This study presents the results of testing nuclear data libraries by analyzing statistical criteria obtained from comparing experimental and calculated rates for (n,2n), (n,p), (n,pn), (n,n'gamma) (n,alpha) and (n,gamma) reactions measured on samples Ni-nat, Zr-nat, Nb-nat, Cd-nat, Ti-nat, Co-nat,Cu-63(96%),Cu- 65(99.70%), Zn-64(99.70%), In-nat, Al-nat, Mg-nat, Fe-nat, Au-nat and Th-nat, which were placed in the experimental channels of micromodels of the fusion blanket. The "fast" (the cylinder & Oslash; 230 mm and 520 mm length was filled with similar to 67 kg of molten salt 0.52NaF + 0.48ZrF4) and the "thermal" blanket (the same cylinder was placed in a dry channel inside a cubic container filled with water with dimensions of 52.0 x 52.0 x 52.0 cm were investigated. The reaction rates were measured using the activation method. Modeling with transport codes MCNP5, KIR, PHITS-3.31, SuperMC3.4.0 was performed using the ENDF/B-VII.0 library for neutron transport as well as seven neutron data libraries for reaction rates simulation, including: JEFF-3.3, JENDL-4.0, ENDF/B-VIII.0, ROSFOND-2010, FENDL-3.0, TENDL - 2019 and IRDFF-II.

Some aspects of the use of coatings for various functional purposes on the first wall plasma-facing surface of a thermonuclear reactor are considered. An important characteristic of coatings is adhesive and fatigue strength under cyclic impact of quasi-stationary heat loads, as well as resistance to high pulsed thermal loads. This paper describes thermal cyclic tests with surface thermal load of water-cooled mockups with various coatings on the heat-receiving surface. The B4C coating, made by atmospheric plasma spraying on a tungsten substrate, demonstrated excellent durability over 1400 thermal cycles at 4.7 MW/m2. CVD tungsten coating on a copper substrate demonstrated good results after 1000 thermal cycles at 3.3 MW/m2, but after a similar number of cycles at 5 MW/m2, cracks were detected on the surface. The stainless steel coating on a copper substrate demonstrated resistance to loads up to 11.9 MW/m2, as well as excellent durability over 1000 thermal cycles at 8.2 MW/m2.

The organization of the plasma–wall interaction remains an urgent problem for long-term operation of a tokamak with an intense thermonuclear fusion reaction. The concept of a lithium cycle and the design of a sectioned divertor for the Demonstration Fusion Neutron Source (DEMO-FNS) tokamak are proposed. The parameters are estimated, and requirements for the components of the lithium cycle are formulated. Technical solutions for the lithium cycle flow rate ≅10 g/s are selected. It is estimated that 0.1-µm liquid lithium layer on the surface of the first wall can protect its solid coating. On the basis of a simple model, it is shown that, at a wall temperature of 200–300°C, a thickness of 0.1 µm can be achieved in ≅1 min. The film can reach quasi-stationary values of 13–15 µm in 3–4 h. Above 340°C, the film does not form because of the increase in the thermal evaporation of lithium. The wall temperature of 700°C of the divertor section with the lithium pool is chosen so that lithium deuteride and lithium tritide do not form in it. They can form in the liquid metal lithium protective layer of the wall at temperatures less than 300°C. In order to significantly reduce the explosion and fire hazard when working with hot liquid lithium, it is proposed to increase the size of the DEMO-FNS divertor section with the lithium pool by 2 to 3 times, which may allow the transition from water to helium coolant.

In the design of the DEMO fusion reactor, as well as the facility of a fusion neutron source FNS, as one of the options, it is assumed that a three-loop cooling system will use a blanket of liquid metals. Liquid lithium from the primary loop transfers heat in intermediate heat exchangers to liquid sodium of the second loop, which in turn transfers heat to the water of the third loop in the steam generator. The possibility is considered of using a “neutral” coolant in the second loop—a eutectic lead–bismuth alloy, which does not chemically interact with either liquid lithium or water in the third loop, which makes it possible to exclude contact of liquid lithium or sodium with water. A preliminary calculation of the intermediate liquid lithium–eutectic alloy heat exchanger has been performed, and it is shown that the use of this alloy will allow for the cooling of the fusion reactor with the same number of intermediate heat exchangers and without a significant increase in their size.

Steady-state operation of fusion neutron sources (FNSs) will be maintained by a non-inductive current drive (CD) from injected high-energy particles and bootstrap current. Neutral beam injection (NBI) is an efficient current driver and a main source of neutrons in FNS plasma with NBI contribution expected to reach 60%-99% in total fusion intensity. The overall NBI performance in the FNS machine can be represented as a product of the injector beamline efficiency, the beam capture in plasma, and the effects produced by the beam particles. NB output in plasma can be expressed as a direct CD and/or beam-thermal fusion gain. The main feature of the presented methodology is that the injector and NB effects in plasma are optimized consistently: transmission losses along the injector beamline, beam capture by plasma, and its final performance in tokamak operation are included in the optimization procedure with a goal to obtain maximum effect in beam-plasma scenarios. Direct coupling of models allows one to perform a combined analysis of CD efficiency and neutron output for a given NB source exit power. The integrated model reproduces the beam spatial and angular distribution with a maximum level of geometrical accuracy while being highly flexible and efficient as compared to conventional NBI models. The injector multichannel structure, scrappers, and transmission duct geometry are replicated in detail; power losses along the beamline including direct particle interception are calculated with an account of electromagnetic fields, gas flows, and technological tolerances. NBI performance in a tokamak is calculated using the beam-plasma suit of codes developed in 2018-2022. Plasma magnetic shape is described by aspect ratio, elongation, and triangularity parameters. An efficient modeling technique allows one to obtain fast particle statistics and loss channels-along the injector beamline and within the plasma volume-in a few seconds on moderate laptops. The range of optimum parameters for beam-plasma scenarios with maximum NBI efficiency in FNS tokamak is limited by established NB production technology.

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international C oordination on I nternational C hallenges on L ong duration OP eration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

The results obtained by determining the flux density of neutrons produced with energies of up to 20 MeV upon the irradiation of a beryllium target 1.3 mm thick with a beam of 21.3-MeV protons are presented. The proton flux density was determined by means of standard instruments and was controlled with the aid of the monitoring reactions ^nat Cu (p,x)^62 Zn and ^nat Cu (p,x)^63 Zn, while the neutron flux density was determined using the reactions ^27 Al (n,p)^27 Mg and ^27 Al (n,α)^24 Na. The proton and neutron spectra at the center of experimental samples were calculated using the PHITS code.

This paper presents the results of the experimental determination and computational simulation of the ambient dose equivalent rate for a metallic thorium cylindrical miniblock and the (n,2n), (n,f), and (n,gamma) reaction rates in a thin Th-232 metal foil irradiated with neutrons of the NG-24M generator spectrum. The ambient dose equivalent rate was determined by dosimeters-radiometers. The reaction rates were determined by the activation method using Ge spectrometers without destroying the irradiated samples. Computational simulations of ambient dose equivalent and reaction rates were performed, respectively, using the radiation transport codes PHITS, MCNP5, and KIR2, which use various nuclear data libraries: JEFF-3.2 and -3.3; JENDL4.0; ENDF/B-VII.0, -VII.1, and -VIII.0; ROSFOND; FENDL; and TENDL. The authors give an estimate of the U-232/U-233 relative accumulation upon natural thorium irradiation in a fusion facility blanket with defined neutron spectrum. The nonirradiated and irradiated thorium nuclide composition change simulation and visualization were performed using analytical solutions of an ordinary system of homogeneous linear differential equations describing nuclide transmutations.

The paper is devoted to the creation of an improved neutron plasma model of the hybrid reactor facility (HRF) DEMO-FNS, which has been used by codes implementing the Monte Carlo method for modeling radiation transport (MCNP and SuperMC). The old simplified model of the neutron source did not take into account the effect of the neutral injection on the intensity distribution of the fusion reaction rate, and it also had an unrealistic shape (ellipsoidal in vertical section). A new model has been developed on the basis of current data of temperature and density distributions in plasma. The model is built on the basis of current ideas about temperature and density distributions in plasma, takes into account the high contribution of neutral injection to the formation of a population of fast ions, and is also specified by an array of elementary volumes. This makes it possible to set an arbitrary shape of the neutron generation intensity distribution with the required accuracy. The influence of the plasma shape on the distribution of the primary (not scattered) neutron flux along the poloidal angle, as well as on neutron transfer and related processes outside the plasma (in the blanket, and first wall), is shown.

The dependence of the neutron yield of the FNS-ST (spherical tokamak) fusion neutron source on the fraction of tritium in the core D+T plasma is analyzed for the case of using tritium neutral beam injectors with 200-keV energy and 6-MW power. The FNS-ST operating regimes are explored using the SOLPS4.3 and ASTRA codes for different values of core plasma density n(e), T fraction in the plasma, and particle diffusivity. The FC-FNS code is used to estimate the fluxes of the fuel components in the fuel cycle (FC), which are produced by different injection systems: gas puffing, pellet injection, and neutral beam (T) injection. It is shown that in the case of the T beam injection, in the operating range of parameters, the neutron yield can reach 6.0 x 10(17) s(-1), which is the value comparable to that obtained for the scenario of D-beam injection into the balanced D+T plasma. In the case of the T-beam injection, in the range of parameters, for which the neutron yield is close to its maximum, the amount of tritium in the FC is lower than in the case of the D-beam injection. The neutron yield can be increased to 6.5 x 10(17) n/s if full separation of the D and T is introduced for the gas pumped out from the divertor and puffed back into the torus. With this approach, in the case of the tritium beam, the amount of tritium in the FC is T-inv of similar to 170 g. If this approach is used in the case of the deuterium beam, the neutron yield can reach 7.0 x 10(17) n/s. However, in this case, the amount of tritium contained in the FC increases to 215 g. The results of the analysis performed are used for optimizing the FC of the FNS-C (compact) fusion neutron source, which is planned for construction in the framework of the comprehensive program of the State Corporation Rosatom "Development of Engineering, Technology and Scientific Research in the Field of Using Atomic Energy in the Russian Federation for the Time Period up to 2030."

The paper presents several modifications of a vacuum chamber of the designed fusion neutron source (DEMO-FNS), which will improve its reliability and safety. Employment of a "double" internal shell makes it possible to decrease the temperature and thermal stresses in it. Modifying the outer shell design makes it possible to decrease the thermal load on cryogenic thermal shielding. Changing the water flow scheme inside the vacuum chamber body ensures reliable cooling of iron-water shielding in it. The method of thermal calculations is described. The proposed modifications will be used in other fusion designs.

The thermonuclear fusion between fast (super-thermal) particles injected in plasma as a neutral beam and the ions of the background plasma is expected to be the main source of fusion neutrons in FNS (fusion neutron source) design based on tokamak. Neutral beam contribution in fusion reactivity and in the total neutron yield depends on the high-energy ion fraction in the integral energy distribution. NESTOR code [1] calculates nuclear fusion rates in the FNS plasma volume, taking into account an external source of high-energy fast ions. Neutral beam model reproduces in detail the actual beam structure in phase space at the injection port plane; while the fast ion distributions in magnetically confined plasma are calculated using a combination of slowing-down classical formulae and magnetic field topology in the tokamak chamber. Here we discuss the issues relevant to the overall neutron production and the contribution of fast ions to the neutron output in plasma.

The modern challenges of nuclear energy are the replenishment of dwindling reserves of nuclear fuel and the development of a closed nuclear fuel cycle while complying with strict radiation safety requirements. A fusion neutron source has unique capabilities to solve these problems. The preliminary results of a neutronic analysis of the FNS-C fusion-fission hybrid neutron source with a thorium-uranium aqueous blanket by the Monte Carlo method computer simulation, using the MCNP-4 code with the ENDF/B-VII cross-section library, gives satisfactory results for the study of the possibility of creating a compact source of fusion neutrons based on a small spherical tokamak for commercial use. The obtained results show that the FNS-C hybrid blanket generates enough tritium to fully ensure the uninterrupted operation of the FNS-C throughout the year. The reproduction coefficient of U-233 is 1.027 at a consumption of 1304 kg/year of the fissile material in the aqueous blanket containing Th-232 enriched to 1.47% U-233. The FNS-C is operated with an effective neutron multiplication factor k(eff) similar to 0.99 with reactivity rho = -0.006249 in the presence of delayed neutrons, which corresponds to the safest state of the core of thermal neutron fission reactors. The thermal power of the FNS-C at k(eff) similar to 0.99 is similar to 3 GW, which is comparable to the thermal power of fission reactors. This indicates the potential possibility of creating a safe thorium-uranium breeder power reactor based on a fusion neutron source. The results of the study were obtained for the simplified approximate geometrical FNS-C model. To confirm the preliminary results, it is necessary to develop a more accurate calculation model of the FNS-C machine.

Neutral beam injection is supposed to be the main source of high-energy particles, driving non-inductive current and generating primary neutrons in fusion neutron sources design based on tokamaks. Numerical simulation of high-energy particles’ thermalization in plasma and fusion neutron emission is calculated by novel dedicated software (NESTOR code). The neutral beam is reproduced statistically by up to 109 injected particles. The beam efficiency and contribution to primary neutron generation is shown to be dependent on the injection energy, input current, and plasma temperature profile. A beam-driven plasma operation scenario, specific for FNS design, enables the fusion rate and neutron generation in plasma volume to be controlled by the beam parameters; the resultant primary neutron yield can be efficiently boosted in plasma maintained at a relatively low temperature when compared to ‘pure’ fusion reactors. NESTOR results are applicable to high-precision nuclear and power balance estimations, neutron power loads distribution among tokamak components, tritium generation in hybrid reactors, and for many other tasks critical for FNS design.

FNS-ST is a fusion neutron source project based on a spherical tokamak (R/a = 0.5 m/0.3 m) with a steady-state neutron generation of ~1018 n/s. Neutral beam injection (NBI) is supposed to maintain steady-state operation, non-inductive current drive and neutron production in FNS-ST plasma. In a low aspect ratio device, the toroidal magnetic field shape is not optimal for fast ions confinement in plasma, and the toroidal effects are more pronounced compared to the conventional tokamak design (with R/a > 2.5). The neutral beam production and the tokamak plasma response to NBI were efficiently modeled by a specialized beam-plasma software package BTR-BTOR, which allowed fast optimization of the neutral beam transport and evolution within the injector unit, as well as the parametric study of NBI induced effects in plasma. The “Lite neutral beam model” (LNB) implements a statistical beam description in 6-dimensional phase space (106–1010 particles), and the beam particle conversions are organized as a data flow pipeline. This parametric study of FNS-ST tokamak is focused on the beam-plasma coupling issue. The main result of the study is a method to achieve steady-state current drive and fusion controllability in beam-driven toroidal plasmas. LNB methods can be also applied to NBI design for conventional tokamaks.

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.

Fusion neutron source (FNS)-spherical tokamak (ST) will be a FNS based on ST ( $R =0.5$ m, $a =0.3$ m, $k =2.75$ , and $B =1.5$ T) with a steady-state neutron yield ~1018 n/s. The noninductive current drive and neutron generation is to be maintained by neutral beam injection (NBI). Beam in toroids (BTOR) code (Python) is used to reproduce NBI and plasma shaping with high degree of detail and to perform parametric studies of neutral beam (NB) driven effects. The beam model light neutral beam (LNB) is inherited from the original beam transmission (BTR) code, used since 2005 for NBI design and optimizations. The BTOR methodology combines NB statistical description with high-performance analytical models of particle tracing, which allows one to get results several orders faster, if compared to the conventional (Monte–Carlo) approach, and still with a good agreement. The results obtained for the FNS-ST device prove that the entire beam efficiency highly depends on the beam–plasma mutual size and shaping, including the beam cross section and aiming, plasma aspect ratio, elongation, and triangularity. The results are especially important for a steady-state current drive and fusion control in low aspect ratio tokamaks; the BTOR method is also applicable to a conventional tokamak design.

To study the effect of the beam–plasma system geometry on current drive and generation of neutrons in plasma, an efficient model was used that combines the statistical description of the injected beam with the analytical methods for calculating particle trajectories. The calculation results were consistent with the classical models of interaction of neutral beams with plasma. The model presented is simple and high-performance that makes it possible to trace up to 1012 fast ions in plasma and quickly optimize the facility parameters with allowance for the plasma and beam geometry. It is shown that the dimensions and inner angular structure of the atomic beam considerably affect the process of fast ions capture in plasma and overall efficiency of neutral beam injection (NBI), including the current drive efficiency and generation rate of fusion neutrons during the interaction of fast ions with plasma ions. An example is presented of optimizing the NBI parameters for the FNS-C fusion neutron source based on the spherical tokamak. The current drive efficiency (NBI‑driven current) and neutron yield of the beam–plasma fusion reactions were chosen as target criteria for the system optimization.

This paper presents the results of an experiment determining (n,2n), (n,p), (n,pn), (n,alpha), (n,n'gamma), and (n,gamma) reaction rates in 15 test samples of both natural and high-enriched composition: Mg-nat, Al-27, Ti-nat, Fe-nat, Co-59, Ni-nat, Cu-63 (99.5%), Cu-65 (99.7%), Zn-64 (99.4%), Zr-nat, Nb-93, Cd-nat, In-nat, Tm-169, and Au-197. Computer simulations in the NG-24M neutron generator spectrum were carried out using the MCNP5 and KIR2 radiation transport codes with different nuclear data libraries (JEFF-3.2, JEFF-3.3, JENDL-4.0, ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, ROSFOND-2010, FENDL-3.0, TENDL-2019, and IRDFF-II). The elaborated full-scale model for neutron transport analysis included the geometry and composition of the neutron generator, experimental samples, and laboratory room. The mean square deviation factor was used to compare the experimental and the simulated results. The best predictive results for both the MCNP5 code and the KIR2 code were obtained with the FENDL-3.0 and ENDF/B-VIII.0 libraries.

The design of the demonstration fusion neutron source (DEMO-FNS) demonstration fusion–fission hybrid facility is carried out at the National Research Center ‘Kurchatov Institute’ in Russia. For the implementation of the DEMO-FNS project, it is important to assess the effect of the fast neutron spectrum on radiation-induced damage to the device materials, primarily the materials of the first wall, which is the most problematic unit of the device. Due to the lack of real hybrid fusion devices, the study can only be carried out using computer simulations of the experiment. A computer simulation using the Monte Carlo method was carried out to calculate the developed three-dimensional full-scale model of the DEMO-FNS reactor with a blanket for the transmutation of minor actinides. The MCNP-4 code was used with cross-sections from the FENDL-2.1 and ENDF/B-6 files, as well as with cross sections for calculating radiation displacements. The neutron spectrum in the first wall of DEMO-FNS was determined at the power of a volumetric deuterium–tritium neutron of 1.42 × 1019 n s−1. It was found that beryllium as a plasma-facing material has too short lifetime under the neutron load on the first wall and divertor plates of DEMO-FNS. The replacement beryllium to tungsten is problematic and requires special study. Copper alloys have sufficient resistance to the effects of neutron radiation of DEMO-FNS for one full power year, but will require several replacements when operating the device for more than 10 years. The steels used in fast reactors can meet the operational requirements of the first wall of DEMO-FNS. To solve the problem of choosing materials for the first wall of DEMO-FNS, new experimental researches on changes in the physical properties of these materials in the fast neutron spectrum are needed.