Development of Transuranium Element Recovery from High-Level Radioactive Liquid Waste

In the present study, the jet breakup characteristics of molten material is experimentally investigated in nonboiling condition using Wood's metal to isolate the key features of jet breakup phenomenon from the conjugated nature of melt breakup and steam generation. The experimental apparatus consists mainly of melt generating furnace and melt crucible equipped with variable nozzle diameter, a rectangular water tank of 350×350×800mm equipped with temperature controlling heater and thermocouples. The jet diameters were 10 and 20 mm and the jet velocity was varied by pressurizing the melt container. Wood's metal of 70°C melting temperature was used. Visualization of jet breakup provided characteristics of jet breakup in water. The range of jet velocity was 2.2–5.5 m/s. The debris were collected and sieved and it was shown that the debris sizes of 1.0–2.8 mm had the largest mass fraction, up to 50%. In the present experimental conditions, the Kelvin-Helmholtz instability is considered the most probable cause of jet breakup.     

自蔓延高温合成法固化高放核废料的研究

采用新型氧化剂CrO3,自蔓延高温合成了含Sr2 的钙钛矿.分析了自蔓延高温反应的理论基础,并以XRD、SEM和电感耦合等离子体原子发射光谱仪,分析合成产物的物相、显微结构及浸出率.结果表明:CrO3可将反应燃烧温度由常规1900K提高至2945K;当SrO含量为10wt%时,可以获得较好的钙钛矿物相;当SrO含量为20wt%时可获最低浸出率2.8×10-3g/m2·d.此外,不同配比固化体浸出率皆小于0.13×10-1g/m2·d.     

铂系金属沉积对高放废液玻璃固化的影响北大核心CSCD

高放废液中的钌(Ru)、铑(Rh)、钯(Pd)等铂系金属元素在玻璃固化过程中会发生团聚并沉积在熔炉底部。铂系金属的沉积会导致熔炉内局部的玻璃密度、熔体电导率和黏度发生变化,容易引起焦耳加热陶瓷熔炉的加热电极短路和出料口堵塞等问题,同时对玻璃固化体的质量产生影响。本文基于对国内外有关玻璃固化过程中铂系金属沉积机制及其性能影响研究进展的调研,综述了铂系金属在玻璃熔体中的结晶机理,分析观察其生长、团聚和分布规律;研究了铂系金属对玻璃熔体电导率和黏度影响的机理机制,对影响电导率和黏度的各种因素进行了总结;论述了铂系金属沉积对玻璃固化体核素浸出率的影响;最后提出和归纳了解决玻璃固化过程中铂系金属沉积的办法。     

A Screening Model for Calculating the Risks of Transporting High Level Radioactive Waste

Abstract

In transporting high level radioactive waste (including spent nuclear fuel), shippers (and sometimes carriers) need to evaluate the risks of potential radiation exposure to the public and transport workers. A simple model is presented that can be applied to nuclear waste transport risk assessments. The model considers radiation risks arising from incident free exposure, accidental release-caused exposure to on-link population, off-link population, crew, transport workers, etc. Important parameters and factors that affect the radiation dose level are grouped using the physics of the different exposure phenomena. The total radiation risk (in person-rems) is given by a linear combination of the groups of these factors, each representing a different type of exposure. The radiation exposure risk assessment is reduced to the evaluation of a single linear algebraic equation containing five distinct terms and each term containing the groups of parameters and a constant coefficient. The estimation of the values of the constant coefficients was accomplished by selecting a sample of 65 origin-destination (0-D) pairs and simulating the shipment of high level nuclear waste or spent nuclear fuel between each O-D pair, and evaluating the radiation dose risks to each group of population. RADTRAN 4 was used in the detailed assessments. The coefficient values were tested for statistical robustness using a sampling hypothesis and t-statistics. These values are presented. The simplified model presented here represents a viable and economical option as a radiological risk assessment tool, to be used in mode or route options screening.     

Accelerated Leach Tests of SRL-165 High-Level Waste Glass in Deionized Water

Accelerated short-term leach tests in a laboratory are neccessary in order to estimate, with reasonable accuracy, the long-term leaching behavior of high-level waste glass. In the present study, static leach tests of an SRL-165 high-level waste glass were carried out in deionized water at two different glass-surface-area to solution-volume ratios (SA/V-ratio), namely 0.85 and 0.079 cm^?1 at 90°C, and 0.85 cm^?1 at 40°C.

First, an equation was examined which related Si-concentrations with time, temperature and SA/V-ratio under the present static conditions. The parameter determined at 90°C, 0.85 cm^?1 can be used to calculate the Si-concentration at 40°C, 0.85 cm^?1. Second, at the low SA/V- ratio of 0.079 cm^?1, the concentrations of Ca and Mg in the leachates peaked and then decreased a little. The equation used above does not explain the variation of the concentrations of Ca and Mg at a low SA/V-ratio. The precipitation of Ca and Mg onto the glass surface is probably caused by the adsorption efficiency of the surface layer or the formation of crystalline materials at the low SA/V-ratio of 0.079 cm^?1. Third, the in-depth profiles of some elements obtained by secondary ion mass spectroscopy (SIMS) were qualitatively in agreement with the results of solution analyses. This indicates the particular usefulness of SIMS for analyzing leaching behavior of the glass in in-situ burial studies where solution analyses are often impractical.     

A Thermoelectric-Conversion Power Supply System Using a Strontium Heat Source of High-Level Radioactive Nuclear Waste

A thermoelectric-conversion power supply system with radioactive strontium in high-level radioactive waste has been proposed. A combination of Alkali Metal Thermo-Electric Conversion (AMTEC) and a strontium fluoride heat source can provide a compact and long-lived power supply system. A heat source design with strontium fluoride pin bundles with Hastelloy cladding and intermediate copper has been proposed. This design has taken heat transportation into consideration, and, in this regard, the feasibility has been confirmed by a three-dimensional thermal analysis using Star-CD code. This power supply system with an electric output of 1MW can be arranged in a space of 50m² and approximately 1.1m height and can be operated for 15 years without refueling. This compact and long-lived power supply is suitable for powering sources for remote places and middle-sized ships. From the viewpoint of geological disposal of high-level waste, the proposed power supply system provides a financial base for strontiumcesium partitioning. That is, a combination of minor-actinide recycling and strontium-cesium partitioning can eliminate a large part of decay heat in high-level waste and thus can save much space for geological disposal.     

Distribution of Elements in Phases Produced by Reaction between CaO-Stabilized Zirconia and Simulated High-Level Radioactive Waste

A study for identifying the relationship between turbulent parameters and local wall thinning inside a carbon steel piping was performed. Experiments and numerical analyses for several types of downscaled piping components were conducted, and the obtained results were compared. Based on the results indicating that flow behaviors inside piping components can be sufficiently simulated by numerical analysis, numerical analyses for the models magnified to the actual sizes of plants were carried out. To determine the relationship between turbulent parameters and local wear rates, numerical analyses were performed for 17 piping components of 7 types installed in a feedwater system. Turbulent parameters resulting from numerical analyses were compared with the local wear rates based on the measured thickness data. From thecomparison of the results, the vertical flow velocity component (V_r) flowing again to the wall after separation due to geometrical configurations or direct collision against the wall at an angle of some degrees was found to be analogous to the configuration of local wall thinning. From the least-squares fitting result, it was derived that the average relationship between V_r and the local wear rate is proportional to 0.55-fold V_r adding 0.1 to the standard deviation of 0.65.     

放射性废物安全的阶段性研究成果--《放射性废物安全通论》

放射性废物安全是辐射防护与辐射源安全科学同放射性废物管理技术相互作用的产物。在国际上,放射性废物安全已经成为辐射防护与辐射源安全科学发展的前沿领域之一。《放射性废物安全通论》是以文集形式即将出版的一本专著,是近20年来笔者等在此领域进行学习、思考、实践、讨论和总结的一个记录。本文综述了该书的要旨。     

高放废物处置评价中应用的有关程序

对有关在高放废物处置中应用的程序进行了调研,根据应用方向将这程序分成源顶计算程序、地下水与核素迁移程序、热－湿－力耦合程序及其它程序4类。     

Clearance for Radioactive Waste

正The clearance plays an important role in radioactive waste management.It is an effective tool for waste minimization and also an important application of optimization of radiation protection in waste management.The study of clearance is significant to make waste management legitimate,safe and economical.     

Radioactive Material Package Seal Tests

Abstract

General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the United States (US) Code of Federal Regulations Part 71. The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high and low temperature environments under dynamic and static conditions. A seal technology programme, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this programme to characterise the behaviour of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer's low temperature ratings are based on methods that simulate this use. The seal materials tested in this programme with a fixture similar to a RAM cask closure, with the exception of silicone S613–6O, are not leak tight (1.0 × 10–7 std cm³.s^?1) at manufacturer's low temperature ratings.     

最少量化在实验研究产生的放射性废物管理中的实施

通过源项的引入控制、实践正当化、安全与防护最优化、废物产生的课题化管理、分类收集分别处理等措施的实施，落实废物最少量化在核科研中的实施，减少对环境的危害，降低废物处理处置费用，促进核技术的发展应用。     

对法国高放废物处置安全审评的分析研究

介绍了法国高放废物处置研究现状和规划,对法国高放处置场的审评技术单位法国核与辐射安全研究院（IRSN）所开展的高放处置安全研究和审评工作及其提出的审评原则和审评要点进行了分析研究,并对我国的高放处置安全审评工作提出了建议。     

Radioactive Waste Information Management System

According to the requirement and current situation of national industrial radioactive waste management and disposal, present work develops drummed radioactive waste information management system based on network. The system may provide effective real-time management both for administration and enterprise. And the ）,-detecting classification facility of barrel radioactive waste is also prepared.     

放射性废物中等深度处置

放射性废物中等深度处置是介于近地表处置和地质处置的一种处置方式。本文介绍了国际上放射性废物中等深度处置的发展概况及我国可能适于中等深度处置的废物源项,提出我国实施放射性废物中等深度处置需考虑的问题及今后开展相关工作的建议。     

Cement Solidification of Radioactive Waste Liquid

Withdecommissionofthelaboratoryofneutronactivityanalysis,radioactivewasteliquidproducedneedbetreatedurgently.Inthewasteliquidthemajornuclidesare60Coand152Eu,thesaltcontentis250g/Landtotalactivityis2×107Bq.Aftersufficientdemonstration,thewastewaterissolid…