ДВУМЕРНАЯ ТРАНСПОРТНАЯ МОДЕЛЬ НЕЙТРАЛОВ НА ОСНОВЕ ВЕРОЯТНОСТЕЙ ПРОХОДА И УХОДА ПОТОКОВ С
We determine coil arrangements for reproducing a minimum-B mirror magnetic field, optimized with respect to plasma stability, plasma cross-section ellipticity and particle drift surfaces.The reproduction has to be done with precision, as field errors may give rise to plasma instabilities or collisionless plasma losses due to the guiding centres' drift away from the confinement region.We have developed a set of twisted «fishbone» coils to allow an array of coils to be flexibly stacked, as required for a precise magnetic field reproduction.Results suggest that high mirror ratios of around 10 can be obtained using a fishbone coil arrangement.The mirror ratio can be further increased by finite plasma beta.Parameters representative of a compact 10 MW fusion neutron source have been derived.
A large amount of spent nuclear fuel (SNF) from nuclear power plants has been accumulated globally to date, and there is still no established strategy for handling it.While SNF can be partitioned, the predominant isotope Uranium-238 can be used to produce secondary fuel in fast nuclear reactors, and plutonium be burned in thermal nuclear reactors as a part of MOX fuel.Fission products can be disposed in geological repositories, as they decay in 200-300 yearsmuch sooner than SNF.A major challenge is to handle minor actinides (MAs), particularly americium and curium, which are long-lived elements and are currently not recycled.They have different nuclear properties and cannot be treated like plutonium.It is possible to have americium and curium effectively burned up (fissioned) through irradiation with fusion neutrons.This paper explores the idea of employing fusion power plants for recycling those elements.An appropriate model was generated, which used americium and curium quantities small enough to avoid any strong impact on the reactor systems and operation.At the same time, the model allowed for high MA burnup rates.Nuclear facility used in the model was a torus-shaped thermonuclear reactor with plasma major and minor radii of 1000 and 300 cm, respectively.Such facility could take up additional 10 t of fuel (americium plus curium) with no significant impact on its physical characteristics.The americium and curium burnup rates, calculated with the MNCPX code, were within acceptable limits.Fission neutrons were found to contribute to the production of tritium, which may be important from the standpoint of the reactor's self-sufficiency in tritium supply.Calculations proved that the reactivity of the reactor as a fission burner was low, enabling a safe operation.In addition to the MA incineration and tritium breeding capacities, fission reactions provided for a moderate (tens of percent) power gain.
Based on the heuristically generalized perturbation theory (HGPT) adapted to subcritical systems [1], a procedure for online operational monitoring of the subcriticality level of a hybrid fusion-fission system in which an "external" neutron source is generated in a magnetically confined tokamak-type plasma is described.This procedure [2], commonly referred to as power control-based subcriticality monitoring (PCSM), consists of compensating of slow and small movements of a specialized control rod in the nuclear fission zone, previously calibrated using a standard procedure, with equally slow and small alterations of the fusion neutron source.PCSM is verified by solving the multigroup transport equation for the direct flux, as well as for the importance function, associated with normalized fission power [3], and a technique is proposed for modifying the fusion neutron source strength based on plasma compression/expansion.Plasma and confining magnetic field adjustments, needed to implement the PCSM, are estimated with due account of the 0-D-plasma power balance.
In connection with the start-up of the T-15MD tokamak, testing of numerous magnetic probes of the electromagnetic diagnostic system is an important problem, because measurements from magnetic probes are used to control the plasma discharge. The article analyzes the first experiments on measuring the magnetic field performed on the T-15MD tokamak. A program has been created in which Kirchhoff equations for currents in poloidal coils and eddy currents in a vacuum chamber and passive stabilization conductors are solved. The distribution of the magnetic field and voltage in the magnetic probes are calculated and compared with experimentally measured values. As a result, it is found that some probes have incorrect connection polarity and that the signals from two probes are mixed up during recording to the database. It is shown that large currents are induced in the vacuum chamber, which will influence the plasma equilibrium, especially at the initial stage of current rise. An algorithm is proposed to improve the 2D model of the T-15MD chamber, which increases the accuracy of calculating the eddy currents and thereby improves plasma discharge control.
The choice of an idea for a divertor with evaporating liquid lithium that meets the requirements for removing the thermal load from the edge plasma and provides an acceptable level of change in the ionic composition of the main plasma for the DEMO-FNS tokamak being developed in Russia has been discussed. The results of numerical simulation and optimization of the design of divertors with multiple volumes sectioned by slotted diaphragms have been presented. The parameters of lithium streams flowing into the edge layer have been estimated for the temperature range of divertor chambers from 500 to 1000 K under the conditions of the gas-kinetic and free-molecular modes of lithium vapor outflow from the divertor. Analysis of the processes that reduce the outflux of lithium from the chambers and its penetration into the main volume of the plasma inside the separatrix showed that sectioning effectively reduces the outflow streams to acceptable levels of ≈10 20 atom/s.
This paper presents the high flux neutron shielding design and extensive neutronics calculations of GDT based fusion neutron source ALIANCE. Neutron distribution of ALIANCE is strongly inhomogeneous along the axis: significant portion of the neutron flux is generated near the two mirrors, while the rest of it is spread over the remaining central volume of plasma. The shielding design includes 40 cm stainless steel as the main shielding layer and an additional 5 cm tungsten carbide shielding layer at mirror plugs to protect superconducting coils from neutron damage and reduce nuclear heating. The simulations have been carried out by using Monte Carlo transport code SuperMC with nuclear data library FENDL 3.1. Results show that the nuclear heating on the mirror coils can be reduced by more than two thirds with additional tungsten carbide shield, and fast neutron fluence by 30 %. The highest nuclear heating and the highest fast neutron fluence zones are located at the mirror coils, and the values are about 300 W/m(3) and 9 x 10(18) n/cm(2) respectively, which meets the threshold of ITER superconducting coils. The specific activities of shielding layers are of order of 10(12) Bq/kg. The structural materials? specific activities will decrease to 4 x 10(11) Bq/kg in one year after shutdown, and their decay heat will quickly drop below 2 kW/m(3) after one day. Besides, all the structural materials of ALIANCE can be recycled by different recycling technologies. The modeling and calculations reported in this paper will be beneficial for the pre-conceptual engineering design of ALIANCE.
меняться, причём направление изменения зависит от исходного значения этой скорости.Сверхзвуковой поток ускоряется
An important part of high-temperature plasma study is the determination of the electron temperature dynamics in the tokamak plasma. At spherical tokamaks, one can use Thomson scattering diagnostics as well as soft X-ray emission diagnostics (SXR). The capabilities of electron temperature measurement by the first diagnostics are limited by the repetition rate of laser pulses and their number in one tokamak discharge. Data of the second diagnostics are continuous in time and are determined by the time resolution of the detectors; however, obtaining the electron temperature using these data encounters a number of difficulties considered in this study. A method of combined processing of results of these diagnostics using machine learning algorithms was developed for overcoming these difficulties and applying the adVoprosy Atomnoi Nauki i Tekhniki, Seriya: Termoyadernyi Sintezages of both diagnostics. Training data include soft X-ray diagnostic data, hard X-ray diagnostic data, and CIII line emissivity diagnostic data. Thomson local scattering measurements were used as labels for supervised machine learning. The developed technique provides significant extension of the possibilities of determining the electron temperature at the Globus-M2 tokamak.
The method of neutron calculations for the diagnostic equipment of the ITER fusion reactor has been considered. The important role of these calculations in the design of systems used in the environments with radiation exposure has been noted. The process of developing diagnostic equipment and modeling neutronic characteristics for installations with the source of radiation in the form of thermonuclear plasma has been described. The analysis of the modern software used to determine the neutron parameters of various installations, including for ITER, has been carried out. Particular attention has been paid to defining the concept of statistical calculation error and ways to reduce the variance of calculated values. The results of the computational modeling of gamma and neutron fields, radiation energy release, and damage to materials in the area of the diagnostic system of spectroscopy of hydrogen lines (H-alpha) located in the equatorial port no. 11 of ITER have been presented.
The T-15MD tokamak with the warm toroidal field coils, elongated plasma cross section, and a low aspect ratio was commissioned at the National Research Center Kurchatov Institute in 2021. An economically viable fusion reactor requires long discharges or even a steady-state mode. The short pulse of the toroidal field is among the main obstacles on the way to the steady state. Therefore, as a next step after T-15MD, it is proposed at the NRC Kurchatov Institute to consider the design of a superconducting tokamak (SCT), which holds the basic geometric parameters of the T-15MD: R/a = 1.5 m/0.67 m with a field on the axis of B-0 < 5 T and a long pulse of current of I-p < 5 MA. The focus of the paper is the system of the toroidal magnetic field. A three-layer design of the toroidal field coil is proposed: an inner layer of a high-temperature superconductor (HTSC), a Nb3Sn middle layer, and a NbTi outer layer. On the basis of structural strength calculations, candidate materials for coil cases were selected. For cooling the coils, a semi-longitudinal pumping of low-pressure liquid helium is proposed. The design of the HTSC cable in conduit consisting of two halves is considered in detail. A mesh cryostat design that provides a convenient access for diagnostics and plasma heating is proposed. SCT systems similar to those used in the T-15MD are analyzed.
Electron cyclotron resonance heating is one of the methods of additional plasma heating in the T-15MD tokamak. The article considers the possibility of plasma heating using an extraordinary wave at the third harmonic of electron cyclotron resonance at a frequency of 140 GHz. It has been shown that, in accordance with the calculations by the quasi-linear OGRAY code, a high fraction of single-pass absorption (up to 96%) can be expected even when the ECR is turned on with an ohmic background. Increase in this fraction with the increase in plasma temperature agrees with the theoretical predictions. Vertical power launch seems to be the most attractive. Possible engineering solutions for implementation of such a scheme are proposed.
ЛУЧЕВАЯ МОДЕЛЬ ПУЧКА ДЛЯ ОПТИМИЗАЦИИ ПАРАМЕТРОВ НЕЙТРАЛЬНОЙ ИНЖЕКЦИИЕ.Д.Длугач, Б.В
The description of the zero-dimensional engineering-physical code GLOBSYS (Globus spherical tokamak system code), designed for parametric analysis of the next step of the program Globus-M, Globus-M2, is given. Within the framework of the zero-dimensional approximation, the definitions of the main scaling parameters of the plasma (poloidal beta, the fraction of bootstrap current, the energy lifetime of the plasma), as well as the specifics of calculating the inductance and resistance of the plasma in spherical tokamaks, are refined. The results of calculations of the plasma parameters by the code were compared with the experimental data of one of the Globus-M2 discharges (no. 38800) with neutral beam heating and showed good agreement. It is proposed to perform a comparison of calculations based on the code with the achieved and predicted parameters of the spherical tokamaks NSTX, NSTX-U, MAST, MAST-U, and ST40 in a separate paper. The goals of the next step (Globus-3) are formulated, the main ones of which are long pulse, high toroidal field, and powerful heating, which allow us to consider Globus-3 as a hydrogen prototype of a neutron source. The infrastructural restrictions on the Globus-3 parameters are given, which require further analysis of various versions of the electromagnetic system. Using the example of Globus-M2 discharge no. 38800, the effect of restrictions on the flow balance and heating of the elements of the electromagnetic system is shown.
The calculation of the sink strengths of vacancy voids for radiation defects, which are the parameters of phenomenological models of radiation damage of materials, requires knowledge of the energy of interaction of the radiation defects with the elastic fields created by vacancy voids in the bulk of the material. Direct calculation of the interaction energy by molecular statics demands enormous computational resources and therefore is not suitable for sink strength calculations. In this article, we propose a computationally efficient approach to calculating the interaction energy, which does not introduce a significant error in the calculations. This approach is based on the combined use of different methods: molecular statics is used to calculate the dipole tensors of radiation defects and the elastic strain fields created by vacancy voids, while the interaction of voids with radiation defects (elastic dipoles) is calculated using anisotropic linear elasticity theory. The validity of such an approach is demonstrated by directly comparing its results with the results obtained only by the method of molecular statics, which uses as a test problem the calculation of the interaction between spherical vacancy voids with diameters of 2 and 20 lattice parameters and self-point defects for the BCC metal Fe. Elastic strain fields of the spherical vacancy voids with diameters from 2 to 20 lattice parameters in the BCC metals Fe and V are calculated by molecular statics.
This paper comprises results presented at the FUNFI4 conference.The reported research focuses on the prospective utilization of a fusion-fission hybrid system (FFHS) as a powerful neutron source capable of transmuting minor actinides (MAs; Np, Am, Cm) from spent nuclear fuel (SNF).Calculations simulating nuclide kinetics in MA-bearing metal fuel were performed for three 40-MW fusion power FFHS reactors, intended for different purposes (demonstration, pilot-industrial, and industrial).In addition, the research assessed potential requirements for FFHSs and their role in Russia's nuclear power program.A model created by AO «Proryv» was used to analyze the development of the Russian nuclear power system with integrated FFHSs.MA quantities expected to be produced and transmuted in the integration scenario were estimated.The results suggest that just one hybrid facility's capacity would be enough to achieve a ~28% MA decrease in the Russian power system by 2130.
Neutron sources are the important experimental platforms for the R&D of advanced nuclear energy and nuclear technology application.The High Intensity Neutron Generator (HINEG) has been developed in China with different missions including neutronics design validation, material & components irradiation test, nuclear waste burning and nuclear technology application.HINEG-I has achieved the fusion neutrons with the yield of 6.4•10 12 n/s at maximum, and has been coupled with the Lead-based Zero Power Critical/Subcritical Reactor named CLEAR-0.Such a facility is actually an accelerator-driven Fusion-Fission Hybrid System.Series of typical experiments have been carried out on HINEG-I, including neutronics and code validation, core physics study of advanced reactors, neutron radiography, neutron detector calibration, neutron biological effects, neutron radiation hardening, and so on.HINEG-II is an accelerator-based neutron source with the yield of 10 14 -10 15 n/s.It aims to apply to multi-purposes, e.g.neutron capture therapy, isotope production, etc.The design and R&D for key technologies of HINEG-II are performed on-going.HINEG-III is initially conceived as a GDTbased or accelerator-based neutron source with the intensity of 10 17 -10 18 n/s.The objectives of HINEG-III are to conduct test of nuclear materials, components test and reliability data collection of nuclear components, nuclear waste burning test, etc.This contribution presents an overview of the series recent activities.
This article discusses the issue of selection of material for resistive current leads of superconducting magnets. A wide range of their working modes has been considered. It has been demonstrated that copper is by far not the best material for fabrication of current leads, especially operating at currents exceeding the optimum. In such modes, they are prone to rapid overheating, whereas current leads from alloys can withstand current loads exceeding optimum current by several times. Moreover, in the absence of current, the heat gains from current leads fabricated from alloys are 50% lower than those from copper, which is their obvious advantage. The physical reason for this effect is a strong temperature dependence of copper specific resistance, which leads to positive feedback in the course of increase in temperature of current leads. In addition, the current leads fabricated from alloys have other advantages: significantly higher time of their overheating, as well as wider opportunities for increase in the cooling surface.
