Estimate of the Efficiency of Hydrogen–Metal Hydride (Intermetallic Compound) Systems in the Separation of the Three-Isotope Mixtures H–D–T

A method is developed for calculating apparatus for separating a three-isotope mixture H–D–T by the sorption method using hydrides of metals or intermetallic compounds. Relations are established between the partition coefficients and mass-transfer characteristics, determined for binary isotopic mixtures of hydrogen, and the parameters determining the isotopic equilibrium and the kinetics of mass transfer in the system gas–solid hydride phase with three isotopes. The efficiencies of various sorbents for deprotization of deuterium–tritium mixture are compared. 4 figures, 2 tables, 7 references.     

Concentration of water samples for determining the tritium content

A method of concentrating the tritium in water samples by means of distillation under vacuum (~100 mm Hg) is suggested and experimentally verified. The degree of tritium enrichment is determined by the change in concentration of stable isotopes (oxygen or deuterium) on distillation of the sample. Two variations of the calculation are proposed: for the stationary state with the attainment of the maximtu~ possible degree of fractionation and cessation of enrichment at the nonstationary state.Translated from Atomnaya Énergiya, Vol. 18, No. 4, pp. 367–372, April, 1965     

Rectification Decontamination of Tritium-Containing Waters

The results of experimental investigations of the main rectification processes for removing tritium from reactor, process, and waste waters are presented. The separation coefficients are presented for rectification of water (1.028), ammonia (1.05), the azeotrope H₂O–HTO–HNO₃ (1.098), and D₂O–DTO–DNO₃ (1.039). Practical schemes for detritization setups are examined.     

Liberation of deuterium gas from molybdenum upon heating after penetration of D+ ions

Conclusion A series of peaks of liberation of deuterium gas from molybdenum has been discovered in this paper upon the heating of a sample after ion penetration. The maxima of these peaks lie, in the 230–650°K range. The diffusion coefficient and the activation energy in the individual peaks of j(T) have been measured. It has been discovered that D in the individual peaks covers a range of at least 10 orders of magnitude when the temperature is constant, and the discrete activation energy series comprises 45–126 kJ/mole. The measured values of D and Q can be larger or smaller than those which are measured in experiments on hydrogen transport. At high irradiation doses an interrelationship has been discovered in the first group of peaks between secondary defects in molybdenum and the liberation of captured deuterium gas. It is suggested that secondary defects can serve, in different cases both as centers for deuterium retention and as the means for its easier escape to the surface. It has been assumed possible to explain the high-temperature peaks as the release of deuterium from radiation vacancy-deuterium atom complexes with a different content of vacancies and deuterons.Translated from Atomnaya Énergiya, Vol. 51, No. 1, pp. 31–34, July, 1981.     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

氚老化对钯钇合金膜氢渗透性能的影响

在聚变堆燃料循环系统中,钯合金膜将被用于氢同位素与杂质气体间的渗透分离以及含氚杂质中氚的催化回收。长期连续的氚操作将使合金膜体内因氚衰变而累积3He,产生氚老化效应。本工作研究了贮氚老化对Pd8.5Y0.19Ru（原子百分数）合金膜的氕、氘渗透性能的影响。研究结果表明：对于膜内体氦浓度He/M为0.042的氚老化膜,在573～723K温度范围内,氕、氘渗透率被严重降低,膜的氕氘渗透分离系数则有所提高。     

Fusion Safety Studies in Russia in 1996

The paper seeks to provide a summary report of observations and results of some Russian fusion safety studies performed in 1996. Release of tritium and helium from neutron irradiated beryllium at relatively high neutron fluences has a burst nature. With the growth of the beryllium temperature-increase rate to 90 K/s, the temperature of tritium burst release decreases from 800 to 450–500°C and for helium decreases from 1200 to 500°C. Characterization of carbon and tungsten dust produced in experiments simulating plasma disruptions revealed that dust particle distribution of sizes for graphites and carbon fiber composites has a bimodal nature with maxima in the range of 0.01–0.03 and 2–4 m for composite UAM and in the range of 0.14–0.18 and 2–4 m for graphite MPG-8. Chemical reactivity of beryllium with air was studied as well. A mathematical model for beryllium weight gain under its chemical interaction with air at temperatures of 700–800°C as a function of beryllium porosity, temperature, and interaction duration was developed.     

Mutual diffusion in the system zirconium-15 at. % uranium

The author has studied mutual diffusion in the system zirconium— 15 at. % uranium. The diffusion coefficients in the body-centered high-temperature phase were determined at 950–1450° C with -radiation of uranium. It is shown that alloying zirconium with uranium reduces the diffusion mobility throughout the whole of the investigated temperature range and increases the activation energy. The temperature dependence of the diffusion coefficient of uranium in zirconium, obtained by extrapolating the diffusion parameters to zero uranium concentration, is represented by the equation D= 1.40·10^–4 exp (–27,700/RT) cm²/sec.Translated from Atomnaya Énergiya, Vol.22, No.4, pp.290–292, April, 1967.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.     

Penetration of hydrogen isotopes through the superalloy ÉI698 at high pressures and temperatures

Conclusions We determined the parameters of the processes of penetration and diffusion of the hydrogen isotopes through the superalloy ÉI698 in the 600–1050°K range at pressures up to 100 MPa. It was shown that the penetrating flux is in proportional to p in the 0.1–100 MPa pressure range.The diffusion coefficient of hydrogen (deuterium) does not depend on the concentration of the dissolved hydrogen (deuterium) that attains a value of 1.3·10⁶ cm³/cm³ Me at a pressure of 100 MPa and a temperature of 1050°K. The concentration of the dissolved hydrogen is not critical with respect to hydrogen-embrittlement of the ÉI698 alloy subjected to loading for a period of 600 h under an internal pressure of 100 MPa. Based on the obtained values of the penetration rates of hydrogen and deuterium through the ÉI698 alloy, the decrease in the quantity of the gas in the vessel can be taken into account properly when carrying out experimental studies on the processes of mu-catalysis.Translated from Atomnaya Énergiya, Vol. 65, No. 6, pp. 395–399, December, 1988.     

Absorption parameters of deuterium ions in molybdenum

Conclusions The capture parameters of deuterium ions in molybdenum have been studied in this paper. It has been shown that the capture coefficients decrease as the dose and the irradiation temperature increase and they increase as the energy and flux density of the ions increases. Upon heating after irradiation or exposure in a vacuum even at room temperature the number of deuterons captured in molybdenum decreases. At a raised temperature the capture coefficients measured from the gas generation spectra with heating after irradiation turn out to be smaller than the coefficients measured from the gas generation right during the irradiation. It has been established in measurements of the spectra of deuterium gas generation that the accumulation of deuterium in molybdenum occurs with a different binding energy of the deuterons in the metal. As the deuterium accumulates in the sample, those sites at which the deuterium is most strongly bound are gradually filled up and saturated. At large doses most of the deuterium is poorly retained in the metal and can be liberated into a vacuum either in the course of irradiation or during the subsequent heating. First of all, the deuterium is liberated most rapidly from the weakly bound states. It has been shown that in addition to thermal gas generation processes there is gas generation stimulated by the ion beam retarded in the sample. The contribution of radiatively stimulated gas generation at a low temperature and a high irradiation dose is especially important. A comparative analysis has been made of the results obtained by different authors, and the possible causes for the discrepancy between the experiments have been explained. It has been shown that the various results can be produced both by a difference in the irradiation parameters (which are sometimes assumed without justification to be unimportant and are not even indicated in the papers) and by a difference in measurement procedures.Translated from Atomnaya Énergiya, Vol. 57, No. 2, pp. 104–108, August, 1984.     

Effect of temperature on graphite oxidation behavior

The temperature dependence of oxidation behavior for the graphite IG-11, used in the HTR-10, was investigated by thermogravimetric analysis in the temperature range of 400–1200 °C. The oxidant was dry air (water content <2 ppm) with a flow rate of 20 ml/min. The oxidation time was 4 h. The oxidation results exhibited three regimes: in the 400–600 °C range, the activation energy was 158.56 kJ/mol and oxidation was controlled by chemical reaction; in the 600–800 °C range, the activation energy was 72.01 kJ/mol and oxidation kinetics were controlled by in-pore diffusion; when the temperature was over 800 °C, the activation energy was very low and oxidation was controlled by the boundary layer. Due to CO production, the oxidation rate increased at high temperatures. The effect of burn-off on activation energy was also investigated. In the 600–800 °C range, the activation energy decreased with burn-off. Results of low temperature tests were very dispersible because the oxidation behavior at low temperatures is sensitive to inhomogeneous distribution of any impurity, and some impurities can catalyse graphite oxidation.     

铀吸、放氘/氚的热力学同位素效应

实验测定了铀吸收氘和氚单质气体的p-c-T曲线及解吸氘和氚气体的p-t曲线,获得了不同温度下吸气和解吸的平衡压,根据范德荷夫方程得到了铀吸收和解吸氘、氚的热力学参数ΔHΘ和ΔSΘ。数据显示,铀吸收氘和氚气体时,在相同温度相同原子比下,吸氘的平衡压比吸氚的平衡压低,解吸氘的平衡压也比解吸氚的平衡压低,且存在明显的滞后效应。从反应的焓变和熵变来看,铀吸、放氘/氚气体时存在较小的热力学同位素效应。     

D+注入Li4SiO4的表面化学状态及其热解吸行为

锂陶瓷氚增殖剂的氢同位素行为是聚变堆固态产氚包层关心的重要课题。本文将3 keV D⁺注入Li₄SiO₄,采用X射线光电子能谱在线分析注氘前后材料表面的化学状态,同时采用热解吸谱（TDS）实验技术,研究注氘后Li₄SiO₄中氢同位素的热解吸行为。实验结果表明：D⁺注入会改变Li₄SiO₄表面的化学环境,产生多种辐照缺陷和化学键合状态;氘滞留量和热解行为受注氘时样品的温度影响较大,可在一定程度上预测产氚包层中氚的滞留行为。     

Measurement of the Characteristics of the Ventilation Pipe of the Filtration System for the Steam-Gas Medium of a Nuclear Power Plant

The dependence of the heat power of a pipe (air–air heat exchanger) on the flow rate and the temperature of the heating air is investigated experimentally. The temperature fields along the height of the pipe are measured. In the experimental velocity range 1.2–9.5 m/sec and heating-air temperature range 120–280°C, the heat power depends strongly on the temperature and negligibly on the flow rate of the heating air. Increasing the cold-air flow rate increases the thermal power in direct proportion.     

Vapor pressure of T2O

The procedure is described and the results of the determination of the vapor pressure of samples of high-percentage tritium water (containing 83.4 and 98.1 mole % T₂O) in the 12–95 °C temperature range are given. To calculate the pressure of the gaseous products of the radiolysis of water and nuclear transformations the measurements were carried out by the static method at two different volumes of the apparatus. It was found that the boiling points of HTO and T₂O are 100.8 and 101.6 °C, the heats of vaporization at these temperatures are 9.9 and 10.0 kcal/mole, and the standard entropies are 19.3 and 19.0 units, respectively.     

CLAM钢中氕氘的渗透与去除

低活化铁素体/马氏体钢（RAFM钢）是聚变堆产氚包层的优选结构材料。氢同位素在结构材料中的扩散渗透特性关系到产氚回收率、燃料循环及运行安全。本工作对国内研发RAFM钢之一的CLAM钢进行了气体驱动的氘渗透实验,得到573～873 K温度范围内氘的宏观溶解度S（mol/(m³•Pa^0.5)）为0.264exp(-22 447/RT),扩散系数D（m²/s）为1.38×10^-7exp(-17 271/RT),渗透率Φ（mol/(m•s•Pa^0.5)）为3.64×10^-8exp(-39 718/RT)。还进行了氕氘气体混合物的渗透实验,确认了渗透同位素效应;探索了钢中溶解氘的真空热释放去除。     

Effect of Internal Breeding of Tritium and Helium-3 on the Ignition of an ICF Fuel Pellet

Self-heating condition and following ignition in an Inertial Confinement Fusion (ICF) fuel pellet is evaluated by calculating the power equations, dynamically. In fact, the self-heating condition is a criterion that determines the minimum parameters of a fuel (such as temperature, density and areal density) that can be ignited. Deuterium is the main component of ICF fuels as large amounts of it are naturally available. In addition, the use of deuterium as a fuel in ICF causes the production of tritium and helium-3. However, pure deuterium has a high ignition temperature (\(\hbox {T}\ge 40\,\hbox {keV}\)) which makes it inefficient. In this paper, the power equations are solved, dynamically, and it has been indicated that internal tritium and helium-3 production at early evolution of compressed deuterium fuel causes ignition at lower predicted temperatures.     

Fusion of Multi-Pinched Plasma Beams Converging with Spatial Symmetry

Fusion reactions can be achieved by using deuterium from sea water as the fuel.The amount of deuterium in one gallon of sea water contains energy equivalent to three hundred gallons of gasoline.Satisfactory conditions of plasma temperature and density necessary to initiate fusion have been achieved in various research facilities.However,the confinement time is not sufficient for ignition due to plasma instabilities.Here we show that fatal plasma instabilities could be suppressed by the ingenious arrangement of multi-pinched plasma beams converging symmetrically in space based on the minimization principle of plasma potential energy.Confirmation tests are proposed using tiny wires containing deuterium.If successful,the results could lead to a feasible approach to obtaining commercial fusion power from sea water,hence without the need to use expensive and radioactive tritium as the fuel.