Fusion Reactor Concept Research Articles

Hydrogen and deuterium permeation in Hastelloy N

Hastelloy N was chosen as the fluoride salt-contacting structural material for the Molten Salt Reactor Experiment due to its excellent compatibility with the fuel salt FLiBe. FLiBe is currently investigated for several advanced fusion and fission reactor concepts where tritium generation in the FLiBe is anticipated. Knowledge of hydrogen transport properties through Hastelloy N is important to understand how tritium would permeate through this material and result in an unintentional release. In this study, the hydrogen and deuterium permeability, diffusivity, and solubility were measured from 500 to 700 °C at a primary-side pressure of 10 kPa in a well-characterized sample of Hastelloy N. The prepared polycrystalline Hastelloy N had C and O impurities present on the surface. These impurities were investigated using Auger Emission Spectroscopy and Ar depth profiling. The adventitious C was removed upon the first Ar sputter cycle whereas O persisted deeper into the sample. For permeation experiments, applied deuterium pressures ranged from 13 Pa to 100 kPa and deuterium transport remained in the diffusion-limited regime (J ∝ P0.5) throughout the pressure range examined. Two methods are employed to measure the effective hydrogen and deuterium diffusivity: rise and decline. The decline method produced improved statistical model fits for calculating the effective diffusion coefficient compared to the rise method. The resultant transport properties compared well to published values for other nickel alloys.

The Princeton Field-Reversed Configuration for Compact Nuclear Fusion Power Plants

The Princeton Field-Reversed Configuration (PFRC) nuclear fusion reactor concept is an innovative approach to fusion power generation prioritizing low neutron production and small size. A combination of analytical modeling and numerical simulation shows that the novel heating approach generates an FRC with closed field lines. Simulation data from a single-particle Hamiltonian code predicts ms-scale plasma heating in reactor-scale conditions while PIC codes predict formation of warm FRC plasmas from initial mirror fields. The PFRC-1 and PFRC-2 experiments have heated electrons to energies well in excess of 100 eV and plasma durations to 300 ms, more than 10\(^4\) times longer than the predicted tilt instability growth time. From these data, we have created a development plan and anticipated performance metrics for a fusion reactor based on the PFRC concept. The resulting 1–10 MW PFRC reactors would be suitable for diverse applications, from submarines to urban environments to space propulsion. PFRC is a steady-state, driven magnetic confinement device. Plasma, inside a cylindrical array of coils, is confined and heated by external RF antennae. PFRC would be ultra-low radiation due to both its fuel and small size. The choice of advanced fuels, deuterium and helium-3 (D–3He), may be enabled by the high-\(\beta\) FRC configuration. The small size of the reactor would enable rapid exhaust of the dangerous tritium ash. Low radiation would make the reactor safer to operate and, in combination with simple geometry and small size, dramatically lowers development and maintenance costs. This review paper gives an introduction to the physics of the PFRC and a summary of the PFRC-2 experiment results to date. It then discusses the future program plan and how PFRC reactors would be commercialized.

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

AR reactor neutronics multi-code validation* *This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow

The affordable, robust, compact (ARC) reactor is a tokamak fusion reactor concept currently under development by Commonwealth Fusion Systems and Massachusetts Institute of Technology. There are three important neutronics considerations for the operation of the ARC reactor: (1) breeding of enough tritium in the blanket to sustain the D–T reaction in the plasma; (2) ensuring low fluence on the superconducting toroidal field coils; and (3) assessing neutron volumetric heating in structural components. This work aims to perform a validation of the neutronics analysis approach by code-to-code comparison. State-of-the-art software stacks are employed for the neutronics analysis of the ARC reactor, and a computer-aided design (CAD) model is used directly for Monte Carlo (MC) neutron transport calculations. Three software stacks, Attila-MCNP, OpenMC-DAGMC, and Shift-DAGMC, are used to perform neutronic analyses of a 90° sector CAD model of the ARC reactor. Results show that the flux tallies calculated by the three software stacks are very close. Also, the volumetric heating and tritium breeding values have less than 0.6% relative difference between codes.

Development of an object-oriented, thermal-hydraulics model for ARC FLiBe loop safety assessment

The development of new fusion reactor concepts is a challenging task for designers and safety analysts. Little information is usually available from previous experiences or experimental campaigns, especially when the design is innovative. Hence, system modeling is of fundamental importance. A proper system model should allow the investigation of different design options, and a preliminary assessment of the relative safety features. Anticipated or accidental transients can be studied, exploring the design space and highlighting possible criticalities. Object-oriented modeling is extremely advantageous to carry out this task. ARC pre-conceptual design (MIT-PSFC) may greatly benefit from this kind of analysis. A 1-D system-level, thermal-hydraulics model of ARC FLiBe loop developed in Modelica language is presented in this work. Because of the innovative nature of ARC design, most of the components are not available in the Modelica standard library. Thus, the key components of the loop are defined and modeled. No experimental results are currently available for model validation; therefore, the model consistency is assessed by verification and benchmark against analytical and numerical models. A Python wrapper is developed to explore multiple transient conditions by automating pre-processing and post-processing. Component failures are injected in the thermal-hydraulics model by a Monte Carlo routine. It is found that the model can efficiently describe different transients, with a low error on key parameters (pressure drop, fluid temperature). Furthermore, the model can be easily adapted to different design, thanks to the modular structure of object-oriented models. Similarly, it can be implemented in broader applications for safety analysis by coupling with suitable soft computing techniques.

ARC reactor neutronics multi-code validation

Abstract The affordable, robust, compact (ARC) reactor is a tokamak fusion reactor concept currently under development by Commonwealth Fusion Systems (CFS)and Massachusetts Institute of Technology. There are three important neutronics considerations for the operation of the ARC reactor: (1) breeding of enough tritium in the blanket to sustain the D-T reaction in the plasma; (2) ensuring low fluence on the superconducting toroidal field coils; and (3) assessing neutron volumetric heating in structural components. This work aims to perform a validation of the neutronics analysis approach by code-to-code comparison. State-of-the-art software stacks are employed for the neutronics analysis of the ARC reactor, and a computer-aided design (CAD) model is used directly for Monte Carlo (MC) neutron transport calculations. Three software stacks, Attila-MCNP, OpenMC-DAGMC, and Shift-DAGMC, are used to perform neutronic analysesof a 90°sector CAD model of the ARC reactor. Results show that the flux tallies calculated by the three software stacks are very close. Also, the volumetric heatingand tritium breeding values have less than 0.6% relative difference between codes.

A study on ohmic plasma initiation for JA DEMO

We have investigated ohmic plasma initiation for JA DEMO, a design concept of the tokamak demonstration fusion reactor examined in Japan. A calculation model describing the tokamak plasma initiation phase consisting of the two-dimensional analysis of the coil and eddy currents and the zero-dimensional plasma analysis is developed. It is revealed that the conducting shell does not necessarily hinder plasma initiation, while it provides the stabilizing effect of vertical displacement events. The available loop voltage and prefill gas pressure for initiation in JA DEMO are clarified for developing the initiation scenario. We demonstrate a possible scenario of ohmic plasma initiation, which is consistent with the design specifications of superconducting poloidal field coils.

Impact of blanket material and configuration on the size of a tokamak fusion reactor

Abstract On the basis of the Korean fusion roadmap, conceptual design of the DEMO and RD the HCLL concept among liquid blanket concepts does as well, due to the good shielding capability of PbLi. The outcomes of this study provide useful guidelines for the feasibility assessment and selection for the blanket concept of the tokamak fusion reactor.

Cost analysis and energy return on investment of fuel cell and gas turbine integrated fusion-biomass hybrid system; application of a small scale conceptual fusion reactor GNOME

The objective of this study was to propose a fusion-biomass hybrid system with fuel cell and gas turbine as a technical option for fusion application to generate electricity. Fusion reactor with dual coolants lithium lead (DCLL) blanket design employing SiCf/SiC could provide over 900 °C of high temperature heat for waste biomass gasification process. Waste biomass such as municipal waste solid and residues from agricultural and forestry sites could be sufficiently processed. A small scale conceptual fusion reactor GNOME is designed under the assumption that the system will be deployed in a place where a low amount of waste biomass is generated. Results showed that 364 MW GNOME reactor could produce 410 MWth power through divertor and blanket according to fusion power plant system analysis code (FUSAC). The hybrid system enables to generate 454 MWe with 80% capacity factor which is 7008 h operation in a year. Net electricity of 289 MWe could be produced through solid oxide fuel cell (SOFC) and gas turbine showed 30% of overall system efficiency due to the high self-consumption of GNOME reactor. The levelized cost of electricity (LCOE) is 208 $/MWh and the energy return on investment (EROI) is 3.9.

Effect of collisionality dependence of thermal force on impurity transport under a lower collisional condition in DEMO scrape-off layer plasma

A preceding study (Homma et al 2018 Contrib. Plasma Phys. 58 629–37) presented an advanced model for thermal force whose applicable condition is extended from collisional to relatively low collisional plasma. According to this extended model, the thermal force reduces when plasma collisionality decreases. In the present study, the extended thermal force model has been implemented into a SOL-divertor integrated simulation code SONIC, in order to study the impact on impurity transport in DEMO-relevant scrape-off layer (SOL) plasma, due to thermal force reduction. A set of test simulations has been carried out supposing a reference steady-state operation scenario of the Japanese DEMO fusion reactor concept (JA DEMO). The thermal force has reduced by as much as 20%–70%, reflecting relatively lower collisionality in DEMO SOL plasma. The simulation results have demonstrated that the introduction of the collisionality dependence of thermal force leads to as much as 20%–80% of effective decrease in impurity density and its content ratio widely over the SOL upstream area under the DEMO relevant condition. Compared to the case with the conventional thermal force model, relative change rate of impurity content at representative poloidal positions are as follows: low-field side (LFS) X-point −59%, poloidal top area −22%, HFS upstream area up to −80%, and no major impact around LFS upstream.

General Relativistic Modification of the Lockheed Compact Fusion Reactor Concept

Taking into account Einstein’s general theory of relativity, a modification of Lockheed’s compact fusion reactor concept is proposed by replacing the two superconducting tori with rapidly rotating tori rotating in opposite directions. According to the general theory of relativity, two Coriolis force fields in opposite directions are set up, both of them having a negative mass density in their corotating reference systems, with a vanishing negative mass density in the center in between the rotating tori, where the hot fusion plasma is centered. Because of the Nernst effect going in the opposite direction, large toroidal currents are set up, repelling the hot plasma from the much cooler tori. This results in closed magnetic field lines for stable plasma confinement. The remaining problem, the removal of the heat released by neutron absorption in the metallic tori, can be resolved by a pulsed operation, axially injecting cool deuterium-tritium gas, from which the heat is externally removed by a radiator.

A novel fusion reactor with chain reactions for proton–boron11

AbstractUsing a combination of laser–plasma interactions and magnetic confinement configurations, a conceptual fusion reactor is proposed in this paper. Our reactor consists of the following: (1) A background plasma of boron11 and hydrogen ions, plus electrons, is generated and kept for a certain time, with densities of the order of a mg/cm3 and temperatures of tens of eV. Both the radiation level and the plasma thermal pressure are thus very low. (2) A plasma channel is induced in a solid target by irradiation with a high power laser that creates a very intense shock wave. This mechanism conveys the acceleration of protons in the laser direction. The mechanisms must be tuned for the protons to reach a kinetic energy of 300–1200 keV where the pB11 fusion cross section is significantly large (note that this value is not a temperature). (3) Those ultra-fast protons enter the background plasma and collide with boron11 to produce three alphas. Fusion born alphas collide with protons of the plasma and accelerate them causing a chain reaction. (4) A combination of an induction current and a magnetic bottle keeps the chain reaction process going on, for a pulse long enough to get a high energy gain. (5) Materials for the background plasma and the laser target must be replaced for starting a new chain reaction cycle.

Modeling and numerical analysis of a magneto-inertial fusion concept with the target created through FRC merging

The Magnetized Plasma Fusion Reactor concept is proposed as a magneto-inertial fusion approach based on the target plasma created through the collision merging of two oppositely translating field reversed configuration plasmas, which is then compressed by the imploding liner driven by the pulsed-power driver. The target creation process is described by a two-dimensional magnetohydrodynamics model, resulting in the typical target parameters. The implosion process and the fusion reaction are modeled by a simple zero-dimensional model, taking into account the alpha particle heating and the bremsstrahlung radiation loss. The compression on the target can be 2D cylindrical or 2.4D with the additive axial contraction taken into account. The dynamics of the liner compression and fusion burning are simulated and the optimum fusion gain and the associated target parameters are predicted. The scientific breakeven could be achieved at the optimized conditions.

Direct-drive inertial confinement fusion: A review

The direct-drive, laser-based approach to inertial confinement fusion (ICF) is reviewed from its inception following the demonstration of the first laser to its implementation on the present generation of high-power lasers. The review focuses on the evolution of scientific understanding gained from target-physics experiments in many areas, identifying problems that were demonstrated and the solutions implemented. The review starts with the basic understanding of laser–plasma interactions that was obtained before the declassification of laser-induced compression in the early 1970s and continues with the compression experiments using infrared lasers in the late 1970s that produced thermonuclear neutrons. The problem of suprathermal electrons and the target preheat that they caused, associated with the infrared laser wavelength, led to lasers being built after 1980 to operate at shorter wavelengths, especially 0.35 μm—the third harmonic of the Nd:glass laser—and 0.248 μm (the KrF gas laser). The main physics areas relevant to direct drive are reviewed. The primary absorption mechanism at short wavelengths is classical inverse bremsstrahlung. Nonuniformities imprinted on the target by laser irradiation have been addressed by the development of a number of beam-smoothing techniques and imprint-mitigation strategies. The effects of hydrodynamic instabilities are mitigated by a combination of imprint reduction and target designs that minimize the instability growth rates. Several coronal plasma physics processes are reviewed. The two-plasmon–decay instability, stimulated Brillouin scattering (together with cross-beam energy transfer), and (possibly) stimulated Raman scattering are identified as potential concerns, placing constraints on the laser intensities used in target designs, while other processes (self-focusing and filamentation, the parametric decay instability, and magnetic fields), once considered important, are now of lesser concern for mainline direct-drive target concepts. Filamentation is largely suppressed by beam smoothing. Thermal transport modeling, important to the interpretation of experiments and to target design, has been found to be nonlocal in nature. Advances in shock timing and equation-of-state measurements relevant to direct-drive ICF are reported. Room-temperature implosions have provided an increased understanding of the importance of stability and uniformity. The evolution of cryogenic implosion capabilities, leading to an extensive series carried out on the 60-beam OMEGA laser [Boehly et al., Opt. Commun. 133, 495 (1997)], is reviewed together with major advances in cryogenic target formation. A polar-drive concept has been developed that will enable direct-drive–ignition experiments to be performed on the National Ignition Facility [Haynam et al., Appl. Opt. 46(16), 3276 (2007)]. The advantages offered by the alternative approaches of fast ignition and shock ignition and the issues associated with these concepts are described. The lessons learned from target-physics and implosion experiments are taken into account in ignition and high-gain target designs for laser wavelengths of 1/3 μm and 1/4 μm. Substantial advances in direct-drive inertial fusion reactor concepts are reviewed. Overall, the progress in scientific understanding over the past five decades has been enormous, to the point that inertial fusion energy using direct drive shows significant promise as a future environmentally attractive energy source.

Creep Lifetime Analysis on First Wall Structure Material for CFETR

China fusion engineering test reactor is an ITER-like superconducting tokamak reactor, which has a big duty time factor. Lifetime of structure material in blanket play an important role in the safety, reliability and economy of fusion reactor, which should be careful considered in the engineering design. High temperature is always one of the main targets in the fusion reactor concept design. And creep performance under high temperature of structure material is important in the failure mechanisms and lifetime analysis. Creep lifetime analysis of first wall in helium cooled solid breeder blanket was conducted for candidate structural materials. It is found that 316L(N) have a better creep lifetime performance than Eurofer. For an acceptable creep lifetime at 0.3 MW/m2 surface heat flux, there could be space for wall thickness change and improvement of cooling ability and outlet temperature of first wall.

The science program of the TCV tokamak: exploring fusion reactor and power plant concepts

TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic electron cyclotron resonance heating (ECRH) to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was employed to control the current density profile using electron cyclotron current drive. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace. Systematic studies of suprathermal electrons and ions in the presence of ECRH were performed. The L–H threshold power was measured to be ∼50–75% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core. The geodesic acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested successfully. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. The edge gradients were found to govern the scaling of confinement with current, power, density and triangularity. The dynamical interplay of confinement and magnetohydrodynamic modes leading to the density limit in TCV was documented. The heat flux profile decay lengths and heat load profile on the wall were documented in limited plasmas. In the snowflake (SF) divertor configuration the heat flux profiles were documented on all four strike points. SF simulations with the EMC3-EIRENE code, including the physics of the secondary separatrix, underestimate the flux to the secondary strike points, possibly resulting from steady-state E × B drifts. With neon injection, radiation in a SF was 15% higher than in a conventional divertor. The novel triple-null and X-divertor configurations were also achieved in TCV.

Designing Radiation Resistance in Materials for Fusion Energy

Proposed fusion and advanced (Generation IV) fission energy systems require high-performance materials capable of satisfactory operation up to neutron damage levels approaching 200 atomic displacements per atom with large amounts of transmutant hydrogen and helium isotopes. After a brief overview of fusion reactor concepts and radiation effects phenomena in structural and functional (nonstructural) materials, three fundamental options for designing radiation resistance are outlined: Utilize matrix phases with inherent radiation tolerance, select materials in which vacancies are immobile at the design operating temperatures, or engineer materials with high sink densities for point defect recombination. Environmental and safety considerations impose several additional restrictions on potential materials systems, but reduced-activation ferritic/martensitic steels (including thermomechanically treated and oxide dispersion–strengthened options) and silicon carbide ceramic composites emerge as robust structural materials options. Materials modeling (including computational thermodynamics) and advanced manufacturing methods are poised to exert a major impact in the next ten years.

Oxidation/Corrosion Behaviour of ODS Ferritic/Martensitic Steels in Pb Melt at Elevated Temperature

Lead-based melts (Pb, Pb-Bi) are considered as candidate coolants and spallation neutron targets due to their excellent thermophysical and nuclear properties. However, the corrosion of structural materials remains a major issue. Oxide dispersion strengthened (ODS) ferritic/martensitic steels are considered for high temperature application for both fission and fusion reactor concepts. The oxidation/corrosion kinetics in a static oxygen-saturated Pb melt at temperature of 550°C as well as the morphology and composition of scales formed on ferritic/martensitic Fe-9Cr-1.5W and ferritic Fe-14Cr-1.5W ODS steels have been investigated. Both materials showed homogeneous multiple, dense scales that consisted of typical combination of Fe3O4 as outer sublayer and (Fe,Cr)3O4 as inner sublayer. A nonuniform growth of inner oxide sublayers into the metal matrix as well as a good adhesion to the metal substrate is observed. With the prolongation of exposure from 240 to 1000 h, observed scales grow from 35 µm to 45 µm for ODS Fe-9Cr steel and from 40 µm to 60 µm for ODS Fe-14Cr steel with the thinning rates of 0,22 and 0,31 mm/year correspondingly. The mechanism of scales formation is discussed.

Nuclear Data from AMS & Nuclear Data for AMS – some examples

We summarize some recent cross-section measurements using accelerator mass spectrometry (AMS). AMS represents an ultra-sensitive technique for measuring a limited, but steadily increasing number of longer-lived radionuclides. This method implies a two-step procedure with sample activation and subsequent AMS measurement. Applications include nuclear astrophysics, nuclear technology (nuclear fusion, nuclear fission and advanced reactor concepts and radiation dose estimations). A series of additional applications involves cosmogenic radionuclides in environmental, geological and extraterrestrial studies. There is a lack of information for a list of nuclides, as pointed out by nuclear data requests. An overview of some recent measurements is given and the method is illustrated for some specific neutron-induced reactions.

Self-consistent Calculation of Neutronic E ects in the Development of Fusion Reactor Concept

Self-consistent Calculation of Neutronic E ects in the Development of Fusion Reactor Concept