A Computational Study of a Capillary Discharge Pellet Accelerator Concept for Magnetic Fusion Fueling

An ablation-dominated capillary discharge using low atomic number elements for plasma formation to flow into an ablation-free extension barrel is a concept that provides a high energy–density plasma flow sufficient to propel fuel pellets into the tokamak fusion plasma chamber. In this concept, the extension barrel is made from a non-ablating material by coating the interior wall of the barrel with nanocrystalline diamond to eliminate mixing the propelling plasma with any impurities evolving from the barrel ablation. The electrothermal plasma code ETFLOW models the plasma formation and flow in the capillary discharge and the flow into the extension barrel to accelerate frozen deuterium pellets. The code includes governing equations for both the capillary and the extension barrel, with the addition of the pellet’s terms. It also includes ideal and non-ideal plasma conductivity models. The joule heating term in the energy conservation equation is only valid in the capillary section. The pellet momentum and kinetic energy are included in the governing equations of the barrel, with the addition of the effect of viscous drag terms. The electrothermal capillary source generates the plasma via the ablation of a sleeve inside the main capillary housing. The acceleration of the pellet starts in the extension barrel when the pressure of the plasma flow from the capillary reaches the release limit. The code results show pellet exit velocities in excess of 2 km/s for source/barrel systems with low-Z liner materials in the source for 5, 20, 45, and 80 mg pellets. The study shows that an increase in the length of both the source and the extension barrel increases the pellet exit velocity with the limitation of slowdown effects for plasma expansion and cooling off inside the barrel.     

Enhanced Performance of Electrothermal Plasma Sources as Fusion Pellet Injection Drivers and Space Based Mini-Thrusters via Extension of a Flattop Discharge Current

Electrothermal plasma sources have been introduced as a method to propel frozen hydrogenic pellets for fueling of future magnetic fusion reactors. These sources are also useful as mini-thrusters in space shuttles, pre-injectors in hypervelocity launchers and igniters in electrothermal-chemical Guns. The source is a capillary discharge that generates the plasma from the ablation of a liner in an ablation-dominated regime, or from the flow of gas into the capillary in an ablation-free regime. Most electrothermal plasma sources uses pulse power delivery system with a pulse length in the range of 100 μs with FWHM of 50 μs. This research is a computational study on the effect of extending the top of the discharge current pulse to the range of 1,000 μs on the source exit parameter to achieve higher pressures and better exit velocities. Calculations using 0.4 cm diameter, 9.0 cm length Lexan polycarbonate capillary source, using ideal and nonideal plasma models, show that extended flattop pulses at fixed amplitude produce more ablated mass which scales linearly with increased pulse length, however, other plasma parameters remain almost constant. Results suggest that quasi-steady state operation of an electrothermal plasma source may provide constant exit pressure and velocity for pellet injectors for future magnetic fusion reactors deep fueling.     

Flow Characteristics and Charge Exchange Effects in a Two-Dimensional Model of Electrothermal Plasma Discharges

Electrothermal (ET) plasma discharges are capillary discharges that ablate liner materials and form partially ionized plasma. ET plasma discharges are generated by driving current pulses through a capillary source with peak currents on the order of tens of kA and pulse lengths on the order of \(100\,\upmu \hbox {s}\). These plasma discharges can be used to propel pellets into magnetic confinement fusion devices for deep fueling of the fusion reaction, ELM mitigation, and thermal quench of the fusion plasma. ET plasma discharges have been studied using 0D, 1D, and semi-2D fluid models. In this work, a fully 2D model of ET plasma discharges is presented. The newly developed model and code resolve inter-species interaction forces due to elastic collisions. These forces affect the plasma flow field in the source and impede the development of plasma pressure at the exit of the source. In this work, these affects are observed for discharge current pulses peaking at 10 and 20 kA. The sensitivity of the model to the inclusion of charge exchange effects is observed. The inclusion of charge exchange has little effect on the integrated, global results of the simulation. The difference in total ablated mass for the simulations caused by the inclusion of charge exchange reactions is <1 %. Differences in local plasma parameters are observed during discharge initialization, but after initialization, these differences diminish. The physical reasoning for this is discussed and recommendations are made for future modeling efforts.     

Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different ⁶Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.     

C<Subscript>60</Subscript>-Fullerene Hyper-Velocity High-Density Plasma Jets for MIF and Disruption Mitigation

We present an innovative idea to use hyper-velocity (>30 km/s) high-density (>10¹⁷ cm⁻³) plasma jets of D-T/H and C₆₀-fullerene for magneto-inertial fusion (MIF), high energy density laboratory plasma (HEDLP), and disruption mitigation in magnetic fusion plasma devices. The mass (~1–2 g) of sublimated C₆₀ and hydrogen (or D-T fuel) produced in a pulsed power source is ionized and accelerated as a plasma slug in a coaxial plasma accelerator. For MIF/HEDLP we propose to create a magnetized plasma target by injecting two high-Mach number high-density jets with fuel (D-T) and liner (C₆₀/C) structure along the axis of a pulsed magnetic mirror. The magnetized target fusion (MTF) plasma created by head-on collision and stagnation of jets is compressed radially by a metallic liner (Z-pinch) and axially by the C₆₀/C liner. For disruption mitigation, the C₆₀ plasma jets were shown to be able to provide the required impurity mass (J Fusion Energy 27:6, 2008).     

Calcuiational Method for Neutron Heating in Dense Plasma Systems, (II)

Abstract

The effect of neutron heating on the burn characteristics of inertial confinement fusion pellets is investigated by applying the calcuiational method developed in an earlier paper (Part I). The basic equations are time-dependent transport equations for fusion neutrons and recoil ions, being written in terms of the modified Eulerian coordinates originally proposed by Wienke (1974). After incorporating these equations into the one-dimensional hydrodynamics code MEDUSA, burn simulations are made for isobaric D-T pellets models compressed to 1,000 times the liquid density. It is found that in reactor-grade pellets, the inclusion of neutron heating decreases the fuel gain from the values obtained by neglecting the neutron heating. Calculations neglecting the energy transport by recoil ions overestimate the neutron energy deposition to plasma ions. The energy spectrum of neutrons emitted out of a typical D-T burning pellet is also shown. These neutrons contain fast components whose energies reach more than 20MeV.     

Rview of pellet fueling

The injection of frozen pellets composed of the isotopes of hydrogen has become the leading candidate for refueling fusion power reactors based on the tokamak concept. This lofty position has been reached partly as a result of efforts to find an attractive solution to the perplexing problem of depositing atoms of fuel deep within the magnetically confined, hot plasma, and because of some recent experimental successes. To some extent, the relative merits of this technique will depend upon the distance that the cryogenic pellet will penetrate such a plasma, and the early exploratory research has addressed this problem on both theoretical and experimental fronts. The conclusion from the theoretical effort is that a protective blanket consisting of hydrogenic gas or cold plasma will envelope the pellet and partially shield the surface from the intense plasma heat flux. The blanket prolongs pellet lifetime, but penetration to the plasma center might require pellet injection velocities in excess of 10 km/s. The need for central penetration has not yet been established either theoretically or experimentally. The experiments performed to date have verified the existence of a shielding mechanism in general, and pellet ablation models that incorporate neutral gas shielding in particular are in adequate agreement with the experiments. Magnetic shielding effects are expected to contribute to, but not dominate, self-shielding in the higher plasma temperature regimes of the future. The tokamak plasma has demonstrated a surprising resilience even to massive density perturbations caused by the large refueling pellets used in present experiments. The characteristic discharge behavior is qualitatively not unlike that observed with gas puffing; but, for the first time, central plasma fueling has been studied, and this does not appear to be superior to refueling by partial pellet penetration. If relatively large pellets containing a significant fraction of the total plasma charge are acceptable in the present resistive plasma regimes, then it can be argued that they should have little impact on the gross stability of a hot thermonuclear tokamak plasma. Large pellets are preferable from the standpoint of attaining deep penetration, and this has important implications for the technology of pellet injection. The interesting velocity regime of 1 km/s has already been achieved with simple gun-type devices and this should be adequate for near-term tokamak experiments. Further improvements are anticipated, but the 10 km/s and above regime is uncertain; and, if current theory and experiments extrapolate to the future, such velocities might be desirable but unnecessary.     

Online integrated visual inspection and sorting system for fast reactor fuels

Mixed oxide (MOX) fuel for prototype fast breeder reactor (PFBR) is designed to have initial burn up of 100,000 MWD/T. The major differences from thermal reactor fuel are relatively smaller dimension with central hole and higher plutonium concentration (21% and 28% of PuO₂) MOX pellets which are loaded into 2.5 m long clad tubes with depleted UO₂ blanket pellets at either end of the MOX stack. The relatively smaller dimension of fuel pellets for PFBR results in large volume at fabrication and inspection. To ensure fast and accurate inspection and sorting of as sintered pellets with less radiation exposure to personnel an integrated on line pellet inspection system for remote visual inspection and sorting of pellets based on diameter has been developed. Details of the integrated pellet inspection system developed at Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur along with the results of the performance trials has been described in this paper.     

Railgun pellet injection system for fusion experimental devices

A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s⁻¹ using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s⁻¹ using a 3 m long railgun.     

Development of the pellet injection system on the J-TEXT tokamak

Pellet injection is an attractive technology for core-fueling and magnetohydrodynamic study in magnetic-confinement fusion devices like tokamaks and stellarators. It can inject solid hydrogen/deuterium pellets into the plasma with deeper density deposition compared with other fueling methods, such as gas puffing. A three-barrel H₂ pellet injection system was installed on the J-TEXT tokamak and experiments were carried out. The pellets are formed in three barrels cooled by a cryocooler and compressor system at around 9 K, and are 0.8 mm/1 mm diameter and 0.8 mm length. The pellet is launched by helium propellant gas and injected from the low-field side of the plasma. The normal range of pellet speed is 210–310 m s⁻¹ for different propellant gas pressures. Due to the three-barrel structure, the number of injected pellets can be adjusted between one and three. Pellets can be launched sequentially with arbitrary time intervals, which enables flexible applications. The results of the experiments show that pellet fueling efficiency can reach 50%. The energy confinement time increased by about 7.5‒10 ms after pellet injection.     

Neutronics Calculations in Pellets and Blankets of Laser Fusion Reactor Concept SENRI-I

The neutronic properties of SENRI-I, a reference design of laser fusion reactor proposed by Institute of Engineering, Osaka University, are discussed on the basis of the one-dimensional neutron transport calculations in burning DT plasmas and blankets. The softening of the fusion neutron energy spectrum, the neutron heating and the neutron multiplication are studied and discussed for the compressed DT pellets with various thickness of fuel plasmas and lead or lead-polyethylene tampers.

The neutronic and thermal features in the blanket of the SENRI-I design are also examined. The tritium breeding ratio is high enough (~1.6), depending on the neutron energy spectrum from a pellet. The maximum temperature increase per 1,000 MJ DT fusion reactions is ~3°C in the inner liquid Li layer and ~1.5°C in the stainless steel first wall. A parametric study is also presented on the effect of varying the thickness of the inner Li blanket ΔRⁱ to examine the thickness required for the enough tritium breeding ratio and energy deposition.     

Uranium-zirconium hydride fuel properties

Properties of the two-phase hydride U_0.3ZrH_1.6 pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650 °C (the design limit recommended by the fuel developer is 750 °C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3× that of oxide fuel) requires an initial radial fuel-cladding gap of ∼300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrH_x, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and ∼1.70 at the periphery. Because the density of the hydride decreases with increasing H/Zr ratio, the result of H redistribution is to subject the interior of the pellet to a tensile stress while the outside of the pellet is placed in compression. The resulting stress at the pellet periphery is sufficient to overcome the tensile stress due to thermal expansion in the temperature gradient and to prevent radial cracking that is a characteristic of oxide fuel. Several mechanisms for reduction of the H/Zr ratio during irradiation are identified. The first is transfer of impurity oxygen in the fuel from Zr to rare-earth oxide fission products. The second is the formation of metal hydrides by these same fission products. The third is by loss to the plenum as H₂.The review of the fabrication method for the hydride fuel suggests that its production, even on a large scale, may be significantly higher than the cost of oxide fuel fabrication.     

Interaction of Impurity （Li, Be, B and C） and Hydrogen Isotope Pellet Injection with Reactor-relevant Plasmas

Based on the two-dimensional kinetic ablation theory of the hydrogen pellet ablation developed by Kuteev [B.V. Kuteev, Nuclear Fusion, 35 (1995) 431], an algorithm of erosion speed and ablation rate calculations for Li, Be, and B impurity pellets in reactor-relevant plasma has been derived. Results show compatibilities of lithium pellet injection used in α-particle diagnostics are positive in comparison with other solid impurity pellets (e.g. Be, B and C). Using the 2-D Kuteev lentil model, including kinetic effects, we find that currently existing pellet injection techniques will not meet core-fueling requirements for ITER-FEAT. A pressure as high as 254 MPa must be applied to a pellet accelerator with a 200 cm-long single-stage pneumatic gun, in order to accelerate a pellet with a radius rpo = 0.5 cm to a velocity of vpo, 24 × 105 cm/s penetrating 100 cm into the ITER plasma core. Comparisons of pellet velocity- and radius-dependent penetration depth between the Neutral Gas Shielding and the Kuteev's     

Elastic and viscoplastic pellets fragmentation modeling using an axisymmetrical 1D finite element code

During first rise to power in Power Water Reactor, fuel pellets crack because of thermal expansion. The phenomena of pellet cracking and fragments relocation have a major influence on rod behaviour and especially on the cladding behaviour in the case of pellet–cladding interaction.This article presents the modeling used to take into account the fragmented state of the pellet in the EDF fuel rod thermo-mechanical code, CYRANO3^®. The aim is to simulate more realistic stress and strain fields in the pellet.The investigated method consists in adding parameters in the 1D finite elements calculations in order to integrate the multi-dimensional fragmentation effects in the axisymmetrical 1D code CYRANO3^®. These parameters modify the material behaviour by describing the fuel as an anisotropic damaged material. The modeling accounts for the opening and closing of radial pellet cracks. It has been implemented in the code for elastic and viscoplastic fuel behaviours.     

A model for the behavior of thorium uranium mixed oxide kernels in the pelletizing process

A behavior model of nuclear fuel kernels in the pelletizing process was developed to predict the microstructure of (Th,5%U)O₂ sintered pellets. Methods, equipments and components were developed in order to measure the density, the specific surface area and the crushing strength of the kernels and produce fuel pellets. It enables a correlation between the kernels properties and the microstructure, density and open porosity that were obtained in the fuel pellet produced with these kernels. It was possible to obtain a mathematical expression that allows one to calculate, from the kernel density and specific surface, the density that will be obtained in the fuel pellet for each compactation pressure value. The investigation showed which kernels properties are desired to obtain fuel pellets that satisfy the quality requirements for a stable performance in a power reactor. This model has been validated by experimental results and fuel pellets were obtained with an optimized microstructure that satisfies the fuel specification for an in-pile stable behavior.     

Supersonic Flow Patterns from Electrothermal Plasma Source for Simulated Ablation and Aerosol Expansion Following a Fusion Disruption

A pulsed electrothermal plasma source of a capillary discharge operating in the confined controlled arc regime is investigated to simulate the source term for ablation-induced regime of fusion reactor following hard disruption, in which ablation of diverter surface produces large aerosol transporting into the vacuum vessel. The source is attached to a converging–diverging micro-nozzle transition region to allow for the plasma flow and expansion into a large volume simulating large chamber of fusion reactor aerosol expansion to facilitate modeling of the plasma transport. This transition region connects to a 4 mm diameter capillary source and has a 3.33 mm converging section with a 2° converging angle, followed by a 146.7 mm diverging section with a 60° diverging angle, thus making an overall transition length of ~150 mm. The diverging section has an exit diameter of 50.82 cm to open into a large volume of the same exit diameter and a length of 1 m. Preliminary computation results indicate about 21 Mach number at the diverging exit and drops down to 0.7 Mach number after suffering from multiple shocks in the large uniform expansion volume. The plasma parameters entering the large chamber are maintained constant along the axis of the chamber for a simulated 1-D condition.     

Optimization of the radiation shielding mass for the magnet coils of the VISTA spacecraft

An attempt has been made for the optimisation of the radiation shielding of a spacecraft design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris, and with the help of a magnetic nozzle. The allowable nuclear heating in the super conducting magnet coils (up to 5 mW/cm³) is the crucial criterion for the dimensioning of the radiation shielding structure of the spacecraft. The optimized design reduced the shield mass from 600 tons to 93 and 88 tons with natural and enriched lithium, respectively. The space craft mass was 6000 tons. Total peak nuclear power density in the coils is calculated to be 5.0 mW/cm³ for a fusion power of 17,500 MW. Peak neutron heating density is 2.6 mW/cm³ and peak γ-ray heating density is 2.9 mW/cm³ (all on different points). However, volume averaged heat generation in the coils is much lower, namely 0.30, 0.73 and 1.03 mW/cm³ for neutron, γ-ray and total nuclear heating, respectively.     

五种组合氢同位素靶丸消融率的修正(英文)

对五种不同组合的固态氢同位素靶丸H2,HD,D2,DT和T2在聚变等离子体中的消融率首次作了修正研究。结果表明由于同位素效应引起的靶丸半径烧蚀率修正从氢靶丸的1下降到氚靶丸的0.487,因此在消融率计算时是不可忽略的。这些修正可导致更深的质量沉积和改善加料效率。     

Fuel transient deformation

Cladding strains resulting from fuel-cladding mechanical interaction in a transient tested fresh fuel pin are assessed against laboratory measurements of high-temperature creep and hot pressing of mixed-oxide fuel pellets.A fuel pin containing nine different fuel sections with varying fuel pellet geometry and density was transiently tested in a MARK IIIA flowing sodium loop under conditions typical of a 1$/s overpower transient. Post-test cladding strain measurements indicated that the largest strains were generated by solid pellets with small gaps while large gap annular pellets generated the smallest strains.High-temperature creep and hot pressing tests on mixed-oxide fuel have been performed at temperatures up to 2600°C. The results indicate that at temperatures above 2300°C, an additional component of creep is operative; while the densification due to hot pressing was considerably less than expected by extrapolating the low-temperature behavior.Both the in- and out-of-reactor data suggest that fuel creep into the center void or hole of a fuel pin is a more effective means of reducing fuel-cladding stress than densification by hot pressing into fuel pellet porosity.     

Fabrication of Sintered Annular Fuel Pellet for HANARO Irradiation Test

The fabrication method of an annular pellet with highly precise diametric tolerances, same dimensions, and various sintered densities has been investigated. To examine the in-pile densification and swelling of the annular pellet, 5 different types of annular pellet were prepared for a HANARO irradiation test. In order to obtain annular fuel pellets with the same dimensions and various sintered densities, we control the green density of an annular compact, the sintering temperatures, and the sintering time. For a diametric tolerance control, we have introduced a new compaction process that combines the usual double-acting pressing and cold isostatic pressing. Annular fuel pellets with the same dimensions and various sintered densities were fabricated successfully, and all the pellets satisfied the pellet specification of the HANARO irradiation test. Sintered annular pellets show an excellent inner diametric tolerance of less than ±12 μm without an inner surface grinding.