Reactor Kinetics Research Articles

Process characterization in industrial vessels by flow-following sensor particles

Instrumented flow-following sensor particles have been developed for investigation of hydrodynamic and biochemical processes in chemical reactors and bioreactors, where standard measurement techniques are not applicable. The sensor particles allow autonomous long-term measurement of spatially distributed process parameters in the chemically and mechanically harsh environments of e.g. agitated industrial vessels. Each sensor particle comprises of an on-board measurement electronics that logs the signals of the embedded sensors. A buoyancy control unit enables automated taring to achieve neutral buoyancy of the sensor particles. Moreover, controlled floating of the sensor particles is possible to expose them for recovery from the liquid surface. Macro-flow tracking of the sensor particles is validated with circulation time reference measurements by means of salt tracer experiments in a stirred model reactor and computational fluid dynamics (CFD) simulations. Moreover, process characterization with sensor particles is demonstrated in three further applications, namely a biogas pilot digester, an air–water column and a biological wastewater treatment basin. Acquired data were used to fit mixing model parameters, namely effective circulation time, circulation number, degree of suspension and Péclet number.

On a moment analysis of the stochastic nuclear reactor dynamics

On a moment analysis of the stochastic nuclear reactor dynamics

A neural-network based variable universe fuzzy control method for power and axial power distribution control of large pressurized water reactors

This paper proposes a neural network-based variable universe fuzzy controller (NN-VUFC) for power and axial power distribution control of large pressurized water reactors (PWRs). First, a two-node reactor kinetics model is introduced, and the fuzzy inference rules are formulated for the reactor power and axial power distribution control according to their deviations from corresponding setpoint values. Then, based on the idea of variable universe, the scaling factor of the input fuzzy universe is designed to maximize the utilization of fuzzy rules. Finally, a neural network is designed to optimize the proportionality coefficient of the scaling factor online to further improve the fuzzy control performance. Simulation results of the CPR1000 reactor under typical transient conditions show that the NN-VUFC can realize good control of both the reactor power and axial power distribution, and its control performance is effectively improved compared to traditional fuzzy controllers.

HEART, a specific code for thermal-electrical analysis of heat pipe cooled nuclear reactor

Heat pipe cooled reactor (HPR) features high power density, compact structure, long operation life and high reliability, which is an excellent candidate for small nuclear power source for deep space or underwater missions. In this paper, a conceptual Nuclear Silent Thermal-Electrical Reactor (NUSTER) is preliminarily designed, using 109 sodium heat pipes for passive cooling. The thermoelectric generators (TEGs) are employed in the reactor to convert fission heat to electric power. Based on the design, a HEat pipe cooled Advanced Reactor Transient analysis code (HEART) is developed, including point reactor kinetics model, multi-channel model, core heat transfer model, heat pipe model, thermoelectric generator model and coolant model. The models proposed in HEART are preliminarily verified against the design value and experiment data, and the maximum deviation is less than 12.3%. Steady-state performance of the NUSTER are successfully predicted by HEART, and results indicate that the solid-state core has good temperature flattening ability. The surface temperature of the heat pipe is less than 1300 K, and the average temperature drop of the TEG module in central channel is 724.1 K, which can produce an electrical power of 301.94 W. The models developed in HEART can also be applied to other heat pipe cooled reactors, providing valuable experience in design and thermal-hydraulic evaluation of HPRs.

Port-Hamiltonian Control of Nuclear Reactors

The port-Hamiltonian form (PHF) of nuclear reactor dynamics is first given with the corresponding Hamiltonian function determined by the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics. Then, a passivity-based control (PBC) is proposed by adding extra damping injections to the PHF, and it is shown by theoretical analysis that satisfactory closed-loop stability can be guaranteed. The newly designed control law is applied to a nuclear heating reactor, and the simulation results not only verify the theoretical analysis but also show the satisfactory reactor power-level control performance.

Fluidization Dynamics of a New Two-Staged Fuel Reactor for Chemical Looping Gasification

A new staged fuel reactor (FR), which integrates a bubbling bed and a spouted fluidized bed, is proposed in the present work with regard to lowering the pressure drop of the staged FR during chemical looping applications. In order to gain insights into the gas–solid fluid dynamics of the new design, a cold experimental setup was established. The fluidization dynamics of the staged FR was investigated based on the pressure pulsation signal analysis using time domain analysis, power spectral density analysis, wavelet analysis, and the Hurst exponent analysis by the R/S method. There are four flow types of funnel-like flow, spout-fluidization, spouting motion, and unsteady spouting motion to be classified, and the transition boundaries of the corresponding flow types were determined. The design provides a way to lower the overall pressure drop of FR and then reduces the fan energy consumption to hold the oxygen carrier particle circulation.

Multi-objective atom search optimization of biodiesel production from palm empty fruit bunch pyrolysis

In this study, palm empty fruit bunch (PEFB) pyrolysis, bio-oil improvement, and generating electricity were all simulated using Aspen plus. This research employed a kinetic reactor for pyrolysis at 500 °C based on 1,000 tons of PEFB per day. The simulation results indicated that 1 kg/hr. PEFB generated 0.11 kg/hr of char, 0.21 kg/hr of gas, and 0.67 kg/hr of bio-oil, which is in good agreement with literature. The relationship between biodiesel yield, CO2 emissions, and utility costs was then investigated the effect of the distillate-to-feed ratio of biodiesel distillation, heat exchanger temperature, and the flash drum pressure from the process simulation by using central composite design (CCD). The coefficient of determination (R2) values for biodiesel yield, CO2 emissions, and utility costs were 0.9940, 0.9941, and 0.9959, respectively, which was a reason for the excellent model fitting. The optimum response (the biodiesel yield, the CO2 emission, and the utility cost) was obtained at 5,562.73 kg/hr, 33,696.55 kg/hr, and 2,953.99 USD/hr., respectively, with optimum conditions for the distillate-to-feed ratio of 0.899999, temperature of 56.0356 °C and pressure of 18.1479 bar. After that, a quadratic polynomial equation from the RSM was employed as the fitness function to evaluate the fitness value of the multi-objective optimization (MOO) by atom search optimization (ASO) to maximize biodiesel yield and minimize the CO2 emissions and utility costs. The ASO performance was generated into the Pareto optimal solution of 200 generations. The optimal CCD was then compared with the ASO results. It was found that the ASO could reduce CO2 emissions by 1.33% and reduce utility costs by 5.03% while increasing biodiesel yields by 7.01%. It can be observed that the ASO was more efficient at finding parameters than the CCD.

Capturing multiphysics effects in hydride moderated microreactors using MARM

High temperature hydrides are being considered as neutron moderators for LEU microreactor systems due to their ability for creating a compact core with minimum loading of a fissile material. Under the influence of a temperature gradient, hydrogen in metal hydrides is prone to diffusion. This migration of hydrogen during the reactor operation results in local changes in moderator material properties; such as changes in neutron cross sections leading to changes in power and temperature distributions. High fidelity multiphysics modeling is required to properly understand the reactor dynamics during its operation. MCNP, a Monte-Carlo based particle transport code, and Abaqus, a finite-element engineering simulation software, are used to develop a multiphysics framework named MARM for high temperature reactor analysis. MARM is short for MCNP and Abaqus based Reactor Multiphysics software package being developed at the Los Alamos National Laboratory. This suite of codes is aimed to enable multiphysics analysis for full core applications by incorporating neutronics, heat transfer, and mass diffusion solutions. This paper demonstrates the significance of capturing multiphysics effects in a hydride-moderated reactor system. This paper also provides an outline of MARM multiphysics development and its future plans.

Non linear stability analysis of supercritical water cooled reactor: Parallel channels configuration

The present work is devoted to the study of non-linear dynamics of the supercritical water reactor (SCWR) with a parallel channel configuration. To fulfil this objective, point reactor kinetics has been coupled with a parallel channel thermal-hydraulic model. The reduced ordered model is developed by coupling point kinetics equations with one group of delayed neutrons, fuel heat transfer and 1-D fluid flow model. The existing studies are commonly limited to linear stability analysis, therefore, present work focused on to demonstrate the nonlinear dynamics of the SCWR with parallel channel reactor core. The stability boundaries are shown in thermal hydraulic (pseudo phase-change and pseudo subcooling number) and intrinsic reactivity feedbacks (Doppler and density reactivity feedback) parametric spaces. Several bifurcation characteristics associated to density wave oscillations namely, subcritical, supercritical and generalized Hopf bifurcations are observed, which is also confirmed by calculating first Lyapunov coefficients. Corresponding to these phenomena, several numerical simulations are carried out at different parametric values, which confirms the existence and nature of the limit cycle behavior associated with these bifurcations. In addition, parametric sensitivity analysis on the stability characteristics are also carried out.

High accuracy measurement of the prompt neutron decay constant in CROCUS using Gamma Noise and bootstrapped uncertainties

We experimentally demonstrate the advantages of gamma detection for noise measurements to determine the prompt neutron decay constant α of a nuclear reactor using the power spectral density (PSD) method, coupled to a new uncertainty estimation scheme based on bootstrapping. We compare the relative precision of gamma noise measurements and neutron noise measurements performed in the CROCUS research reactor. The reference neutron noise experiment employs two large 235U fission chambers, whereas the new gamma noise experiment is based on two CeBr3 scintillators. The results demonstrate that gamma noise is superior to neutron noise with respect to precision, giving an advantage with regards to measurement time to reach desired uncertainty levels: the uncertainty budget shows a relative uncertainty of 3.6% and 1.3% on the prompt neutron decay constant with neutron and gamma noise, respectively. Moreover, thanks to bootstrapping, the reporting of full distributions – rather than assuming Gaussian spreads – of kinetic parameters is shown to be relevant in order to provide an accurate representation of the experimental results in some cases. In addition, we notably discuss the lower level discrimination threshold and its effect on the gamma PSDs. Gamma ray cross PSDs exhibit a behaviour contrary to that of a standard neutron acquisition system, and we discuss the implications. Gamma noise could prove to be a cheaper, more precise, and more flexible method to determine reactor kinetics.

An enhanced formalism for the inverse reactor kinetics problem

Abstract The inverse kinetics problem in reactor physics is a standard formalism to unfold reactivity on the basis of registered power (flux) profile. The classical inverse point kinetics framework has been retrofitted herein to comprise thermal reactivity feedback effects. The instantaneous fuel and coolant temperatures are thus computed by way of the exponential time-differencing scheme and the corresponding thermal reactivity feedback is plugged into the inverse kinetics module. The core external reactivity is therefore unfolded employing only two consecutive time-steps of the power (flux) profile. A history independent yet straightforward numerical routine is accrued enjoying noticeable robustness with regards to the time-step resolution.

Optimized fractional-order PID controller based on nonlinear point kinetic model for VVER-1000 reactor

Abstract Nuclear reactor dynamics are nonlinear and time-varying, so the power level control is a challenging problem in nuclear power plants (NPPs) to ensure both its operation stability and efficiency. An important measure to improve the safety of the reactor core of NPP is the implementation of robust control for the core by adjusting the inserted reactivity of the control rods. Thus in the present paper, fractional-order PID (FOPID) controller is developed as it is well known for its simplicity and robustness against disturbances. A Genetic Algorithm (GA) is used to determine FOPID controller parameters to achieve the desired power level for the generation III+ reactor VVER-1000. Implementing the GA, a suitable objective function is proposed to search for the optimal FOPID parameters. The nonlinear model of the VVER-1000 nuclear reactor is presented based on the point kinetic equations with six delayed neutron groups and temperature feedback from lumped fuel and coolant temperatures. Two cases for the VVER-1000 reactor are investigated; the changes in the power loads and the control rod withdrawal that leads to reactivity disturbance. Moreover, the uncertainties that result from model parameters perturbation are added to examine the controller robustness. The simulation results show that the proposed optimized FOPID controller can track the desired power level of the VVER-1000 reactor and robustly cope with any load changes, disturbances, or any parameters uncertainties. Also, it proves the superiority of the proposed optimized FOPID controller over other PID controllers in ensuring the safe and effective operation of the VVER-1000 reactor.

Inert-matrix fuel for transmutation: Selected mid- and long-term effects on reprocessing, fuel fabrication and inventory sent to final disposal

Partitioning and transmutation (P&T) fuel cycles provide a technical approach to ease the problem of radioactive waste disposal. Some of the partitioned components of the waste stream are irradiated while others can be used for energy production or are sent to final storage. Minor actinides are planned to be irradiated in a fast spectrum nuclear facility to transmute them into stable or short-lived isotopes. As minor actinides have negative effects on reactor dynamics, subcritical, accelerator-driven systems are proposed to increase their fraction in the fuel. An example is the MYRRHA research reactor to be built in Mol, Belgium.This reactor is modeled for depletion calculations. The behavior of special fuel elements that mirror fuel composition as envisioned for large scale transmutation facilities, namely inert-matrix fuels with an increased minor actinide content, are investigated within this reactor environment. It turns out that gamma dose rates, activity and residual heat from the spent fuel elements present significant challenges for implementing a P&T program. Spent inert-matrix fuel element show significantly higher levels than spent fuel elements from fast reactors. This requires long cooling periods and poses unprecedented challenges to reprocessing technology. The problem is amplified by the fact that it is generally agreed upon that due to low transmutation efficiencies several transmutation steps would be necessary. Looking at the radiotoxicity index, the efforts suggested to reduce the minor actinide content in a final repository are justified. The long-term safety case of deep geological repositories, however, implies that certain long-lived fission products are more relevant. The build-up of some of these radionuclides is investigated for two hypothetical German P&T scenarios. Naturally, the amount of fission products increases with continued irradiation. But namely the fraction of Cs-135 increases over-proportionally when inert-matrix fuel rich on minor actinides is used.

Modelling and simulating CO and CO2 methanation over Ru/γ-Al2O3 catalyst: An integrated approach from carbon capture to renewable energy generation

The looping of carbon capture with power-to-gas technology is getting tremendous focus for the efficient inception of the methanation process. This will not only meet the growing energy demand but also reduce fossil fuel dependency and the climate change effects associated with it. In this work, a simulation approach was developed to model the methanation of CO and CO2 on a ppm basis over a ruthenium catalyst. The fast equilibrium reaction of CO2 to CO and slow kinetically-controlled hydrogenation of CO to CH4 was modelled within the same reactor. The catalyst particles of 0.5% Ru on γ-Al2O3 pellet supports were simulated on the Aspen Tech program. The fixed bed reactor was split into segments, consisting of equilibrium and kinetic reactors to model the relevant reactions. The effect of the reaction temperature, feed composition, and reactor volume was investigated. These parameters affected the reactor’s conversion with the reaction rate and the number of reactor segments required for convergence. An approach was developed to size a reactor for specific CO and CO2 conversions. With a reactor size of 13.85 m3, a 99.0% and 97.55% conversion of CO and CO2, respectively was achieved with 96.50% CH4 yield. This proved to have a high level of accuracy, with a maximum difference between the graphical and simulated conversions of 1.96%. Furthermore, the reaction mechanism was also elucidated, showing associative adsorption based CO2 and CO methanation over Ru-catalyst, forming oxygenated intermediates that subsequently hydrogenated to CH4. Altogether, results suggest this combined carbon capture approach and its utilization with power-to-gas technology can effectively contribute to sustainable renewable energy.

Neutronics analysis of a source-driven subcritical system using G4-STORK

As a promising design of the next-generation reactors, source-driven subcritical systems (SC) are compelling for their safety and transmutation features. However, singular properties of SCs bring insuperable difficulties to their neutronics analysis with conventional numerical methods. Instead of applying oversimplified approximations, the limitations of deterministic approaches and quasi-static Monte Carlo approaches can be surmounted using G4-STORK (Geant4 STOchastic Reactor Kinetics), a time-dependent Monte Carlo particle tracking code for reactor physics applications. In this work, exploiting the exclusive features of G4-STORK, we demonstrate a novel analysis methodology by presenting a specialized neutronics analysis of an SC. A detailed inspection of neutron distributions, integral parameters, and source-induced kinetics are provided to reveal the singular behaviors neglected by oversimplifications.

Storage-Coupled Nuclear Combined Cycle

AbstractA new design paradigm for nuclear power plants is needed to complement the increasing adoption of low marginal cost variable renewable energy resources. The situation is reflected in the wholesale electricity price–duration curve with four distinct economic opportunities: (a) a hundred or so hours per year of high-value peaking power; (b) about 4000–5000 h of moderate electric prices; (c) about 2000 h per year when renewables set the marginal price at or near zero; and (d) about 1000 h of flexible ramping between the b and c regions. The current approach to the low-carbon energy transition reduces the need for baseload power and requires curtailment of conventional, nuclear, and even renewable generation, decreasing their capacity factors and increasing their fixed charges for electricity generation. Flexible low-carbon dispatchable power plants capable of daily cycling along with storage and time shifting of low-cost nondispatchable renewable power will be needed. Although nuclear power plants have demonstrated load-following capability, cycling can be limited by reactor kinetics (xenon poisoning) as well as by thermal stresses and fatigue considerations in the steam cycle. Storage of nuclear heat is hampered by the relatively low operating temperatures of existing nuclear reactors (but not advanced reactors) that lowers thermal to electric conversion efficiency, which in turn increases the required quantity of storage medium and the cost of storage. The quantity of storage medium can be reduced by integration of thermal energy storage with high-grade heat as in the liquid salt combined cycle (LSCC). The LSCC uses high-temperature gas turbine exhaust heat to increase the electricity output per unit of storage medium, uses the stored energy to add operating flexibility to a bottoming steam cycle, and substantially reduces the fuel heat rate. The low fuel heat rate improves economic competitiveness compared to alternative gas turbine-based power plants, especially when burning expensive fuels such as hydrogen. LSCC could be coupled to a nuclear power plant for time shifting both nuclear and renewable electricity and could support high utilization of a co-located hydrogen electrolysis plant. Further cost reduction could be achieved by using solid media for thermal energy storage, with the liquid salt used as a heat transfer medium.

Kinetic Modeling and Reactor Design of the Direct Synthesis of Dimethyl Ether for Co2 Valorization. A Review

Kinetic Modeling and Reactor Design of the Direct Synthesis of Dimethyl Ether for Co2 Valorization. A Review

Визначення консервативних умов моделі реактивнісної аварії на РБМК‑1000

Analytical studies have been performed on the model of a reactivity accident at the 4th Unit of the Chornobyl NPP with an RBMK-1000-type reactor. The model of the RBMK-1000 reactor was developed based on the equations of nuclear reactor kinetics. In the model, the reactivity changes as a result of external influences (the movement of control rods; changes in the coolant temperature at the reactor inlet), as well as a result of feedback on reactivity effects (changes in fuel temperature, coolant density, 135Xe concentration). The change in the coolant density takes into account the formation of steam in the reactor core, and the coolant pressure is introduced into the model as an external factor, according to the results of the registered data during the accident on 26.04.1986. There is a high sensitivity of the RMBK-1000 reactivity model to the absolute values of reactivity coefficients that have occurred on the eve of the accident (negative reactivity coefficient by fuel temperature, positive steam coefficient of reactivity). Therefore, the study is conducted for different combinations of values of the efficiency of the control rods, reactivity coefficients, as well as other factors affecting the course of the accident — emergency protection triggering time, and reactor power level before the accident. Considering that the main stage of the accident lasted less than 10 s, fuel destruction is possible when the critical value of fuel enthalpy is reached, at which the fuel dispersion process begins. The results of modeling of reactivity accident on RBMK-1000 with the values of parameters of reactivity effects, which best correspond to the chronology of recorded events as well as to the recorded values of technological parameters, are presented.

Neutron Multiplication and Reactor Kinetics Parameters in Stochastic Media Based on Randomized Noise Functions

Neutron Multiplication and Reactor Kinetics Parameters in Stochastic Media Based on Randomized Noise Functions

Theta Method Applied to Point Reactor Kinetics Equations with Two Energy Groups

Theta Method Applied to Point Reactor Kinetics Equations with Two Energy Groups