Fusion Neutron Source Research Articles

Analytical study of low energy proton interactions in the SORGENTINA’s fusion ion source-part I: beam-dump

An analytical study, corroborated by Monte Carlo simulations, is presented which describes the interaction of a 300 keV–830 mA proton beam with a Cu and Al dumping system. The analytical calculations rely on the theoretical framework of the Particle Induced X-ray Emission, while the Monte Carlo simulations are performed by means of the GEANT4 toolkit. The case study is related to the project SORGENTINA-RF fusion neutron source and in particular to the tests devoted to assess the performance of the ion source of the plant. The results provide a detailed physical insight of the main processes occurring in the beam dump material, and are also important to give some realistic figures of the radiation emission expected during operation.

Neutron yield using a semi-Monte Carlo simulating method for D-T fusion neutron sources

Neutron yield and spatial distribution are crucial parameters of a gaseous target D-T neutron source. The exact calculation of these parameters in large-size geometries with complex fluid dynamics is very important for gaseous target neutron source design and subsequent applications. The gaseous target on operating conditions put more requirements on the Multiphysics simulating method. The fusion neutron yield simulating method has been developed using semi-Monte Carlo based on the finite element analysis for Multiphysics. The method has been verified using the model of a D-T fusion neutron source with a windowless gas target and compared with experimental results and the GEANT4 simulations. Compared with the experimental results, the simulated neutron yield error is 8.31%, and compared with the result of GEANT4, the relative error of total neutron yield is 3.34%. The spatial distributions of neutron yield are in good agreement with the results of GEANT4. In particular, the computing time is reduced from ∼200 core-hours with GEANT4 to ∼50 core-hours with the developed semi-Monte Carlo method, greatly improving the calculation efficiency.

Conceptual Neutronics Study of a Hybrid Fusion Neutron Source FNS-C with Aqueous Blanket

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 233U is 1.027 at a consumption of 1304 kg/year of the fissile material in the aqueous blanket containing 232Th enriched to 1.47% 233U. The FNS-C is operated with an effective neutron multiplication factor keff ~ 0.99 with reactivity ρ = –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 keff ~ 0.99 is ~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.

Study on A-FNS shielding analysis with variance reduction parameter generation code

Design activities of advanced fusion neutron source A-FNS have been conducted to perform the irradiation tests for fusion reactor materials. It is necessary to evaluate the effective dose rate of the lateral access cell in the A-FNS facility by shielding calculations. In the calculation, a variance reduction method using an effective bias data is necessary to reduce statistical errors. The weight window generation (WWG) function had been used in the analyses with MCNP code and McDeLicious source subroutine. There were some difficulties in the usage of WWG such as many iterative calculations, experiences, and skills required to obtain appropriate bias data. Therefore, the automatic variance reduction parameter generator code ADVANTG has been introduced to the calculation, creating a source description in the general source definition (SDEF) of MCNP for the A-FNS by approximating the neutron and photon generations of the McDeLicious source subroutine. The ADVANTG code was successfully applied to the calculation, and the dose rate of the lateral access cell in the A-FNS was obtained with small statistical errors by a once-through calculation.

Neutronics analysis of a subcritical blanket system driven by a gas dynamic trap-based fusion neutron source for 99Mo production

Neutronics analysis of a subcritical blanket system driven by a gas dynamic trap-based fusion neutron source for 99Mo production

Influence of electrodes' geometrical properties on the neutron production rate of a discharge fusion neutron source

Trapping ions, such as deuterium and tritium, inside a potential well to generate neutrons is a well-established technology through electric and magnetic fields via the inertial electrostatic confinement fusion (IECF) and the tokamak, respectively. In the IECF, the straightforward configuration is a concentric cathode connected to a negative bias, surrounded by a grounded anode that serves as a vacuum vessel. Theoretically, neutrons are generated inside the vessel through fusion between ions that are accelerated by applying several tens kV voltage and tens mA current. Many parameters affect the plasma conditions and fusion in the system, hence the neutron production rate (NPR). This study investigates the cathode transparency and the number of apertures effect on NPR. For this end, eleven cathodes were fabricated from stainless steel in three different groups with different transparency and number of apertures. NPRs were investigated as a function of the cathode transparency and number of apertures at low power operating mode ∼1 kW. Experimental results revealed that higher NPR was produced from lower grid transparency and vice versa; this behavior was explained through beam–surface fusion with grid surface. In addition, a higher NPR was generated from the grid with many apertures; this was attributed to the effect of the deuterium ionization improvement by the number of ionizing electrons through the grid channels.

Dissolution of Molybdenum in Hydrogen Peroxide: A Thermodynamic, Kinetic and Microscopic Study of a Green Process for 99mTc Production.

99mTc-based radiopharmaceuticals are the most commonly used medical radioactive tracers in nuclear medicine for diagnostic imaging. Due to the expected global shortage of 99Mo, the parent radionuclide from which 99mTc is produced, new production methods should be developed. The SORGENTINA-RF (SRF) project aims at developing a prototypical medium-intensity D-T 14-MeV fusion neutron source specifically designed for production of medical radioisotopes with a focus on 99Mo. The scope of this work was to develop an efficient, cost-effective and green procedure for dissolution of solid molybdenum in hydrogen peroxide solutions compatible for 99mTc production via the SRF neutron source. The dissolution process was extensively studied for two different target geometries: pellets and powder. The first showed better characteristics and properties for the dissolution procedure, and up to 100 g of pellets were successfully dissolved in 250-280 min. The dissolution mechanism on the pellets was investigated by means of scanning electron microscopy and energy-dispersive X-ray spectroscopy. After the procedure, sodium molybdate crystals were characterized via X-ray diffraction, Raman and infrared spectroscopy and the high purity of the compound was established by means of inductively coupled plasma mass spectroscopy. The study confirmed the feasibility of the procedure for production of 99mTc in SRF as it is very cost-effective, with minimal consumption of peroxide and controlled low temperature.

High deuteron and neutron yields from the interaction of a petawatt laser with a cryogenic deuterium jet

A compact high-flux, short-pulse neutron source would have applications from nuclear astrophysics to cancer therapy. Laser-driven neutron sources can achieve fluxes much higher than spallation and reactor neutron sources by reducing the volume and time in which the neutron-producing reactions occur by orders of magnitude. We report progress towards an efficient laser-driven neutron source in experiments with a cryogenic deuterium jet on the Texas Petawatt laser. Neutrons were produced both by laser-accelerated multi-MeV deuterons colliding with Be and mixed metallic catchers and by d (d,n)3He fusion reactions within the jet. We observed deuteron yields of 1013/shot in quasi-Maxwellian distributions carrying ∼8−10% of the input laser energy. We obtained neutron yields greater than 1010/shot and found indications of a deuteron-deuteron fusion neutron source with high peak flux (>1022 cm−2 s−1). The estimated fusion neutron yield in our experiment is one order of magnitude higher than any previous laser-induced dd fusion reaction. Though many technical challenges will have to be overcome to convert this proof-of-principle experiment into a consistent ultra-high flux neutron source, the neutron fluxes achieved here suggest laser-driven neutron sources can support laboratory study of the rapid neutron-capture process, which is otherwise thought to occur only in astrophysical sites such as core-collapse supernova, and binary neutron star mergers.

Physics-based model of irradiation creep for ferritic materials under fusion energy operation conditions

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈3×10−7 dpa/s at temperatures between 300 and 800 K.

Constructive Outline of a Vacuum Chamber of a Fusion Reactor or a Fusion Neutron Source

Constructive Outline of a Vacuum Chamber of a Fusion Reactor or a Fusion Neutron Source

Accurate Fe–He machine learning potential for studying He effects in BCC-Fe

Nanostructured Ferritic Alloys (NFAs) containing oxide nanoparticles are prospective advanced structural materials in future fusion reactors. Lacking fusion neutron sources with an adequate flux, modeling is critical to understand radiation damage in the NFAs, including helium effects. Machine learning interatomic potentials (MLPs) have been demonstrated to be capable of describing atomic interactions in chemically complex systems with comparable accuracy to density functional theory (DFT) but with much less computational costs. Hence, MLPs are chosen to study He bubble accumulation in chemically complex NFAs. As a preliminary step, we develop a Fe–He potential, based on ∼10,000 atomic configurations. The developed MLP accurately predicts the bulk properties of BCC-Fe compared to DFT, including the lattice constant (2.832 Å), elastic constants (c11=271 GPa, c12=141, c44=93), and phonon frequencies (maximum error < 3.5%). The MLP predicts He binding energies in HenV bubbles and Hen clusters with a mean absolute error (MAE) of only ∼70 meV, which is ∼3 to ∼5 times smaller than the MAE of binding energies estimated using empirical potentials. The MLP also accurately predicts the migration barrier of interstitial He (64 meV). Furthermore, the MLP correctly predicts the maximum number of He atoms (critical size) at which a HenV bubble (Hen cluster) can kick-out a self-interstitial atom (SIA), forming a bigger HenV2 bubble (HenV bubble). Subsequently, we use the MLP to perform molecular dynamics simulations to investigate the effect of temperature on the critical size of HenV bubbles and Hen clusters. We find that the critical size decreases with increasing temperature — indicating that smaller bubbles and clusters can kick-out an SIA at higher temperatures, thereby escalating the He bubble growth. The current Fe–He MLP can be further developed to include all chemical interactions in NFAs.

Neutral Beams for Neutron Generation in Fusion Neutron Sources

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.

On the expected performance of the SORGENTINA-RF fusion neutron source

A study devoted to assessing the neutronic performance of the SORGENTINA-RF fusion neutron source is here presented. The calculation is performed considering the characteristics of the accelerated mixed deuterium/tritium ion beam from the ion source, the transport of ions into the titanium layer that constitutes the ion beam target, their implantation and the fusion reaction cross section. An investigation has been done by varying the ion composition of the beam in order to assess if and to what extent diatomic and triatomic ion species affect the neutron emission rate of the plant. This calculation is also instrumental to evaluate the cost-effectiveness of equipping the ion source with a filtering system to remove secondary ions from the beam.

Impact of Neutral Beam and Plasma Shaping on the Current Drive Efficiency and Fast Ion Losses in a Compact Tokamak FNS-ST

Fusion neutron source (FNS)-spherical tokamak (ST) will be a FNS based on ST ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$R =0.5$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$a =0.3$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$k =2.75$ </tex-math></inline-formula> , and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$B =1.5$ </tex-math></inline-formula> 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.

Effects of Metal Hydride Coatings at the Electrodes on Neutron Production Rate in a Discharge-Type Fusion Neutron Source

A glow discharge (GD) fusion neutron source that utilizes nuclear fusion reactions of deuterium has been upgraded. The fusion reactions in this device mainly occur by collisions between the charged or neutral particles and the hydrogen isotopes trapped at the surface of electrodes. In addition, it is known that the metal hydride coating on the electrode enhances the neutron production rate (NPR). Therefore, the elemental distribution, including deuterium, in the depth direction on the electrode is an essential factor in neutron production. However, the distribution on the electrode has not been experimentally investigated. This study aims to analyze the distribution experimentally and indicate the effect of the metal hydride coatings. To achieve this purpose, we prepared the titanium (Ti)-coated cathode and the uncoated cathode, of which the base material was stainless steel. After that, the neutron production test was performed in the range of from 5- to 40-mA currents and from 20- to 60-kV applied voltage. This test indicated that the NPR was improved by coating the cathode with Ti than the uncoated cathode. In addition, depth profiling on the cathodes by glow discharge optical emission spectroscopy (GD-OES) was performed. While the analysis indicated that the concentration of deuterium on both cathodes was increased after the test, there was no significant difference in the concentration of deuterium between both cathodes. Furthermore, the concentration of Ti on the Ti-coated cathode was vastly decreased. The cause of these changes needs to be investigated.

On the definition of the deuterium-tritium ion beam parameters for the SORGENTINA-RF fusion neutron source

The SORGENTINA-RF project aims to develop a 14 MeV fusion neutron source to produce medical radioisotopes with special focus on ^{99}Mo. The facility is based on a positive ion source with an acceleration stage to produce a deuterium (D^{+}) and tritium (T^{+}) ion beam that, impacting on a titanium-coated rotating target, allows fusion reactions to take place. Maximizing the neutron production rate is one of the main issues to be addressed in the project and the optimization of some key parameters of the ion beam is of paramount importance in this regard. In this study, a methodology is discussed to reach a definition of the beam characteristics for an effective and sustainable operation of the plant. The most convenient layout that has been found out is based on a single ion source fed by a deuterium and tritium gas mixture. Eventually, a series of considerations about the operation of the ion source and fuel cycle have been drawn.

Choice of Gas Isotope Composition for Neutral Beam Injectors of the FNS-ST Compact Fusion Neutron Source

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 ne, 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 Т beam injection, in the operating range of parameters, the neutron yield can reach 6.0 × 1017 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 × 1017 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 Tinv of ~170 g. If this approach is used in the case of the deuterium beam, the neutron yield can reach 7.0 × 1017 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.”

Neutral Beam Coupling with Plasma in a Compact Fusion Neutron Source

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 &gt; 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.

Preliminary Assessment of Radiolysis for the Cooling Water System in the Rotating Target of SORGENTINA-RF

The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission rate in the order of 5–7 × 1013 s−1. The plant relies on a metallic water-cooled rotating target and a deuterium (50%) and tritium (50%) ion beam. Beyond the main focus of medical radioisotope production, the source may represent a multi-purpose neutron facility by implementing a series of neutron-based techniques. Among the different engineering and technological issues to be addressed, the production of incondensable gases and corrosion product into the rotating target deserves a dedicated investigation. In this study, a preliminary analysis is carried out, considering the general layout of the target and the present choice of the target material.

Measuring the near-target neutron field of a D–D fusion facility with the novel NCT-WES spectrometer

A new directional neutron spectrometer called NCT-WES (Neutron Capture Therapy Wide Energy Spectrometer) was recently developed within the INFN ENTER_BNCT project. This device constitutes a more compact and portable alternative to state-of-the-art directional neutron spectrometers. The device was used for the first time to characterise the near-target field produced by an accelerator-driven D–D fusion neutron source. NCT-WES operates as a “parallelised” Bonner spheres spectrometer, embedding six semiconductor-based thermal neutron detectors in a cylindrical moderator. Owing on a cylindrical collimating aperture, the device exhibits sharply directional response. To account for the non-uniform irradiation condition experienced in the near-target field, a dedicated NCT-WES response matrix was developed. The neutron spectrum at 0^{circ } from the D–D neutron target, determined by means of the FRUIT unfolding code, is coherent with that previously derived with Bonner spheres. NCT-WES proved to be a promising device for angular spectrometric characterisation of neutron-emitting targets.