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The inverse reactor period α is a fundamental mode eigenvalue of the α-mode nonlinear Boltzmann eigenvalue equation that considers delayed neutron contributions. Thus far, several Monte Carlo methods, including the α-k, weight balancing, and transition rate matrix methods, have been developed to calculate α. This study presents a new Monte Carlo method for predicting α by using the derivatives of the k-eigenvalue with respect to α. Formulae are derived to calculate the first and second derivatives using the differential operator sampling method. The key feature of the new proposed method is its ability to estimate the uncertainty of the predicted α by considering the uncertainty of the k-eigenvalue and its derivatives with respect to α. [ABSTRACT FROM AUTHOR]

Building upon the previous research that solved neutron diffusion equations in simplified slab geometry, this study advances the field by addressing the more complex cylindrical geometry, focusing on neutron diffusion equations that are coupled with delayed neutrons in cylindrical reactors of fractional order. The method of solving used integrates the technique of residual power series (RPS) with the Laplace transform (LT) method. Anomalous neutron behavior is explained by examining the non-Gaussian scenario with various fractional parameters α. The LRPSM Laplace transform and residual power series method employed in this approach eliminates the complex difficulties. This simplicity makes the method particularly coherent with different fractional calculus applications. To validate the proposed method, numerical simulations are conducted with two different initial conditions representing distinct scenarios. The obtained results are presented in suitable tables and figures. It should be emphasized that this system is solved for the first time utilizing fractional calculus techniques. The outcomes are consistent with those achieved using the Adomian decomposition method. [ABSTRACT FROM AUTHOR]

The purpose of this study is to determine the kinetic parameters of the RSG-GAS equilibrium core. The calculated kinetic parameters are the effective delayed neutron fraction βeff, the neutron generation time Ʌ, and the prompt neutron lifetime ℓ since they are related to the safety of nuclear operations. The kinetic parameters were calculated using the Serpent 2 code with the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries. Calculations were performed using various adjoint-weighted methods such as Meulekamp's method, Nauchi's method, the Iterated Fission Probability method, and the Perturbation Technique. The calculated results of the six-group delayed neutron fraction by the Meulekamp and the IFP methods showed no significant difference. Choosing the IFP method as the reference, the maximum difference for βeff (694 pcm) is 0.73%, and the maximum difference for Ʌ and ℓ is 1.89%. The calculated kinetic parameters with ENDF/B-VII.1 and ENDF/B-VIII.0 are quite close, with a maximum difference of 0.9%. The sensitivity analysis results indicate several nuclides and reaction types that dominantly affect the βeff and Λ. The results of the kinetic parameter calculations can be used for the safety analysis of the RSG-GAS equilibrium core. [ABSTRACT FROM AUTHOR]

The article presents the results of the analysis of the validity of simplifications and assumptions accepted by researchers in the models of neutron-physical characteristics of fuel-containing materials (FCM) in the “Shelter” object (SO). The analysis and substantiation of the FCM nuclear safety will be one of the important issues during the SO operation for a long time, until the moment of retrieval and disposal of nuclear materials from the ChNPP Unit 4 destroyed in 1986. A significant number of nuclear materials in the SO are uncontrolled, and the geometry of their location is also little known. Therefore, the use of traditional methods of nuclear safety analysis to determine the conditions for reaching criticality in the volume of FCM can lead to significant uncertainties, depending on the assumptions and simplifications made in FCM models. In contrast to traditional nuclear facilities with design characteristics and control systems, there is no possibility to deterministically influence the FCM, and accordingly to determine their dynamic characteristics in the SO. In such conditions, it is necessary to take a responsible approach to obtain and interpret the values of FCM parameters by control systems of the SO. For example, in 1990, one of the neutron flux detectors recorded an abnormal change in the neutron pulse count rate for about 100 hours. This event is often called a “neutron incident”, and it formed the basis of some models, in which the possibility of reaching criticality in one of the FCM clusters is assumed. The article presents the results of the study on the physical possibility of the anomalous event in 1990, as well as the results of the analysis of simplifications and assumptions made in models of FCM criticality. An important result of the conducted research is the justification of the need to include in the FCM criticality model the effects of reactivity on fuel and water temperature, as well as to take into account delayed neutrons. [ABSTRACT FROM AUTHOR]

The cumulative and independent yields of various fission products within the mass range of 78 to 157 have been measured in the epi-cadmium neutron–induced fission of 238Pu by using an off-line γ-ray spectrometric technique. The lower yield of 136I than the usual trend indicates the formation of delayed neutron emitter 137I. From the cumulative yields, post-neutron mass chain yields were obtained by using the charge distribution correction. From the mass yield data, the peak-to-valley (P/V) ratio, the full-width at tenth-maximum of light and heavy mass wings, the average light mass and heavy mass as well as the average number of neutrons <ν> emitted were obtained. The mass chain yield data in the 238Pu(n,f) reaction were compared with similar data of the 238Pu(nth,f) reaction to examine the role of excitation energy on the nuclear structure effect and P/V ratio. [ABSTRACT FROM AUTHOR]

The paper is concerned with the use of uranium wastes after purification of regenerated fuel from 232U in a molten salt reactor (MSR) in application to minor actinide (MA) burning. The comparison of the uranium and plutonium loadings in an MSR was performed. The results of calculations show that the uranium loading in an MSR is a possible option that has a number of potential advantages: it allows for a higher rate of MA burning during the initial period of MSR operation with approximately the same burnup efficiency and increases the effective fraction of delayed neutrons. The presented calculations are of an approximate nature and first of all show the possibility of effective use of uranium from the waste stream of the enrichment cascade as a fissile material. [ABSTRACT FROM AUTHOR]

Algorithms and codes were developed that facilitate the selection of the locations of ionization chambers detecting variations in reactor power. For each time step, the nonstationary neutron transport equation is solved, and the reactivity changes are found using the inverse solution of the kinetics equation in the point approximation. At each step, a set of the best points for moment in time is determined where the reactivity deviations from the reference results are minimal. At the end of the entire calculation, such points with maximum total occurrence are selected. Then the version of the location of ionization chambers selected taking design constraints into consideration is applied to explore various scenarios of motion of control rods. The study is based on the assumption that the equations of point kinetics are used in the processing of ionization chamber readings at an operating NPP. The factors that limit the use of this methodology are the inapplicability of point kinetics at high reactivity values and the potential infeasibility of reconciling the contradictory results obtained in modeling various scenarios of control rod motion. Compromise decisions should be made by experts. The algorithms and codes developed and verified on the BN-600 test model can be used for both the BN-1200 and BREST reactors and reactors of other types. [ABSTRACT FROM AUTHOR]

This article considers the fundamental possibility of identifying the kinetic characteristics of a reactor under quasi-stationary operating conditions by observing its power fluctuations under the influence of controlled changes in reactivity, as well as the possibility of optimal stabilization of reactor power. The analysis is carried out under conditions where the reactor reactivity and measurement of its power are influenced by disturbing perturbations in the form of uncontrolled random processes. A point kinetics model of the reactor with a single group of delayed neutrons and instantaneous power feedback is used. Examples are presented that illustrate the proposed identification method and the effectiveness of the optimal controller. [ABSTRACT FROM AUTHOR]

Molten salt nuclear reactors (MSRs) constitute a promising technology to produce safe, reliable, abundant low-carbon energy. To design MSR systems and perform safety analyses on them, numerical simulation is a powerful tool. Here, we implemented a coupling between several solvers of the deterministic neutronics code APOLLO3® (the MINARET SN transport and the MINOS diffusion and SPn-simplified transport solvers) and the computational fluid dynamics (CFD) code TRUST/TrioCFD, both developed at the French Alternative Energies and Atomic Energy Commission (CEA). The code coupling is orchestrated using the dedicated C3PO library of the open-source SALOME platform. A new code-coupling strategy is employed whereby the delayed neutron precursor concentrations are computed by the CFD code, which eases the use of traditional deterministic neutronics codes. We verified the correctness of our implementation by performing a numerical benchmark dedicated to fast spectrum MSRs originally devised by the French National Center for Scientific Research. The numerical results we obtained are in excellent agreement with those obtained by recent MSR-dedicated multiphysics simulation tools. This study provides a new convenient neutronic–thermal-hydraulic coupling strategy for MSR core simulation. [ABSTRACT FROM AUTHOR]

Changes in temperature during reactor operation may cause changes in physical parameters, leading to core overheating and accidents. It is essential to analyze and assess the safety parameters of the core under different operating conditions. This paper investigates the effects of fuel temperature, moderator density, boron concentration, and control rod state on AP1000 safety parameters. The study uses RMC and NJOY to calculate the changes in reactivity factor, effective delayed neutron fraction, and neutron generation time of the AP1000 reactor under different operating conditions. The changes in reactivity coefficients, neutron fluxes, and relative power densities of AP1000 reactors are analyzed for normal and accidental operating conditions. The results indicated that the reactivity coefficient remained negative under accident conditions, which ensured the safe operation of the reactor. The delayed neutron fraction, neutron flux, and power density distributions are affected by fuel temperature, moderator density, and control rod position. The control rod worth was sufficient for the emergency shutdown of the reactor under accidental conditions. It is demonstrated that the operation of the AP1000 reactor under study conditions is safe and controllable. [ABSTRACT FROM AUTHOR]

The paper presents the results of critical, correlation, and spectral precision experiments carried out at the FKBN‑2 assembly machine using a cylindrical multiplying system made of low-background α‑phase plutonium. In the experiments, the critical state of the multiplying system was determined, the characteristics of non-stationary prompt neutron processes were investigated, and the number of nuclear reactions in the detectors was measured at a controlled power level of the multiplying system. After constructing a benchmark model for the multiplying system, a computational simulation of integral experiments was carried out. Based on the experimental results, the average lifetime of prompt neutrons and the effective fraction of delayed neutrons for the multiplying system were determined. The obtained results can be used to verify the codes of neutron-physics and refine neutron constants. [ABSTRACT FROM AUTHOR]

Grade control is crucial for ensuring that the quality of extracted ore aligns with the geological model and mining plan. This process optimises production, reduces dilution, and maximises profits. It involves geological modelling, sampling, assaying, and data analysis. However, adhering to short-term planning in mining operations can be challenging due to operational bottlenecks that arise during the grade control process and blast design, along with their associated costs. Industry standards for grade control require acquiring extensive information and knowledge to achieve a high level of certainty, which takes time. Despite that, time constraints may necessitate making decisions under risk with incomplete information. In such cases, it is important to consider the opportunities, risks, likelihood, consequences, and potential success associated with each alternative. This study presents the testing results of alternative quantitative analytical methods on samples from the Barruecopardo tungsten deposit in Spain. Spectrometric techniques, including Delayed Gamma Neutron Activation Analysis (DGNAA), Laser-induced Breakdown Spectroscopy (LIBS), and Field Portable X-ray Fluorescence (FPXRF), were employed to determine the tungsten content. Based on the findings of this investigation, a real-time decision-making tool for grade control in open-pit mining has been developed. This tool utilises representative samples directly from the blasting debris, considering the inherent risks and uncertainties associated with the process. [ABSTRACT FROM AUTHOR]

In the work the computer model of a cell of a system for generating fluxes of therapeutic beams of delayed neutrons, based on the use of delayed fission neutrons, was developed in the Geant 4 environment. The principle of such a neutron source is that when a powerful electron beam interacts with a combined tungsten target and a target containing fissile material, a fission reaction occurs; as a result of which neutrons are emitted. If we move a target activated in this way several tens of meters into a neutron flux generation system consisting of a heater, protection, collimator and reflector, we will obtain a compact neutron source for nuclear medicine. A significant advantage of such a neutron source is the absence of gamma background from the electron accelerator and the combined target, and a bulky protection system is not required. In the Geant 4 environment, the geometry of this cell was developed and a series of experiments were carried out with 107 neutrons. The QGSP BIC HP physical sheet was used. A study of neutron energy spectra showed that more than half of the neutrons whose fluxes are formed using such a cell of the formation system have an energy

The Heavy Water Zero Power Reactor (HWZPR) is a 100‑W research reactor that uses natural uranium metallic fuel and heavy water as moderator. The HWZPR is designed for training and research in reactor physics fields. The main aim of this paper is to evaluate the variation of neutronic parameters of a HWZPR resulting from changes in the fuel lattice pitch using MCNPX code. In such a manner, neutronic parameters including effective multiplication factor (keff) for critical water level, effective delayed neutrons fraction (βeff), prompt neutrons lifetime (lp), and neutron flux distribution are investigated and benchmarked. The results illustrated that, as the lattice pitch is increased, more heavy water is needed for criticality condition according to the decrease in the number of fuel rods and fewer available fissile materials. Comparison of the results with the reactor Safety Analysis Report shows reasonable agreement. [ABSTRACT FROM AUTHOR]

The low rate at which some fissile isotopes, such as 235U, 233U, and 239Pu, undergo spontaneous fission leads to a weak signal, resulting in a high-uncertainty in applying passive neutron counting techniques. Stimulating fission through active neutron interrogation can overcome this issue. At Canadian Nuclear Laboratories, a 252Cf and a deuterium-deuterium neutron source are available. In this study, a neutron counting system was designed to perform passive measurements and active neutron interrogation for a search of special nuclear material. The detection system consists of a cylindrical cavity surrounded by a polyethylene moderator with 3He detectors interspersed throughout. When used for passive measurements, the sample is placed in the cylindrical cavity, whereas in active interrogation mode, the 252Cf neutron source and the sample are placed in close proximity to each other in the cylindrical cavity. Measurements that actively interrogated samples, notably containing (among other isotopes) either 235U or 239Pu whose mass was on the order of fractions of a gram, carried out using the 252Cf neutron source found that the average delayed neutron count rate was on the same order of magnitude as those obtained from passive measurements using several kilograms of natural uranium. The Monte Carlo N-Particle 6 version 2.0 radiation transport code was used to simulate the aforementioned active interrogations and to inform the experimental results. Results showed that, due to the close proximity of the polyethylene moderator to the 252Cf source, the neutron energy spectrum traversing the fissile sample has a significant thermal component that maximizes the fission reaction rate in the interrogated fissile samples, thereby allowing for successful measurements. [ABSTRACT FROM AUTHOR]

This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7 FPGA board for determining reactivity values by solving the point reactor kinetics equations with six delayed neutron groups and for stabilizing the reactor power at preset levels by determining the unbalance voltage signal to control the automatic control rod. With real-time calculations, the newly developed controller can monitor the reactor reactivity and control the reactor power online. The developed controller unit's reactivity measuring and power stabilizing capabilities were assessed using the DNRR in normal operation and assumed emergency conditions and compared with those of the preexisting imported BNO-102R1 module of the DNRR control and protection system, known as ASUZ-14R. The results of the experiments show that the produced FPGA-based unit and the BNO-102R1 unit have the same technical characteristics and features, with the disparities being less than 5% and 1%, respectively, in reactivity measurement and power stabilization. The experimental data of reactivity measurements by the FPGA-based unit and the calculation results were also compared and found that the relative deviations between those are also less than 10%. The developed controller unit is capable of carrying out a variety of training and operational experiments on the DNRR. [ABSTRACT FROM AUTHOR]

We present a new approach to calculating time eigenvalues of the neutron transport operator (also known as α eigenvalues) by extending the dynamic mode decomposition (DMD) to allow for nonuniform time steps. The new method, called variable dynamic mode decomposition (VDMD), is shown to be accurate when computing eigenvalues for systems that were infeasible with DMD due to a large separation in timescales (such as those that occur in delayed supercritical systems). The α eigenvalues of an infinite medium neutron transport problem with delayed neutrons, and consequently having multiple, very different relevant timescales, are computed. Furthermore, VDMD is shown to be of similar accuracy to the original DMD approach when computing eigenvalues in other systems where the previously studied DMD approach can be used. [ABSTRACT FROM AUTHOR]

The neutron diffusion equation (NDE) is one of the most important partial differential equations (PDEs), to describe the neutron behavior in nuclear reactors and many physical phenomena. In this paper, we reformulate this problem via Caputo fractional derivative with integer-order initial conditions, whose physical meanings, in this case, are very evident by describing the whole-time domain of physical processing. The main aim of this work is to present the analytical exact solutions to the fractional neutron diffusion equation (F-NDE) with one delayed neutrons group using the Laplace transform (LT) in the sense of the Caputo operator. Moreover, the poles and residues of this problem are discussed and determined. To show the accuracy, efficiency, and applicability of our proposed technique, some numerical comparisons and graphical results for neutron flux simulations are given and tested at different values of time and order which includes the exact solutions (when Finally, Mathematica software (Version 12) was used in this work to calculate the numerical quantities. [ABSTRACT FROM AUTHOR]

This article theoretically confirms the possibility of achievement of oscillatory instability on a portion of groups of delayed neutrons with short decay periods for some known models-mechanisms formed by a chain of negative feedbacks "power–temperature (density)–reactivity–power." Here, in order to provide instability on a lower number of groups, a higher feedback gain is required. The parametric and regime conditions stabilizing the system are mentioned. [ABSTRACT FROM AUTHOR]

There is an urgent need for new, better instrumentation and techniques for detecting and characterizing special nuclear material (SNM), i.e., highly enriched uranium and plutonium. The development of improved instruments and techniques requires experiments performed with the SNM itself, which is of limited availability. This paper describes the findings of experiments performed at the National Criticality Experiments Research Center conducted using new instruments and techniques on unclassified, kg-quantity SNM objects. These experiments, performed in the framework of the Department of Energy, National Nuclear Security Administration Consortium for Monitoring, Technology, and Verification, focused on detecting, characterizing, and localizing SNM samples with masses ranging from 3.3 to 13.8 kg, including plutonium and highly enriched uranium using prototype detectors and techniques. The work demonstrates SNM detection and characterization using recently-developed prototype detection systems. Specifically, we present new results in passive detection and imaging of plutonium and uranium objects using gamma-ray and dual particle (fast neutron and gamma-ray) imaging. We also present a new analysis of the delayed neutron emissions during active interrogation of uranium using a neutron generator. [ABSTRACT FROM AUTHOR]

In the thermal neutron–induced fission of 233U, the cumulative and independent yields of various fission products within the mass ranges of 77 to 109 and 123 to 155 have been measured by using an off-line gamma-ray spectrometric technique. The lower yields of 86Br and 136I than usual trend indicate the formation of delayed neutron emitters 87Br and 137I. From the cumulative yields, the post-neutron mass yield distribution was obtained after applying the charge distribution correction. The data from the present and earlier work of our laboratory in the 233U(nth,f) reaction were compared with similar data of 232,235U(nth,f) and 238U(n,f) reactions to examine the effect of mass difference of the fissioning systems on the fine structure of the mass yield distribution. The mass yield distribution in the 233U(nth,f) reaction was also compared with those of 229Th(nth,f), 241Pu(nth,f), and 245Cm(nth,f) reactions to examine the effect of charge and mass difference of the fissioning systems and also to examine the different behaviors of standard I and standard II asymmetric modes of fission. [ABSTRACT FROM AUTHOR]

For the sake of understanding the mechanism of deep subcriticality and high heterogeneity of neutron fluence rate in time–space on the neutron kinetics of the Accelerator-driven Subcritical Reactor Subcritical Reactor (ADSR) under varied beam transients and neutron spectra. A Monte Carlo homogenization approach for the neutron time–space kinetics of the ADSR is proposed in this study, and the influence mechanism on the kinetic parameters of the ADSR under varied neutron spectra, subcriticality, and beam transients is examined. The results show that the Monte Carlo homogenization for the α eigenvalue mode is more adaptable to the subcriticality characteristics under varied subcriticality; under beam transients, the relative differences in the kinetic parameters of the different modes of the ADSR with fast/thermal spectra increase with the depth of subcriticality, and the differences in neutron generation time for varied modes are larger than those of effective fraction of delayed neutron. Thus, it is recommended to use a more adaptable Monte Carlo homogenization method for the time–space kinetics of ADSR, and the effects of the high heterogeneity of neutron fluence rate and deep subcriticality in time–space on the neutron generation time should be considered. [ABSTRACT FROM AUTHOR]

In this article, a new analytical scheme of the ARA transform is introduced to solve systems of fractional partial differential equations. The principle of the proposed technique is based on combining the ARA transform with the residual power series method to create an approximate series solution for a system of partial differential equations of fractional order on the form of a rapid convergent series. To illustrate the effectiveness, accuracy, and validity of the suggested technique, an Attractive physical system, the fractional neutron diffusion equation with one delayed neutrons group, is discussed and solved. Two different neutron flux initial conditions are presented numerically to clarify various cases in order to ensure the theoretical results. The necessary Mathematica codes are run using vital nuclear reactor cross-section data, and the results for various values of time are tabulated and graphically represented. [ABSTRACT FROM AUTHOR]

The results of calculating the βeff effective fraction of delayed neutrons and the Λ lifetime of prompt neutrons, both when taking into account and excluding the value function, are presented. The analysis, concerning the effects of accounting for the neutron value function at various degrees of its expansion, as well as depending on the physical and geometric parameters of the system, is conducted. The calculations were performed for twelve experiments from the ICSBEP collection. The analysis of the results states the necessity of accounting for the value function in small systems with a reflector and a large number of neutrons. No accounting for the value function is required for estimating the kinetic functionals in systems that are close to infinite. [ABSTRACT FROM AUTHOR]

In this paper, a system of coupled fractional neutron diffusion equations with delayed neutrons was solved efficiently by using a combination of residual power series and Laplace transform techniques, and the anomalous diffusion was considered by taking the non-Gaussian case with different values of fractional parameter α. The Laplace residual power series method (LRPSM) does not require differentiation, conversion, or discretization for the assumed conditions, so the approach is simple and suitable for solving higher-order fractional differential equations. To assure the theoretical results, two different neutron flux initial conditions were presented numerically, where the needed Mathematica codes were performed using essential nuclear reactor cross-section data, and the results for different values of times were tabulated and graphically figured out. Finally, it must be noted that the results align with the Adomian decomposition method. [ABSTRACT FROM AUTHOR]

The Frequency Transform method is used for the first time to efficiently model a multiple-second transient problem with Monte Carlo (MC). This is achieved by coupling MC with a time-dependent coarse mesh finite difference (TD-CMFD) diffusion solver. TD-CMFD presents a large advantage over commonly used point kinetics equations since it preserves spatial resolution during the transient and provides equivalence with the high-order method through nonlinear diffusion coefficients. As TD-CMFD computes time-dependent and spatially dependent neutronics information, it also computes frequencies that describe the rate of change of neutron and delayed precursor concentrations. These frequencies are used in MC shape function calculations as an approximation for the time derivatives. As the simulation proceeds, MC calculations update the multigroup cross sections, currents, and diffusion coefficients that are needed in TD-CMFD, and in turn, TD-CMFD updates the frequencies. Our results show the success of the Frequency Transform method in prescribed transient problems on the C5G7 geometry and on a fuel pin geometry. The Frequency Transform method showed significant improvement compared to the Adiabatic approximation, which does not use any frequency information in the MC calculation. The improvements in spatial resolution are shown to be a direct result of frequencies. Additionally, a study of how TD-CMFD's nonlinear diffusion coefficients behave in time provides a first-of-its-kind study of how equivalence factors are impacted by transients. [ABSTRACT FROM AUTHOR]

We use the deterministic neutron transport code QuasiMolto to simulate steady-state operation of the Molten Salt Reactor Experiment (MSRE). Comparisons are made to similar results from the MOST benchmark, the MOOSE-based code Moltres, and the design calculations for the MSRE. In the course of these comparisons, we calculate a value of 0.1799 for the graphite-to-fuel power density ratio, which differs significantly from that seen in other works. We also find uniform graphite heating inadequate to reproduce the characteristic graphite temperature distribution of the MSRE. Leveraging the multilevel projective methodology of QuasiMolto, the influence of transport effects on the modeled problem is found to produce average and maximum group flux variations of 2% to 5% and 30%, respectively, with a 12% variation in the reactivity loss due to delayed neutron precursor drift. [ABSTRACT FROM AUTHOR]

Neutron diffusion kinetics equation contains a diffusion type equation and a set of ordinary differential equations, which form a coupled system. In this paper, we give a positivity theorem for neutron diffusion kinetics equation with delayed neutrons and propose a nonlinear positivity-preserving finite volume scheme on general star-shaped quadrilateral meshes, which can deal with discontinuous and anisotropy diffusion coefficient problems. The existence of solution for the nonlinear system is proved under certain constraint on time step size, which is independent of spatial mesh size. Three nonlinear iteration strategies are introduced to solve the nonlinear system. Numerical experiments are performed to demonstrate that the scheme is positivity-preserving and keeps approximately second order accuracy in space. The computational costs of three nonlinear iteration strategies are compared. Another example is also present to show some constraint on time step size is necessary for the convergence of the nonlinear iteration. [ABSTRACT FROM AUTHOR]

In an attempt to find an optimum fuel, which can withstand the harsh operation condition in the VVER-1200 reactor, different fuel types were investigated in this work. Uranium carbide and uranium nitride are proposed as alternative fuels to UO2. MCNPX 2.7 with cross-section library ENDF/B-VII was used to design 3D model of VVER-1200 to investigate the neutronic properties of the suggested fuel types. The output results of the burnup of the proposed fuels have been analyzed to determine which of them achieves the longest fuel cycle. It was found that UO2 and U12C achieved criticality till 35 GWD/MTHM, while U14N and U15N achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. The traditional method of reactivity management has been tested on the proposed fuel to determine to what extent it is valid for them. To ensure the operation safety of the proposed fuel, some safety factors such as moderator temperature coefficient (MTC‎), fuel temperature coefficient (FTC‎), and effective delayed neutron fraction (βeff) have been calculated. The distribution of the thermal neutron flux and normalized power has been displayed to give us a good insight view of the VVER-1200 assembly to determine which of the proposed fuels is optimum compared to UO2. [ABSTRACT FROM AUTHOR]

The nuclear and non-nuclear industry has produced a considerable amount of low and intermediate-level radioactive wastes during the last decades. The material characterization of waste packages recently became more and more important in order to dispose of these waste packages in a final underground repository. Material characterization remains an indispensable criterion to prevent pollution of the groundwater with toxic materials and is usually required by the national licensing and supervisory authorities. Information on the nature of waste materials can be obtained based on existing documentation or, if the documentation is insufficient, on further destructive or non-destructive analysis. Non-destructive methods are to be preferred to minimize radiation exposures of operating personnel as well as costs. Existing non-destructive techniques (Gamma scanning, X-ray, active/passive neutron counting, muon tomography) do not allow the identification of non-radioactive hazardous substances. An innovative non-destructive measurement system called QUANTOM® (QUantitative ANalysis of TOxic and non-toxic Materials) has been developed. It is based on the prompt and delayed gamma neutron activation analysis (P&DGNAA). This technology is able to identify and quantify the elemental composition (Cd, Cu, B, Pb, Hg, Fe, Al, ...) in radioactive packages such as 200-l radioactive drums. This information helps waste producers verify the content of their radioactive wastes, especially regarding the presence of hazardous substances. Different reference materials have been analysed by means of the same technology (P&DGNAA) at the research reactor of BUDAPEST. A comparison of those results for five reference materials is presented. The results show a very good agreement between QUANTOM® and standardized reference analyses. [ABSTRACT FROM AUTHOR]

The generation of accurate kinetic parameters such as mean generation time Λ and effective delayed neutron fraction βeff via Monte Carlo codes is established. Employing these in downstream deterministic codes warrants another step to ensure no additional error is introduced by the low-order transport operator when computing forward and adjoint fluxes for bilinear weighting of these parameters. Another complexity stems from applying superhomogenization (SPH) equivalence in non-fundamental mode approximations, where reference and low-order calculations rely on a 3D full core model. In these cases, SPH factors can optionally be computed for only part of the geometry while preserving reaction rates and K-effective, but the impact of such approximations on kinetics parameters has not been thoroughly studied. This paper aims at studying the preservation of bilinearly-weighted quantities in the Serpent–Griffin calculation procedure. Diffusion and transport evaluations of IPEN/MB-01, Godiva, and Flattop were carried out with the Griffin reactor physics code, testing available modeling options using Serpent-generated multigroup cross sections and equivalence data. Verifying Griffin against Serpent indicates sensitivities to multigroup energy grid selection and regional application of SPH equivalence, introducing significant errors; these were demonstrated to be reduced through the use of a transport method together with a finer energy grid. [ABSTRACT FROM AUTHOR]

Long-lived minor actinides (MAs) are one of the primary contributors to the long-term radiological hazards of nuclear waste, and the buildup of MAs is hampering the development of nuclear power. The transmutation of MAs in reactors is regarded as a potential way to replace direct disposal to reduce the impact of MA on the environment and improve the utilization of fuel. Due to its superior features, such as outstanding neutron economy, no fuel assembly fabrication, high neutron flux, and especially online refueling and reprocessing, the molten chloride salt fast reactor (MCFR) is regarded as one of the potential reactors for MA incineration. In this work, MA transmutation capability and 233U breeding performance for an optimized MCFR have been evaluated in different scenarios. The results show that the MA transmutation capability and 233U breeding performance with online transuranic elements (TRU) and 232Th feeding scenario are improved significantly compared with the case in online 233U and 232Th feeding, when the initial MA loading is 5 mol%, the total mass of MA transmutation and MA incineration is 7160 kg and 1759 kg during the whole 100 years operation under online TRU and 232Th feeding scenario, and the corresponding average annual net production of 233U is 450 kg, however, the MA transmutation amount, MA incineration amount and average annual net production of 233U for online 233U and 232Th feeding scenario is 5298 kg, 1315 kg, and 249 kg, respectively. In addition, the research also shows that the increase in initial loading of MA has no obvious effect on the improvement of the 233U breeding performance but can improve the transmutation efficiency of MA under online TRU and 232Th feeding scenarios. Furthermore, if 233U is continuously extracted online from the core during the operation, the 233U breeding performance will be significantly improved, but it will deteriorate the safety performance, such as the fuel temperature coefficient of reactivity (TCR) and the effective delayed neutron fraction (EDNF), more importantly, it will also put forward higher requirements for the immature online reprocessing technology. [ABSTRACT FROM AUTHOR]

An experimental and calculation method for the determination of the effective fraction of delayed neutrons in fast multiplying systems in the delayed criticality state is presented. A physical quantity is found for such systems that can be determined in experiments and calculations and used for testing and adjusting neutron constants. The method has been applied to the analysis of experimental data for systems consisting of (highly enriched) metal uranium and plutonium. [ABSTRACT FROM AUTHOR]

In molten salt reactors (MSRs) with circulating liquid fuel, delayed neutron precursors can escape from the active core and be distributed over the entire reactor circuit. Taking into account the decrease in the effective fraction of delayed neutrons due to the motion of the emitters is important for further analysis of the reactor kinetics. A new approach is proposed to account for fuel circulation in the fuel circuit on the basis of a combined calculation by the Monte Carlo and finite volume methods. The calculation was implemented using the interface between the codes SERPENT and OpenFOAM software applications (SAs) in the form of the OFSI software. Calculations of the effective delayed neutron fraction for the 10-MWt heat power MSR reactor circuit with a cavity type cylindrical core and the Li, Be, Pu/F fuel salt mixture are performed. On the basis of the results obtained, conclusions are made about the applicability of the technique and ways of its further development are outlined. [ABSTRACT FROM AUTHOR]

Nonstationary calculations of the benchmark model of the MET-1000 fast reactor with a metal fuel created on the basis of the international stationary benchmark with a hard spectrum with a thermal power of 1000 MW are discussed. Prompt dropping of all control rods and that of a single central rod are simulated. The advantages of this benchmark are its simplicity and small runtimes in the symmetry of 60 degrees with conservation of all main features required for modeling the nonstationary processes in fast reactors. [ABSTRACT FROM AUTHOR]

The molten chloride salt fast reactor (MCFR) with a closed Th–U fuel cycle is receiving more and more attention due to its excellent performance, such as high solubility of actinides, superior breeding capacity, and good inherent safety. In this work, the neutronics performances for different minor actinides (MA) loadings and operation modes are analyzed and discussed based on an optimized MCFR. The results indicate that online continuous reprocessing can significantly increase the transmutation performance of MAs. In addition, MA loadings have an obvious effect on the neutronics characteristics of the MCFR, and it is helpful for improving the MA transmutation capability and 233U breeding performance, simultaneously. When MA = 5 mol%, the average annual MA transmutation mass and incineration mass can achieve about 53 kg and 13 kg, respectively, and the corresponding annual net production of 233U is 250 kg. When MA = 33.5 mol%, the annual MA transmutation mass and incineration mass can be about 310 kg and 77 kg, respectively, and the corresponding annual net production of 233U is 349 kg. However, when the MA loadings exceed 10%, the corresponding keff will exceed 1.1 for decades, even if only Th is continuously fed online. The results also indicate that the transmutation ratio (TR) and incineration ratio (IR) of MA increase and reach maximum values in the first decades for all the different MA loadings, which means MA may be fed into the fuel salt to improve its transmutation capability. Moreover, though MA loading will increase the level of radiotoxicity of the core in the early stage of burnup, the radiotoxicity of MA will drop rapidly after a brief rise during the operation. It can also be found that the temperature coefficient of reactivity (TCR) of all different MA loadings can be negative enough to maintain the safety of the MCFR during the whole operation, although it decreases in the beginning of life (BOL) with the increasing MA loading. Furthermore, the evolution of an effective delayed neutron fraction (EDNF) is also researched and discussed, and the EDNF varies most significantly when loading MA = 35.5 mol%, with a range of 273 to 310 pcm over the entire 100 years of operation. [ABSTRACT FROM AUTHOR]

Summary: A thorium‐based chloride salt fast reactor (TCLFR) for transuranics (TRUs) transmutation is proposed to decrease radioactive nuclear waste. The TCLFR core uses two types of fuel: liquid nuclear fuel with TRUs and thorium being dissolved in chloride and solid transmutation rod with only TRUs fuel. The transmutation capability of TCLFR with a core power of 2500 MWth is investigated utilizing three different reactor core configurations, namely ThCore, GdCore, and TcCore, and compared to the standard Molten Chloride Fast Reactor (MCFR) while maintaining most of the geometric parameters. Important burnup parameters such as transmutation rate (TRE) and transmutation ratio (TRO) are extracted and analyzed. Besides, the neutron spectrum shift and the relevant safety parameters, including power peak factor, temperature coefficient of reactivity (TCR) and effective delayed neutron fraction (βeff) are all explored. The results demonstrate that all the TCLFR cores can transmute TRUs effectively. TRUs have a high online refueling amount of TRUs in TcCore and GdCore due to their low conversion ratio (CR). The TRE in the TcCore and GdCore is 292.77 kg/GWth/year and 258.31 kg/GWth/year, respectively, which is significantly higher than the TRE in the MCFR, which is 171.21 kg/GWth/year. As a result of the production of U‐233 compensates for the consumption of TRUs, the refueling amount of TRUs decreases, and the TRE in the ThCore is only 96.29 kg/GWth/year, which is approximately 44% lower than that in the MCFR. [ABSTRACT FROM AUTHOR]

In practical implementations, most nuclear reactors employ digital controllers to accomplish the load following operation. However, most of the pre-existing works only focus on continuous time controller design, which may yield degraded control performance when the control strategy is employed by a digital controller owing to the finite sampling frequency and the inevitable discrete-time approximations. To this end, this paper proposes a discrete-time sliding mode prescribed performance controller (DTSMPPC) via Kalman filter (KF) and disturbance observer (DO) for a pressurized water reactor (PWR), aiming to achieve an easier implementation on the practical nuclear reactor systems. Specifically, the continuous mathematical model of the PWR system with consideration of model errors and disturbances is first transformed to a discrete-time form by virtue of Euler's discretization technique. Then, based on this transformed discrete-time model, the KF and DO, which provide the estimates of unmeasured system states (relative density of delayed neutron precursor, average fuel temperature, xenon concentration, and iodine concentration) and uncertainties, respectively, are constructed using the input/output measurement information acquired from the nuclear reactor system merely. By incorporating the observed information provided by the KF and DO into the control framework, the proposed DTSMPPC enables the load following error to fulfill the accurate and robust performance requirement even in the presence of unmeasured system states and uncertainties, while at the same time ensuring the transient and steady-state load following error within a prescribed zone described by performance functions. The system stability and the aforementioned theoretical findings are proved through rigorous analysis and simulations. Simulation results also reveal that the proposed DTSMPPC via the KF and DO yields a superior load following performance compared with some previous control strategies. [ABSTRACT FROM AUTHOR]

The aim of this paper is to present a theoretical sub-critical wait-time experiment that may be used to validate stochastic nuclear reactor kinetics models which compute the neutron population probability distribution. The regulation governing many zero-power reactors prohibits super-critical nuclear reactor operation. This paper shows that a desirable wait-time probability distribution can be computed by only modifying the neutron source strength. An analytical model, which ignores the contribution of delayed neutron precursors, is presented. Additionally, a six-delayed neutron precursor group model using the Pál–Bell equations is used to investigate the relationship between the nuclear reactor system parameters and the resulting wait-time probability distribution. Finally, this paper shows that two neutron energy group models can be collapsed down to an equivalent single (one speed or mono-energetic) neutron energy group model which accurately reproduce the same wait-time probability distribution. [ABSTRACT FROM AUTHOR]

• Reactivity meter with uncertainty developed. • Three sets of experimental effective kinetics parameters considered. • Reactivity Uncertainties decrease considerably. • Very good agreement to Inhour Equation. • Results shown for a practical application considering IPEN(MB01)-LWR-RESR-015. Three sets of effective kinetics parameters were measured and evaluated for the fuel rod core of the IPEN/MB-01 research reactor. A correlation matrix among all these effective kinetics parameters is taking into account to reduce the inferred reactivity uncertainties. This work considers the propagation of the uncertainties inherent in these sets of effective delayed neutron parameters to the inferred reactivities. The comparisons to the reactivity and its uncertainty values produced by the Inhour Equation came into a very good agreement and reinforces the validity of the whole procedure for the development of this reactivity meter with uncertainties. The reduction of the reactivity uncertainty values compared to previous works is significant. The analyses of some specific cases of the evaluation IPEN(MB01)-LWR-RESR-015 reveal that its reactivity values are in accordance with those of this work. However, its reactivity uncertainty values for positive reactivities seems to cancel the uncorrelated component of the uncertainties. [ABSTRACT FROM AUTHOR]

The goal of this work is to develop a high-order numerical technique to solve a two-dimensional fractional neutron diffusion (FND) model describing dynamical behavior of a lead-cooled fast reactor. This method is based on L 1 technique for temporal discretization and a high-order compact Alternating Direction Implicit (ADI) finite difference scheme for the spatial discretization. Numerical simulations are conducted to demonstrate the applicability of the method. We calculate the neutronic flux in the core and examine how the order of the temporal-fractional derivative influences the neutronic flux. The results obtained with FND model for different values of anomalous diffusion coefficient are compared against those obtained with P1 model and classical neutron diffusion equation. In addition, the computational time (CPU time) is provided in order to justify the computational efficiency of the proposed method. • Numerical method for two-dimensional fractional neutron diffusion model. • Anomalous diffusion (non-Fickian) in lead-cooled fast reactors. • ALFRED nuclear reactor dynamic analysis for different coefficient of anomalous diffusion. • Comparisons with classical neutron diffusion theory (normal diffusion) and P1 equations. [ABSTRACT FROM AUTHOR]

Within the frame of the EU H2020 program SANDA project, sensitivity and uncertainty analyses have been performed for the ESFR, ASTRID and ALFRED reactor concepts and the multi-purpose flexible irradiation facility MYRRHA. Relevant reactor parameters, namely the effective multiplication factor, the effective delayed neutron fraction, the Doppler reactivity coefficient, the void worth and the worth of control rods have been investigated. SCALE, Serpent 2 and MCNP6.2+SUMMON codes together with the covariance data of the JEFF-3.3 nuclear data library have been used for this study. A ranking of the most important isotopes and reactions impacting the mentioned parameters has been derived for all these systems. Uncertainties have been quantified and have been found to be higher than the target accuracies proposed. Therefore, recommendations for improvement of nuclear data are given. [ABSTRACT FROM AUTHOR]

We review the current status of the development of sources of epithermal neutrons sources based on reactors and accelerators for boron neutron capture therapy (BNCT), a promising method of malignant tumor treatment. The scheme is proposed of the source prototype for the production of thermal and epithermal neutrons using the delayed neutrons generated with help of linear electron accelerator at the target containing the fissile material. The results of an experiment are presented in which the half-life curves of radioactive nuclei formed during fission and emitting delayed neutrons are measured. It is shown that an activated target containing fissile material is a compact small-sized source of delayed neutrons. It can be delivered to the shaper, where, using a moderator, an absorber, and a collimator, neutrons of thermal or epithermal energies are formed over a certain period of time, after which this target is sent to the activator, and another target comes in its place. Thus, a pulsed neutron flux is formed. Such a neutron beam can be used in nuclear medicine, in particular, in neutron capture therapy in the treatment of cancer. An important task in the implementation of neutron capture therapy, when irradiating patients, is to control both the intensity and the energy spectrum of the neutron flux. To solve this problem, an earlier developed activation-type neutron ball spectrometer can be used, which will allow optimization of various parameters of the shaper, collimator and filters in order to obtain the most powerful neutron fluxes.