom Horizon 2020 project GEMINI+ are the core neutronics (design) calculations. For these calculations on a relatively small (180 MWth) prismatic HTGR with cylindrical core, the 3-D monte-carlo particle transport and depletion code SERPENT version 2 (VTT, Finland) was selected, the main reasons being the flexibility and versatility of this code. This enables the modelling of all relevant details of the reactor without unnecessary approximations. A particularly useful feature of the SERPENT code is the multiphysics input capability. This allows to map a temperature field over the defined geometry, enabling the calculation of converged power and temperature distribution by means of iteration and data exchange between SERPENT and a (steady-state) thermal- hydraulics code. In this particular case the SPECTRA code (NRG, The Netherlands) was used to provide the temperature distribution. 4 to 5 iterations are sufficient to reach simultaneously converged distributions for power and temperature. The paper gives an overview of the performed analyses for the current (June 2020) design of the GEMINI+ HTGR, and results thereof. Neutronics features seem quite promising, but further improvements and therefore further investigations would be desirable.

Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT

To solve the problem that the thermal-neutronic grid mapping relation is complicated and cannot be preset in a centralized manner due to the existence of irregular geometry in performing the thermal-neutronic coupled simulation calculation for various small power reactors and experimental reactors, this paper studies the thermal-neutronic coupling method based on the unified geometric modeling, using the high-fidelity numerical code for reactor neutronics calculation, NECP-X. This study establishes the mapping relation for the thermal-neutronic coupling on the basis of the neutronics model, and enables the direct transient calculation of the experimental reactor core for the special power excursion reactor test (SPERT) via combination with the transient calculation method in NECP-X. Then, this study calculates the steady-state case for the experimental reactors of SPERT, and compares the calculation results with the results gained from the Monte Carlo code. On this basis, this study conducts transient calculation and analysis for these experimental reactors and compares the corresponding results with the experimental results. The final results show that the eigenvalues from the neutronics calculation by the NECP-X and the rod power distribution calculation results are of high accuracy; that the grid mapping method based on the unified geometric modeling allows a easy and fast thermal-neutronic coupled calculation of the PWRs of complex geometry; and that compared with the experimental values, the curve of change in the total power and reactivity gained from transient calculation with time is more accurate and can provide refined distributions of power and temperature.

Neutronics and both analytical and numerical solutions for the rod centered subchannel thermal-hydraulic model

Abstract This work investigates the capacity to detect safety levels in pressurized water reactors using neutronic and thermal-hydraulic calculations (PWR). The main neutronic parameters, such as criticality calculations, reactivity calculations, burn-up calculations, and power distribution calculations, have been calculated by using MCNPX based on the Monte Carlo method. The hot channel was determined from the power distribution. On the hot channel, the rod-centered sub-channel model was used to investigate the main thermal-hydraulic parameters such as the fuel, cladding, and coolant temperature distribution in both axial and radial directions, the coolant pressure drop, and the departure from nucleate boiling ratio (DNBR). Two techniques were used to solve the rod-centered sub-channel model. MATLAB code was used for the analytical methodology, and COMSOL-Multiphysics computer software was used for the numerical methodology. The numerical technique has demonstrated high accuracy in determining the reactor safety level consistent with the analytical approach.

AN IMPROVED ENERGY DEPOSITION MODEL IN MPACT AND EXPLICIT HEAT GENERATION COUPLING WITH CTF

The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation of delayed neutron precursors. To allow the explicit heat generation coupling, the interfaces between MPACT and CTF were updated to transfer separate heat sources for different material regions (fuel, clad, moderator and guide tube). The results show that the distributions of the energy deposition between MPACT and MCNP agree very well for various 2-D assembly and quarter-core problems without TH feedback. The MPACT/CTF coupled calculation for the hot full power quarter-core case exhibited a reduced peak pin power by 2.3% and a reduced peak fuel centerline temperature by 17 K when using the explicit energy deposition and heat transfer. The new model also shows a maximum 100 pcm keff effect on assembly depletion problems and an increased overall energy release by 7% in a PWR reactivity-initiated accident (RIA) problem.

Study of QoS-aware reliability transmission methods for edge computing networks in power distribution IoT

As the construction of IoT for power distribution advances, millions of power distribution equipment, electrical quantity sensors and condition sensors will be connected to the IoT network, thus generating huge amounts of heterogeneous power distribution data, whose collection, transmission and calculation will put great pressure on the communication channel and the master station storage and calculation system. To solve the above-mentioned distribution network problems, edge computing technology is introduced and an edge computing platform is deployed on the end-side to achieve lightweight data processing. The contradiction between the huge amount of data generated by IoT terminals and their QoS requirements, and the limited system resources, makes it imperative to study QoS-awareness transmission methods for edge computing networks.

Neutronics benchmark analysis of the EBR-II SHRT-45R with SAC-3D

Neutronics calculation model plays a very important role in applying a specific simulation of a reactor system. The accuracy of the reactor core neutronics calculation results can hugely affect the final results of the reactor system analysis. At present, system analysis codes applied two kinds of methods to conduct the neutronics calculation. In normal conditions, the point kinetics model can get a satisfying result. But in the large scale reactor core or in several accident conditions in which the reactor core power distribution varies greatly in space, the point kinetics model may have a result with a large error caused by ignoring the spacing effect. To solve this problem, a system code with a 3D neutronics calculation model inside is required. SAC is a system analysis code with different versions which is developed by North China Electric Power University. In the SAC-3D version, we apply the hexagonal nodal expansion method for the diffusion equation to conduct the core flux (power) distribution calculation. In this paper, we prepared the calculation model of the EBR-II SHRT-45R reactor core with the neutronics calculation module integrated with SAC-3D and do the neutronics benchmark analysis. By comparing with the benchmark data and other project participants’ results, the good agreement demonstrated the accuracy and capability of SAC-3D to simulate the principle physics phenomenon for the liquid-cooled fast reactor.

Initial Criticality Analysis of Malaysia’s TRIGA Research Reactor Using TRIGLAV Computer Code

Malaysian Nuclear Agency hosts the 1MW Thermal TRIGA MARK II research reactor since 1982. It first initial criticality was achieved on 28th June 1982 with loading of solid fuel elements like Uranium Zirconium Hydride. TRIGA MARK II started its operation of the research reactor on the same year with a 1MW power generation. Training, Research, Isotope Production and General Atomic (TRIGA) is designed to successfully actualize the variety fields of fundamental nuclear research, the manpower training and the production of radioisotopes. This study deals with the initial criticality analysis of the TRIGA research reactor using TRIGLAV reactor physics computer program. For this purpose, a model of its initial core will be developed and simulated using the software and the results will be validated against the experimental result as mentioned in the final safety analysis report (FSAR). The TRIGLAV computer code solves the neutron diffusion equation by using a finite differences method with iteration of fission density. TRIGLAV is based on four group time independent diffusion equation in two dimensional cylindrical (r, θ) geometry. TRIGLAV can also be applied for criticality of reactor, fuel burn-up calculations, and power distribution and flux distributions calculations of the core and also the reactivity predictions of the reactor.

Research on steady-state power distribution calculation technology of electrothermal coupled regional energy system

When large-scale power grids are connected to the power grid, it is conceivable that adverse effects on the operation and planning of power grids can be imagined. Therefore, wind power has characteristics such as intermittency, volatility, anti-peaking characteristics, and large prediction errors. At the same time, the ability to absorb wind power is also weakened by the contradiction between the peak load and the heating load. In addition, due to the thermo-electric coupling characteristics of the cogeneration unit, ie, “heating with electricity,” this characteristic directly determines that its heat output severely limits the regulation of the electric output. The results of the study indicate that the co-production unit’s electric power and thermal power are coupled. Strong nature, the peak load capacity of the unit is greatly affected by the heat load, and the electric power and thermal power distribution lacks a unified calculation and analysis method, which restricts the means of improving the peaking capacity of the unit by controlling the thermal power output of the unit, and through implementation, establishes the main power equipment, The steady-state power model of the thermal equipment, the steady-state power analysis and calculation method of the regional energy system of the power thermocouple, and the realization of adjusting the thermal load related parameters to improve the peak capacity of the thermal power unit.

A novel multi-objective optimization method, imperialist competitive algorithm, for fuel loading pattern of nuclear reactors

Imperialist Competitive Algorithm (ICA) has been successfully applied to the various optimization problems and has demonstrated great results in both the global optimal achievement and the convergence rate. In this work, the ICA has been applied to achieve the best fuel arrangement of the VVER-1000 reactor core during the cycle. Bushehr Nuclear Power Plant (BNPP) core optimal arrangement has been searched by considering the minimization of the power peaking factor, flattening of the radial neutron flux distribution and maximizing the effective multiplication factor (Keff) of the core. Cross section and power distribution calculations have been performed by DRAGON-4 and CITATION codes, respectively. The algorithm implementation has been carried out in MATLAB framework for optimization of one-sixth (1/6) and one-twelfth (1/12) symmetry of the core. The results show that this method can be used as an efficient computational method for finding an optimized multi-objective loading pattern in a reactor core.

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

High Fidelity Multi-Objective Design Optimization of a Downscaled Cusped Field Thruster

The Cusped Field Thruster (CFT) concept has demonstrated significantly improved performance over the Hall Effect Thruster and the Gridded Ion Thruster; however, little is understood about the complexities of the interactions and interdependencies of the geometrical, magnetic and ion beam properties of the thruster. This study applies an advanced design methodology combining a modified power distribution calculation and evolutionary algorithms assisted by surrogate modeling to a multi-objective design optimization for the performance optimization and characterization of the CFT. Optimization is performed for maximization of performance defined by five design parameters (i.e., anode voltage, anode current, mass flow rate, and magnet radii), simultaneously aiming to maximize three objectives; that is, thrust, efficiency and specific impulse. Statistical methods based on global sensitivity analysis are employed to assess the optimization results in conjunction with surrogate models to identify key design factors with respect to the three design objectives and additional performance measures. The research indicates that the anode current and the Outer Magnet Radius have the greatest effect on the performance parameters. An optimal value for the anode current is determined, and a trend towards maximizing anode potential and mass flow rate is observed.

NEMHEX solution of benchmark problems on thermal and fast breeder reactor cores with hexagonal lattice

This paper gives the solution of some benchmark problems for reactor core calculations in hexagonal geometry using a computer code NEMHEX (Nodal Expansion Method for HEXagonal geometry). The problems cover a wide range of thermal reactor cores including large heavy water, fast breeder and VVER reactors in both two and three dimensional. The code is based on higher order nodal expansion method (NEM) for the solution of the neutron diffusion equation in 3-D hexagonal geometry. In this method, flux in each mesh/node and energy group is expanded in terms of polynomial up to third order in X–Y plane and fourth order in axial direction (Z). Weighted residual method is used to calculate the higher order expansion coefficients of the polynomials. Both the partially integrated radial and axial leakages are approximated by the quadratic polynomial. Higher order polynomial expansion [with respect to the work by Lawrence (DIF3D nodal neutronics option for two-and three-dimensional diffusion theory calculations in hexagonal geometry, 1983)] of flux as well as leakage in each node improves the accuracy in power distribution of VVER-1000 type reactor (which is having large size lattice) in terms of maximum reduction in error from 8 to 2%. The final inhomogeneous response matrix equation involves spatial moments of nodal flux and surface averaged partial currents across the surfaces of node. These equations are solved using conventional fission source iteration. This code is used for calculation of effective multiplication factor, neutron flux and fission power distribution. The results of the analysis establish the adequacy of NEMHEX for thermal as well as fast reactor core calculations in hexagonal geometry.

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

Test case for VVER-1000 complex modeling using MCU and ATHLET

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.