Some equilibria in the uranium-plutonium-nitrogen ternary system. An assessment

Some assessments of possible equilibria for the ternary system uranium-plutonium-nitrogen have been made and from these equilibria the pressures of the gas phase species. Pu, U and N₂ for the system have been calculated. The nature of the vaporization behaviour of alloys of the system is also predicted.     

Damage parameters of structural materials in fusion environment compared to fission reactor irradiation

Calculations were performed to quantify the damage parameters in the leading candidate structural and plasma facing materials when used in magnetic and inertial confinement fusion systems and when irradiated in fission reactors. The structural materials considered are ferritic steel, austenitic steel, vanadium alloy and SiC/SiC composite. Plasma facing materials included beryllium, tungsten, and carbon fiber composites. Atomic displacement damage and gas production rates are greatly influenced by the neutron energy spectrum. For the same neutron wall loading, atomic displacement damage is slightly lower in inertial fusion systems than in magnetic fusion systems but gas production is about a factor of 2 lower. In addition, much lower gas production is obtained in samples irradiated in fission reactors. The results help guide irradiation experiments in fission reactors to properly simulate the damage environment in fusion systems and facilitate extrapolating to the expected material performance in fusion systems.     

The ternary system: silicon-uranium-vanadium

Phase equilibria in the system Si-U-V were established at 1100 °C by optical microscopy, EMPA and X-ray diffraction. Two ternary compounds were observed, U₂V₃Si₄ and (U_1−xV_x)₅Si₃, for which the crystal structures were elucidated by X-ray powder data refinement and found to be isotypic with the monoclinic U₂Mo₃Si₄-type (space group P2₁/c; a = 0.6821(3), b = 0.6820(4), c = 0.6735(3) nm, β = 109.77(1)°) and the tetragonal W₅Si₃-type (space group I4/mcm, a = 1.06825(2), c = 0.52764(2) nm), respectively. (U_1−xV_x)₅Si₃ appears at 1100 °C without any significant homogeneity region at x ∼ 0.2 resulting in a formula U₄VSi₃ which corresponds to a fully ordered atom arrangement. DTA experiments clearly show decomposition of this phase above 1206 °C revealing a two-phase region U₃Si₂ + V₃Si. At 1100 °C U₄VSi₃ is in equilibrium with V₃Si, V₅Si₃, U₃Si₂ and U(V). At 800 °C U₄VSi₃ forms one vertex of the tie-triangle to U₃Si and V₃Si. Due to the rather high thermodynamic stability of V₃Si and the corresponding tie-lines V₃Si + liquid at 1100 °C and V₃Si + U(V) below 925 °C, no compatibility exists between U₃Si or U₃Si₂ and vanadium metal.     

聚变反应堆含氚废水处理系统的设计

以联合电解催化交换-气相色谱(CECE-GC)为技术路线基础,对聚变反应堆(International thermonuclearexperimental reactor,ITER)含氚废水处理系统(Water detritiation system,WDS)进行了总体设计和主要子系统的设计.与目前的重水提氚演示系统相比,ITER-WDS的不同之处在于不使用氢氧复合器,不采用碱性电解池而使用固体聚合物电解池(Solid polymer electrode,SPE),增加了Pd/Ag膜渗透系统进行氚的回收.     

Constitutional studies on carbide nuclear fuels for fast breeder reactors: II. The uranium-praseodymium-carbon ternary system

Investigations have been carried out at 1400°C on the ternary system U-Pr-C and the major phase fields have been established. A considerable degree of solid solubility was found and the limits established are reported.     

The role of dispersed barriers in the pulsed irradiation creep of magnetic fusion reactor materials

The effects of irradiation pulsing on the climb-glide creep, and the role of the glide barrier height are investigated. Only tokamak-type pulsing is considered. We develop a new formulation for the creep strain increment per pulse for large barriers, for which more than one pulse is needed for glide to occur. This formulation is applied to typical tokamak-type conditions, including the UWMAK-I and INTOR designs. It is concluded that no significant enhancement over steady irradiation occurs for τ_v ? burn-time. However, in long burn-time Tokamaks with τ_v?on-time and off-time, it is found that the pulsed creep enhancement can be significant. For example, for a duty factor of 0.9 the enhancement is about 3 for small barriers using a dose-equivalent average damage rate when comparing pulsed and steady irradiation. The maximum enhancements are diminished to about 2 when equal instantaneous damage rates are used.     

Thermodynamic analysis of phase equilibria in the iron-uranium-zirconium system

Thermodynamic analysis of phase equilibria in metallic nuclear reactor systems has progressed to the study of the iron-uranium-zirconium ternary system. This work is based on our previous assessments and optimizations of the constituent binary subsystems iron-uranium, iron-zirconium, and uranium-zirconium.     

聚变堆用典型金属工程材料辐照损伤机理

聚变堆中极端辐照环境下,核工程材料的安全与可靠性对保障核能产业可持续发展具重大意义。采用蒙特卡罗程序包Geant4建立了聚变堆辐照环境下的材料损伤模型,从材料的位移损伤率和杂质沉积等方面研究了CLAM钢、F82H钢、?-Fe三种聚变堆用典型金属工程材料分别在中子、质子、重离子轰击下的辐照损伤机理。研究表明,中子对材料的辐照损伤主要为位移损伤;质子和重离子对材料造成的位移损伤呈Bragg峰曲线分布,且损伤区域与粒子射程均集中在材料表层,其中14.67 Me V质子射程为512?m,0.82 Me V 3He离子射程仅为2.1?m。系统分析了聚变堆用典型金属工程材料的损伤形成机理,为进一步研究材料受辐照后宏观性能与微观结构变化提供了理论依据。     

Chemical thermodynamics of fusion reactor breeding materials and their interaction with tritium

Liquid lithium, lithium alloys (solid and liquid) and ceramic lithium compounds are candidate breeding materials for (D,T) fusion reactors.Besides their tritium breeding capability, which results from neutron capture, their thermochemical properties and their interaction with tritium are of particular interest. A good knowledge of the physical and chemical properties of liquid lithium exists; and the systems Li-LiH, Li-LiD and Li-LiT have been studied in great detail. For dilute solutions of D₂ in liquid lithium, Sieverts' law was found to be valid down to an atom fraction of $\text{x}{}_{\mathrm{D}}\text{= 10}{}^{-6}$; in the vapor, lithium polymers up to Li₄ and lithium deuterides are found.In the system liquid Li-Pb, the solubility of D₂ was measured as a function of temperature and alloy composition, and correlated with the activities of the constituent metals. The solubility of D₂ was found to obey Sieverts' law at low concentrations, and is many orders of magnitude smaller than that in liquid lithium. This holds also for solid “Li₇ Pb₂”.Vaporization studies yielded data on the thermal stability of the oxides: Li₂0, γ-LiAlO₂, β-Li_sAlO₄, LiAl₅O₈, Li₂ZrO₃, Li₄ZrO₄, Li₈ZrO₆, Li₂SiO₃ and Li₄SiO₄. Tritium diffusivity was studied in Li₂O, γ-LiAlO₂, β-Li₅AlO₄ and Li₄SiO₄. A large number of gaseous lithides were detected during these studies.     

Investigation of advanced materials for fusion alpha particle diagnostics

Fusion alpha particle diagnostics for ITER remain a challenging task. Standard escaping alpha particle detectors in present tokamaks are not applicable to ITER and techniques suitable for fusion reactor conditions need further research and development [1], [2]. The activation technique is widely used for the characterization of high fluence rates inside neutron reactors. Tokamak applications of the neutron activation technique are already well developed [3] whereas measuring escaping ions using this technique is a novel fusion plasma diagnostic development. Despite low alpha particle fluence levels in present tokamaks, promising results using activation technique combined with ultra-low level gamma-ray spectrometry [4] were achieved before in JET [5], [6].In this research work, we use new advanced detector materials. The material properties beneficial for alpha induced activation are (i) moderate neutron cross-sections (ii) ultra-high purity which reduces neutron-induced background activation and (iii) isotopic tailoring which increases the activation yield of the measured activation product.Two samples were obtained from GERDA[7], an experiment aimed at measuring the neutrinoless double beta decay in ⁷⁶Ge. These samples, made of highly pure (9 N) germanium highly enriched to 87% in isotope Ge-76, were irradiated in real D–D fusion plasma conditions inside the TEXTOR tokamak. Comparison of the calculated and the experimentally measured activity shows good agreement. Compared to previously investigated high temperature ceramic material [8], this candidate detector offers better prospects for signal to background S/B ratio, energy resolution and particle selectivity due to a unique alpha particle signature. Applicability to ITER is discussed. Finally, research needs for further development of this diagnostic technique are outlined.     

The Fort St. Vrain high temperature gas-cooled reactor: II. Helium circulators

In field tests in a fossil-fueled facility, performed concurrently with Fort St. Vrain's construction, data indicated that the helium circulator design was well suited to provide primary coolant circulation for the high temperature gas-cooled reactor. After plant installation, primarily during the hot functional tests, a number of time-consuming delays developed caused by cavitation damage on circulator speed valves, cavitation and fatigue damage on auxiliary water turbine buckets, water turbine nozzle erosion, static shutdown seal cracks and circulator primary closure helium leakage. After extensive analysis and testing, all of these problems were corrected. Circulators have performed satisfactorily at levels up to 70% of rated power.     

Use of 4.7 MeV alpha particles in elemental analysis and fusion reactor materials studies

The possibilities for using 4.7 MeV alpha particles produced at the U-120 CIP Cyclotron for charged particle induced X- and gamma-ray emission applications (PIXE and PIGE, respectively) and for fast neutron radiation damage simulation are presented.The combined analysis using PIGE and PIXE methods either separately or simultaneously is an excellent means of determining the relative abundances of lighter elements with gamma-ray spectra and of heavier elements (Z ⩾ 16) with X-ray spectra.Some aspects of surface deformation effects by neutrons were simulated by means of medium-energy helium ions. An investigation of three types of commercial stainless steel (Romanian W 4016, Soviet 12KH18N10T and Japanese W 4541) was started using 3.0 [1], 4.7 and 6.8 MeV helium ions. The main post-irradiation effects observed are discussed.     

Foreseeable thermal, mechanical, and materials engineering problems of fusion reactor power plants

The engineering problems that will be posed by full-scale fusion reactor power plants are illustrated by examining a representative conceptual design. The temperature extremes that must be accommodated run from 4 K in superconducting magnets to 10⁸ K in the plasma. These temperature differences lead to difficult problems with differential thermal expansion, high heat fluxes, and stringent thermal insulation requirements. The magnetic fields that must be provided run from 25 to 100 kG, and these fields induce forces on elements of the structure of the order of 20 000 tons. The walls of the chamber containing the plasma must withstand intense radiation by 14 MeV neutrons and 1–50 keV ions. Unusual fluid flow and heat transfer problems include two-phase boiling flow of helium in the super-conducting magnets, and the magnetohydrodynamic effects on the flows of red hot lithium and boiling potassium in a high magnetic field. These and many other problems must be solved in such a way as to give a reliable, safe system at a reasonable capital cost, and this must be done with materials whose nuclear, physical, and fabrication properties and resistance to corrosion meet all of the requisite boundary conditions.     

Void-swelling in irons and ferritic steels: II. An experimental survey of materials irradiated in a fast reactor

The void-swelling response of a wide range of ferritic alloys irradiated in the Dounreay Fast Reactor to displacement doses up to 30 dpa (N/2) and covering the temperature range 380–615°C have been compared. The materials selected included high purity irons, together with commercial mild and low alloy steels, high chromium (12–14%) ferritic and martensitic stainless steels and a range of high purity binary iron-chromium alloys containing chromium contents up to 15%. The pure irons and binary iron-chromium alloys exhibited measurable but relatively low swellings (<1%) whilst all the commercial ferritic steels appeared to be void-swelling resistant, with swellings below the experimental detection limit (0.1%). The pattern emerging is thus one of overall swelling resistance in ferritic materials as a general class. Void-swelling in the pure iron peaked at two irradiation temperatures (~420, ~510°C), and the low magnitude of the swelling was rationalized in terms of the operation of solute-controlled swelling suppression mechanisms involving residual interstitial impurities. The complex functional dependence of peak-swelling on chromium content in the binary iron-chromium alloys was explained in terms of void-swelling suppression based on the presence of weak interactions between chromium atoms in solution and vacancies, modified by depletion of chromium from solid solution by α' precipitation at chromium contents exceeding 10%. The validation of the high swelling resistance of the 12% Cr martensitic stainless steels in a fast reactor environment provides confidence in the selection of these alloys as alternative core component materials for commercial fast reactor systems.     

聚变堆氚增殖层中子学分析

D-T聚变堆包层的主要功能包括氚增殖、能量转换射层蔽等,包层中子学设计的主要原则是满足聚变堆的氚自持,一般要求包层氚增殖比TBR＞1.1.使用与时间有关的扩散理论和本征函数展开方法,研究不同几何线度、6Li丰度的LI2O、LiPb包层材料14MeV源下的系统通量、氚增殖比影响,及在不同6Li丰度下6Li、7Li造氚随时间变化的规律.计算中使用了30群截面数据,微观数据来自ENDF/B-VI及JEF-2.2.