Pd-Ag膜电解浓缩氚的动力学模型及理论计算

建立了模拟Pd-Ag合金膜电解浓缩氚的过程和诸多电解参数的动力学模型。采用求解数学扩散方程的方法得到理论值,并与文献报道的实验数据进行比较。计算结果表明：氚在Pd-Ag膜上的吸附、解吸、浓缩等行为受到诸多条件的制约。在保证密封性能的前提下,对Pd-Ag膜电解槽体的要求是至少应包括有含氚水路循环、阳极气体消除和阴极扩散后氚再生3个组成部分,采取相匹配的级联技术可以提高分离效果;对电解参数的要求是采取尽可能高的电解液温度和稳定的OH^-浓度,合适的电流密度,合理的膜厚度、表面特征和Pd黑结合紧密。     

Generation of the line radiation of argon added to DT gas in Iskra-5 experiments

The first experiments measuring the density of a compressed deuterium and tritium mixture in microtargets of indirect irradiation (x-ray targets) were performed at the Iskra-5 facility. The density was determined according to the broadening of the lines of hydrogen-and helium-like argon added to the DT gas as a diagnostics material. A series of three experiments was performed with x-ray targets in which the central capsule filled with a DT + Ar mixture over a range of shell thicknesses. In two of the experiments, argon emission spectra were recorded and the density of the compressed gas was determined. For a microtarget approximately 280 μm in diameter with a wall approximately 7 μm thick, an analysis of the experimental results yielded an estimated density in the compressed gas of ∼1 g/cm³. Gas-dynamic calculations using the SNDA (spectral nonequilibrium diffusion with absorption) program show that argon emission takes place just after reaching maximum temperature, but much sooner than maximum compression. The results of a calculation for an experiment with low relative Ar concentration are in overall agreement with the experimental data. Additional investigations are needed to interpret experiments at a relatively high concentration. Zh. éksp. Teor. Fiz. 114, 837–848 (September 1998)     

Facility for preparation of gas mixture in muon catalyzed fusion experiments

A facility is described that allows safe handling of high tritium gas activity as dozens kilocuries in a regular laboratory environment. It is used to make and deliver into the target a mixture of specific isotopic composition with the contamination requirement of 10^-7 v.f. for Z>1 elements, and recover it upon completion of operation. With this facility, efforts have been accomplished to investigate into the muon catalyzed fusion on two targets – liquid tritium and high-pressure tritium types. Also, the operation range was 0.1–120 MPa for pressure and 20–800 K for temperature and the amount of tritium used was about 100 kCi. The facility showed reliability in operation without indications of radiation beyond the safety level. This revised version was published online in August 2006 with corrections to the Cover Date.     

高压氘氚靶球的制备与拉曼光谱研究

氢同位素的定量分析与监测在能源与环境领域都有着重要的意义。激光拉曼光谱由于其可以无损分析氢同位素分子,已经成为一种重要的方法,在国际热核聚变实验反应堆(ITER)和美国萨凡纳河工厂得到了广泛应用。利用高压充气装置得到了惯性约束聚变(ICF)高压靶丸,并对靶丸内气体进行原位拉曼光谱测量,通过对高压下氘氚混合气体的拉曼光谱进行分析得到了靶丸内气体的成分比例,验证了靶丸充气工艺参数。实验表明,在CCD的积分时间延长到1 min时,氘(DD),氘氚(DT)和氚(TT)的测量精度可以达到1%,同时对不同时刻靶丸内气体组分的拉曼光谱进行测量,实验结果表明在氘氚渗透和氚衰变两者共同作用下,靶丸内总气体压力随时间不断下降,但是气体组成基本不发生变化。     

Simultaneous catalytic regime of tritium and helium-3 in D-D fusion without external breeding

A catalytic regime of tritium and helium-3 in deuterium-deuterium fusion, including ion-electron collisions, mechanical expansion, bremsstrahlung radiation, inverse Compton scattering losses and reacting particles energy effect has been investigated. In this paper a new fuel configuration, DT _x ³He _y , is formed by adding ³He to DT fuel. According to our calculations this fuel (DT _x=0.0112 ³He _y=0.0399) has greater energy gain than the fuel (DT _x =0.0112) used by Eliezer et al [Eliezer et al, Nucl. Fusion 40, 195 (2000)] and also it does not require external tritium and helium-3 breeding. Furthermore, neutron yields in D-D and D-T reactions are reduced due to the reduced quantity of initial amount of deuterium and tritium.     

A small and compact AMS facility for tritium depth profiling

Depth profiling measurements of tritium in carbon samples have been performed during the past seven years at the AMS facility installed at the Rossendorf 3 MV Tandetron. The samples have been cut from the inner walls of the fusion experiments ASDEX-upgrade/Garching and JET/Culham. The tritium content of the samples from JET required a dedicated AMS facility to prevent any contamination of the versatile 3 MV Tandetron. On the basis of an air-insulated 100 kV tandem accelerator equipped with a gas stripper an AMS facility exclusively devoted to tritium depth profiling was installed, tested and used for routine measurements. After additional successful tests employing diamond-like carbon (DLC) stripper foils at this accelerator, another small and compact 100 kV tandem accelerator with SF₆ insulation and a DLC stripper has been installed at the AMS facility. Results obtained with the different tandem accelerators are presented.     

Laser-optical path to nuclear energy without radioactivity: Fusion of hydrogen–boron by nonlinear force driven plasma blocks

Anomalous interaction of terawatt-picosecond laser pulses allows side-on ignition of solid state density fusion fuel with the unexpected possibility of igniting uncompressed hydrogen–boron p-¹¹B. Suppression of relativistic self-focusing by using very clean laser pulses with an extremely high contrast ratio is essential to achieve ignition thresholds only ten times more difficult than fusion of deuterium–tritium (DT). This opens the possibility for laser driven fusion energy without neutrons and less radioactivity than from burning coal. The complex nonlinear optical properties involved are elaborated.     

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m³Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.     

Diffusion filling with fuel gas of high-gain direct-drive cryogenic targets

To provide continuous operation of a reactor based on inertial confinement fusion (ICF), the thermonuclear burn region should be refilled with fuel with a frequency of 1 million targets per day. The first stage in the target production is diffusion filling of polymeric (CH) shells with fuel gas which is deuterium (D₂) or deuterium–tritium (DT) mixture. The results of simulation of filling reactor-scale CH-shells (Ø ~ 4 mm) to a pressure of ~1100 atm at 300 K in the mode with a constant pressure gradient are presented. Simple and two-layer shells of compact and porous polymers are considered. The problems of constructing an optimum DT-filling scheme avoiding CH-shell fracture due to tritium beta decay are discussed.     

Deuterium–tritium catalytic reaction in fast ignition: Optimum parameters approach

One of the main concerns about the current working on nuclear power reactors is the potential hazard of their radioactive waste. There is hope that this issue will be reduced in next generation nuclear fusion power reactors. Reactors will release nuclear energy through microexplosions that occur in a mixture of hydrogen isotopes of deuterium and tritium. However, there exist radiological hazards due to the accumulation of tritium in the blanket layer. A catalytic fusion reaction of DT _x mixture may stand between DD and an equimolar DT approach in which the fusion process continues with a small amount of tritium seed. In this paper, we investigate the possibility of DT_x reaction in the fast ignition (FI) scheme. The kinematic study of the main mechanism of the energy gain–loss term, which may disturb the ignition and burn process, was performed in FI and the optimum values of precompressed fuel and proton beam driver were derived. The recommended values of fuel parameters are: areal density ρ R ≥ 5g · cm^?2 and initial tritium fraction x ≤ 0.025. For the proton beam, the corresponding optimum interval values are proton average energy 3 ≤ E _p ≤ 10 MeV, pulse duration 5 ≤ t _p ≤ 15 ps and power 5 ≤ W _p ≤ 12 × 10²² (keV·cm³ · ps^?1). It was proved that under the above conditions, a fast ignition DT _x reaction stays in the catalytic regime.     

Tritium gas handling system for muon catalyzed fusion research at RIKEN-RAL muon facility

For muon catalyzed fusion (μCF) experiments at RIKEN-RAL muon facility, a tritium gas handling system for a high purity D–T target, free from ³ component, has been constructed to perform precise measurements of α-sticking probability in the μCF cycle. The system has been constructed to enable us to purify the target D–T gas by removing ³He component, to adjust the D/T mixing ratio, and to measure the hydrogen isotope components at the experiment site. The whole performance has been confirmed and a tritium gas with the inventory of 56 TBq (1500 Ci) has been operated in the system. This revised version was published online in August 2006 with corrections to the Cover Date.     

Feasibility of the Precise Energy Calibration for Fast Neutron Spectrometers

Computational studies aimed at improving the accuracy of measurements performed using neutron generators with a tritium target were performed. A measurement design yielding an extremely narrow peak in the energy spectrum of DT neutrons was found. The presence of such a peak establishes the conditions for precise energy calibration of fast-neutron spectrometers.     

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

The control of the deuterium–tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10–200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1–4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.     

Fast ignition by detonation in a hydrodynamic flow

A general concept of fast ignition by a hydrodynamic pulse is developed. The main statements of the concept are formulated having in mind the need to ignite the pre-compressed thermonuclear fuel of the inertial confinement fusion (ICF) target. Initially, combustion must be initiated inside the hydrodynamic flow during its action on the target. The conditions for propagating a self-sustaining thermonuclear-detonation wave from an igniter on the thermonuclear fuel of the ICF-target must be provided. For this, the deuterium–tritium (DT) igniter placed in the forward part of the hydrodynamic flow should not only be heated up to thermonuclear temperature, but also compressed to a density close to the density of the ICF-target fuel. It is shown that the detonation of the multilayer conical target (containing DT-ice and a heavy pusher) enables fast ignition of the ICF target fuel of 200–500 g/cm³ density at an implosion velocity of 300–500 km/s.     

Propagation of ion shock in solid DT target with nonlinear force-driven plasma blocks

The plasma block (piston) with pressure P ₁ is generated as a result of the nonlinear (ponderomotive) force in laser–plasma interaction. The plasma block can be used for the ignition of a fusion flame front in a solid density deuterium–tritium (DT) target by compressing the fuel that creates an ion shock propagating with velocity u _{ion? shock} in the inside of a solid DT target. The ignition is achieved by creating an ion shock during the final stages of the implosion. We estimated the effect of an ion shock in solid DT target at an early stage with no compression and at the last stage with compression, where density increases by a factor of solid-state density. According to the theoretical model, a large target with a very thin layer of fuel (high-aspect ratio target) would be ideal to obtain the very strong shocks. Results indicate that the maximum compression even by an infinitely strong single shock can never produce more than four times the initial density of DT fuel. The results reported that the threshold ignition energy in a solid DT target is reduced by a factor of 4.     

A simple polymer electrolyte membrane system for enrichment of low-level tritium (3H) in environmental water samples

Tritium (³H) is an essential tracer of the Earth's water cycle; yet widespread adoption of tritium in hydrologic studies remains a challenge because of analytical barriers to quantification and detection of ³H by electrolytic pre-concentration. Here, we propose a simple tritium electrolytic enrichment system based on the use of solid polymer electrolyte membranes (PEMs) that can be used to enrich ³H in 250–3000?mL environmental water samples to a 10-mL final volume. The IAEA PEM-³H system reported here can produce high enrichment factors (>70-fold) and, importantly, removes some of the deterrents to conventional ³H enrichments methods, including the use of toxic electrolysis and neutralization chemicals, spike standards, a complex electrolysis apparatus that requires extensive cooling and temperature controls, and improves precision by eliminating the need for tracking recovery gravimetrics. Preliminary results with varying operating conditions show ³H enrichments to 70-fold and higher are feasible, spanning a wide range of tritium activities from 5 to 150 TU with a precision of ～4.5?%. Further work is needed to quantify inter-sample memory and to establish lower ³H detection limits. The IAEA PEM-³H system is open source, with 3-D CAD and design files made freely available for adoption and improvement by others.     

磁化套筒惯性聚变初步探索研究与实验可行性分析

磁化套筒惯性聚变(MagLIF)构型可充分利用现有大型脉冲功率驱动装置,如聚龙一号等。基于磁流体力学方程组和1∶1比例氘氚(DT)混合燃料聚变模型,开发了零维MagLIF数值模拟程序并进行了初步探索研究。计算结果表明初始负载参数(如轴向磁场强度,预加热温度、时刻,负载半径等)与聚变产额之间有着密切的联系,在给定条件下,可依据计算给出的定性关系进行负载优化设计。值得注意的是,根据计算结果,即使在理想条件下,氘氚燃料要实现能量收支平衡,则驱动器的电流必须大于21.2 MA。这意味着聚龙一号装置(10 MA)无法开展集成化的MagLIF实验,进一步的校验计算验证了上述观点,并在此基础上提出铝套筒分解实验的建议和负载设计参数。所取得的计算结果有利于加深对MagLIF套筒压缩阶段物理过程的认知和理解。     

冷冻靶校准误差和烧蚀层粗糙度对氘氚层均匀性的影响

 理论分析了氘氚层外表面的温差与其粗糙度间的关系;以法国兆焦激光装置LMJ为原型,利用计算流体力学程序Fluent,分别模拟了靶丸轴向偏离黑腔中心不同尺度和烧蚀层存在不同大小的非均匀厚度对氘氚层温度分布的影响,求得了这两种误差引起氘氚层厚度的非均匀度。结果表明：为了满足点火靶的要求,靶丸轴向偏离腔体中心的尺度须在8.5 μm内,烧蚀层轴向粗糙度则应控制在0.72 μm内。     

BRIF and CARIF progress

China Institute of Atomic Energy (CIAE) is currently constructing Beijing rare ion beam facility (BRIF) and is proposing China advanced rare ion beam facility (CARIF). This paper is aiming at introducing the progress of BRIF project and the conceptual design CARIF. The ISOL type facility BRIF under construction is composed of a 100 MeV 300 μA proton cyclotron, an ISOL with mass resolution of 20000, and a super-conducting LINAC of 2 MeV/q, and will be commissioned in 2013. CARIF facility proposed is planned to use both ISOL and PF techniques. It is based on a China advanced research reactor CARR that was critical, with ISOL separation of fission fragment, post acceleration to 150 MeV/u, and fragmentation of neutron-rich fission fragment beam like ¹³²Sn. Such unique combination will allow CARIF to deliver beam intensity better than the best world facilities by more than one order of magnitude.     

Seasonal and Annual Tritium Variations in Belgrade Natural Waters

Characteristics of seasonal and annual variations of tritium concentration completed with hydrological long-term data from 1976–1989 for the Belgrade alluvial aquifer are presented. The highest tritium concentration in precipitation of an average year appears at the beginning of summer with a maximum in June (9 Bq/l) when the amount of precipitation is also high (102 l/m²). The quantity of tritium precipitating in this area was also the largest in the summer, especially in June (922 Bq/M²). Similar variations of tritium concentration was found in the waters of the Danube and the Sava with maxima 11.3 Bq/l (July) and 9 Bq/l (June) respectively. Tritium concentrations orginated from snowmelt influence to the occurrence of higher values during that period which is more distinct in the Sava. The groundwaters and particularly ones in the Ranney wells follow changes of levels, temperature and tritium content in the alluvial of Sava with a time lag from a few days to a month. During the period of observation tritium content was decreasing in all studied waters. The decrease was faster in precipitation (11.2 to 3.0 Bq/l) and the Sava and Renney wells (10.6 to 3.0 Bq/l) than in the Danube (13.1 to 5.2 Bq/l).