共查询到20条相似文献,搜索用时 15 毫秒
1.
Keeman Kim Hyoung Chan Kim Sangjun Oh Young Seok Lee Jun Ho Yeom Kihak Im Gyung-Su Lee George Neilson Charles Kessel Thomas Brown Peter Titus 《Fusion Engineering and Design》2013,88(6-8):488-491
As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb3Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters. 相似文献
2.
Takuya Goto Teruya Tanaka Masahiro Tanaka Akio Sagara 《Fusion Engineering and Design》2012,87(7-8):1286-1290
A new concept of a fusion reactor system, MFE-IFE cooperative system, is proposed. This concept combines the merits of a small-size MFE reactor and a dry-wall IFE reactor and aims at sufficient amount of tritium production and electricity generation without advanced technology. Design window analysis shows a NIF-scale (5 m chamber radius) dry-wall laser fusion reactor with a ~1 GWth fusion output and net tritium breeding ratio (TBR) of 1.74 can sustain an MFE power plant with a fusion power of 3 GWth and net TBR of 0.96. Although more detailed quantitative analyses are required, this concept can be a possible solution for a simultaneous achievement of tritium self-sufficiency and significant net electricity generation. 相似文献
3.
《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》1998,142(3):387-392
Organometallic Vapor Phase Epitaxy (OMVPE) grown layered structures of In0.1Ga0.9As/GaAs have been analyzed by He++ ion backscattering and channeling. From the random spectrum, the InxGa1 − xAs layer thickness was determined to be around 300 Å and the composition of In to be In0.1 within an experimental error of around 5%. The χmin obtained from the In signal is around 7% which shows that the epilayer is of good crystallinity. The normalized yield vs. the tilt angle for the epilayer and the substrate along the off-axis (along [1 1 0] direction of the substrate) shows a shift in the minimum yield χmin dip, which is a direct measure of the strain present. This shift is found to be 0.2 ± 0.05° corresponding to a tetragonal distortion of 0.7 ± 0.2%. Shifting of the minimum yield dip of the overlayer towards left side with respect to the substrate indicates that the strain is compressive which is what is expected. X-ray diffraction is also carried out on the same sample which gives us ϵ⊥=1.01% and the in-plane lattice mismatch is nearly zero. 相似文献
4.
《Annals of Nuclear Energy》1999,26(6):509-521
Radiation shielding structure of a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit has been investigated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle. The nuclear heating in the super conducting magnet coils determines the radiation shielding mass of the spacecraft. It was possible to achieve considerable mass saving with respect to a recent design work, coupled with higher design limits for coil heating (up to 5 mW/cm3). The neutron and γ-ray penetration into the coils is calculated using the SN methods with a high angular resolution in (r–z) geometry in S16P3 approximation by dividing the solid space angle in 160 sectors. Total peak nuclear heat generation density in the coils is calculated as 3.143 mW/cm3 by a fusion power of 17 500 MW. Peak neutron heating density is 1.469 mW/cm3 and peak γ-ray heating density is 1.674 mW/cm3. However, volume averaged heat generation in the coils is much lower, namely 74, 163 and 337 μW/cm3 for neutron, γ-ray and total nuclear heating, respectively. The net mass of the radiation shielding for the magnet coils is 200 tonne by a total mass of 6000 tonne of the space craft. 相似文献
5.
O. Gruber 《Fusion Engineering and Design》2009,84(2-6):170-177
The aim of the ASDEX Upgrade (AUG) programme is to support the design, prepare the physics base and develop regimes beyond the baseline of ITER and for DEMO. Its ITER-like geometry, poloidal field system, versatile heating system and power fluxes make AUG particularly suited.After the transition to fully tungsten coated plasma facing components AUG could be operated without prior boronizations and a low permanent deuterium retention was found qualifying W as wall material. ITER-like baseline H-modes (H98 ~ 1, βN ~ 2) were routinely achieved up to 1.2 MA plasma currents. W concentrations could be kept at an acceptable level of <5 × 10?5 by central wave heating (enhancing impurity outward transport) and ELM pacing with gas puffing. The compatibility of high performance improved H-modes, the ITER hybrid scenario, with an un-boronized W wall was demonstrated achieving H98 ~ 1.1 and βN up to 2.6 at modest triangularities δ ≤ 0.3. This performance is reached despite the gas puffing needed for W influx control. Increasing δ to 0.35 allowed at even higher puff rates still a H98 ~ 1.1.Reliable plasma operation in support of ITER comprised the demonstration of ECRF assisted low voltage plasma start-up and current rise at toroidal electric fields below 0.3 V/m resulting in a ITER compatible range of plasma internal inductance of 0.71–0.97. Disruption mitigation is feasible using strong gas puffs, and the achieved electron densities approach values needed for runaway suppression.Present hardware extensions in support of ITER include the upgrading of ECRH by a 4 MW/10 s system with large deposition variability (tuneable frequency between 105 and 140 GHz, real-time steerable mirrors) for central heating and MHD mode control. A powerful system of 24 in-vessel coils produces error fields up to toroidal mode number n = 4 for ELM suppression and mode rotation control. In connection with a close conducting wall they will open up the road for RWM stabilization in advanced scenarios. For those we are considering LHCD for current drive and profile control with up to 500 kA driven current. The tungsten sources are dominated by sputtering from intrinsic light impurities, and the W influx from the outboard limiters are the main source for the core plasma. ICRH induced electric fields accelerate light impurities, restricting the use of ICRH to just after boronization. 4-strap antennas imbedded in extended wall structures might solve this problem. Finally, doubling the plasma volume with plasma currents above 2 MA in AUG could be the solution for a needed ITER satellite. 相似文献
6.
《Fusion Engineering and Design》2014,89(7-8):1019-1023
The generation and diffusion of runaway electrons (REs) during major disruptions in the HL-2A tokamak has been studied numerically. The diffusion caused by the magnetic perturbation is especially addressed. The simulation results show that the strong magnetic perturbation (δB/B ∼ 1.0 × 10−3) can cause a significant loss of REs due to the radial diffusion and restrain the RE avalanche effectively. The results also indicate that the REs are generated initially in the plasma core during disruptions, and that the toroidal electric field does not exhibit a centrally hollow phenomenon. In addition, it is found that the toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field. 相似文献
7.
This study analyzes the effects of certain heavy-metal-salt fluids on nuclear parameters in a fusion–fission hybrid reactor. Calculated parameters include the tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate, and fissile fuel breeding in the reactor's liquid first wall, blanket, and shield zones; gas production rates in the structural material of the reactor were calculated, as well. The fluid mixtures consisted of 93–85% Li20Sn80 + 5% SFG-PuO2 and 2–10% UO2, 93–85% Li20Sn80 + 5% SFG-PuO2 and 2–10% NpO2, and 93–85% Li20Sn80 + 5% SFG-PuO2 and 2–10% UCO. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A 3 cm wide beryllium (Be) zone was used for neutron multiplier between the liquid first wall and the blanket. The structural material used was 9Cr2WVTa ferritic steel, measuring 4 cm in width. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and the ENDF/B-VII.0 nuclear data library. 相似文献
8.
《Journal of Nuclear Materials》2006,348(1-2):26-32
Mass and charge transfer in a proton-conducting ceramic with internal reformation under the supply of CH4 + H2O was experimentally investigated for application to a fuel detritiation system of a fusion reactor. The oxide used in the present experiment was SrCe0.95Yb0.05O3−a, and the electrodes were composed of Ni–SiO2 paste and Ni wire mesh. The system was described by CH4 + H2O∣Ni∣SrCe0.95Yb0.05O3−a∣NiO∣O2 + H2O. Plots of the I–V (electric current density versus cell potential) characteristic curve were determined under the conditions of different H2O/CH4 concentration ratios and temperatures of 600–800 °C. It was found that the system could work well even without any external CH4 reformer. Mass-transfer process in/on the porous Ni electrode and in the ceramic electrolyte was experimentally clarified. The distribution of carbon depositions in the porous electrode was also determined with EDX by scanning over entire surface in the scope of SEM. The ratio of CH4 to H2 direct decomposition to its steam-reforming reaction was found to be different from location to location in the porous Ni electrode. 相似文献
9.
G. Calabrò F. Crisanti G. Ramogida P. Mantica B. Baiocchi A. Cucchiaro P. Frosi V. Fusco Y. Liu S. Mastrostefano F. Villone G. Vlad R. Fresa 《Fusion Engineering and Design》2013,88(6-8):858-862
The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current IP = 10 MA, toroidal field BT = 8.5 T, with a q95 ≈ 2.3 that would correspond to IP ≈ 20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated. 相似文献
10.
Thomas Geruschke Katharina Lorenz Eduardo Alves Reiner Vianden 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2009,267(8-9):1472-1475
Hafnium ions were implanted into calcium fluoride single crystals. The lattice damage introduced by the implantation was investigated with the Rutherford backscattering (RBS) channelling technique. The lattice location of the implanted ions was determined by performing channelling measurements for the 〈1 1 0〉 crystal direction. A comparison of the angular scan with Monte Carlo simulations leads to the conclusion that >90% of the Hf ions are on Ca sites directly after implantation. Subsequent annealing of the samples was performed in a rapid thermal annealing apparatus. Perturbed angular correlation (PAC) measurements with 181Hf(181Ta) show quadrupole interactions with νQ1 = 300(3) MHz (η = 0.00), νQ2 = 1285(13) MHz (η = 0.43) and νQ3 = 1035(10) MHz (η = 0.00) after annealing up to 1200 K. 相似文献
11.
The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality keff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GWth requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBRcycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality keff ≤ 0.97 and produces less fission power of 1.5 GWth. Its main performance parameters turn out substantially better. 相似文献
12.
《Fusion Engineering and Design》2014,89(11):2523-2528
The transmutation characteristics of minor actinides in the transmutation reactor of a low aspect ratio (LAR) tokamak are investigated. One-dimensional neutron transport and burn-up calculations coupled with a tokamak systems analysis were performed to determine optimal system parameters. The dependence of the transmutation characteristics, including the neutron multiplication factor, produced power, and the transmutation rate, on the aspect ratio A in the range of 1.5–2.0 was examined. By adding Pu239 to the transmutation blanket as a neutron multiplication material, it was shown that a single transmutation reactor producing a fusion power of 150 MWth can destroy minor actinides contained in the spent fuels for more than 38 units of 1 GWe pressurized water reactors (PWRs) while producing a power in the range of 1.8–6.8 GWth. 相似文献
13.
《Journal of Nuclear Materials》2006,348(1-2):122-132
The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry/synchrotron powder X-ray diffraction between 25 °C and 725 °C at a heating rate of 10 °C min−1. The graphite, having been subject to a fast-neutron fluence from ∼1017 to ∼1020 n cm−2 over the life time of the reactor at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 25 to 572 J g−1 and a Wigner energy accumulation rate of ∼7 × 10−17 J g−1/n cm−2. The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 °C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence >1.5−5.0 × 1019 n cm−2) is destroyed while that of samples farther from the reactor core (fast-neutron fluence <1.5−5.0 × 1019 n cm−2) is intact, with marked swelling along the c-axis. The dependence of the c lattice parameter on temperature between 25 °C and 200 °C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c-axis of ∼ 26 × 10−6 °C−1. However, at 200 °C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The 14C activity in the inner thermal column graphite ranges from 6 to 467 kBq g−1. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 °C for 24 h in a hot-cell facility prior to interim storage. 相似文献
14.
《Nuclear Engineering and Design》2005,235(17-19):1799-1805
Small punch (SP) tests were performed to evaluate the ductile–brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm × 10 mm × 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 °C. Small punch tests were performed in the temperature range of RT to −196 °C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (TSP), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T41 J. TSP from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test (ΔT41 J) shows a better correlation with the transition temperature shift (ΔTSP(E)) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. TSP also had a correlation with the reference temperature (T0) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen. 相似文献
15.
Ph. Mertens H. Altmann P. Chaumet E. Joffrin M. Knaup G.F. Matthews O. Neubauer D. Nicolai V. Riccardo V. Tanchuk V. Thompson R. Uhlemann U. Samm 《Fusion Engineering and Design》2011,86(9-11):1801-1804
In the frame of the ITER-like Wall (ILW) for the JET tokamak, a divertor row made of bulk tungsten material has been developed for the position where the outer strike point is located in most of the foreseen plasma configurations. In the absence of active cooling, this represents a formidable challenge when one considers the temperature reached by tungsten (TW,surf > 2000 °C) and the vertical gradient ?T/?z = 5 × 104 K/m.As the development is drawing to an end and most components are in production, actual 1:1 prototypes are exposed to an ion beam with a power density around 7 MW/m2 on the plasma-facing surface. Advantage is taken of the flexibility of the Marion facility to bombard the tungsten stack under shallow angles of incidence (~6°) with a powerful beam of ions and neutrals (>70 MW/m2 on axis). The shallow angles are important, with respect to the toroidal wetted surface, for properly simulating the expected performance under actual tokamak conditions. The Marion tests have been used to validate for a few typical cases the thermal calculations that were steadily developed along with the tungsten tile and, at the same time, to gather information on the actual temperatures of individual components. The latter is an important factor to a finer estimation of the power handling capabilities. 相似文献
16.
Y. Qi T. Prenzel T.A. Harriman Y.Q. Wang D.A. Lucca D. Williams M. Nastasi J. Dong A. Mehner 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2010,268(11-12):1997-2000
A study of the effects of ion irradiation of organically modified silicate thin films on the loss of hydrogen and increase in hardness is presented. NaOH catalyzed SiNawOxCyHz thin films were synthesized by sol–gel processing from tetraethylorthosilicate (TEOS) and methyltriethoxysilane (MTES) precursors and spin-coated onto Si substrates. After drying at 300 °C, the films were irradiated with 125 keV H+ or 250 keV N2+ at fluences ranging from 1 × 1014 to 2.5 × 1016 ions/cm2. Elastic Recoil Detection (ERD) was used to investigate resulting hydrogen concentration as a function of ion fluence and irradiating species. Nanoindentation was used to measure the hardness of the irradiated films. FT-IR spectroscopy was also used to examine resulting changes in chemical bonding. The resulting hydrogen loss and increase in hardness are compared to similarly processed acid catalyzed silicate thin films. 相似文献
17.
《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2004,226(3):274-280
We report the low temperature (below the metal–insulator transition temperature Tim) resistivity and magnetoresistance (MR) behavior of 50 MeV Li3+ beam irradiated La0.7Pb0.3MnO3 for three different fluences. Ion beam irradiation causes a decrease of Tim leading to the increase of insulating regime. Resistivity data of the unirradiated as well as irradiated samples fitted well with an equation of the form ρ = ρ0 + ρ2.5T2.5 which indicates predominant contribution from the electron–magnon interaction (second term). The temperature dependent MR data of samples irradiated with different ion fluences follow the simple relation [MR = a + b/(T + C)] showing appreciable effect of radiation on the parameters a, b and C. The physical significance of the radiation effect on these parameters is not yet very clear. 相似文献
18.
《Fusion Engineering and Design》2014,89(7-8):1380-1385
China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by 6Li(n,α)T reaction. Li2TiO3 pebbles and Be12Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li2TiO3 and Be12Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be12Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported. 相似文献
19.
《Fusion Engineering and Design》2014,89(6):847-852
EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current Ip ≤ 1 MA, toroidal field BT ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s. 相似文献
20.
《Fusion Engineering and Design》2014,89(7-8):1054-1058
This study proposes a probability of the evaporated gas that agitates a growing instability wave in a thin liquid film first wall. The liquid first wall was considered to be in vacuum and the effect of the ambient gas was neglected but the evaporated gas by the high energy fluxes is a probable cause of unstable wave agitation. The criterion is approximately expressed by the density ratio (Q2) and the Weber number (We) as Q2 × We0.5 ≈ 5 × 10−4. Performed indirect experimental supported this criterion. For a case study of liquid Pb-17Li film with a velocity of 10 m/s, the evaporated gas pressure must be below 6.2 × 103 Pa to maintain stable conditions. By recent study, this pressure is generated at 1600 K temperature and it is believed to be attainable by the energy fluxes on the first wall. This result is so far not confirmed so the full verification by experimental is to be performed. 相似文献