放射性污染金属废物包装容器的屏蔽设计

针对活度较高的放射性污染金属废物包装容器的屏蔽设计进行了研究。重点阐述了计算废物货包外γ剂量率的圆柱体源模型,推导了圆柱体源剂量场分布的计算公式。通过圆柱体源模型对Ⅷ型钢箱废物货包的计算值与监测数据对比分析,表明圆柱体源模型计算结果能较准确地反映废物货包外γ剂量率分布。基于圆柱体源模型的计算结果,对废物包装容器进行了合理的屏蔽设计,满足了放射性物质运输和处置标准的有关要求。圆柱体源模型对单体货包外剂量率的计算方法简单易实现,而且其计算值比监测值稍大,提供了一定的安全裕度,适合应用于放射性工程实践中。     

Stainless steel containers for the storage of low and medium level radioactive waste

Corrosion investigations have been performed on the austenitic structural steel AISI 304L, in comparison with the structural steel AISI 316L, in an aerated and a de-aerated solution, which was leached from low and medium level radioactive waste. On the basis of measured potentio-dynamic anodic polarization curves and the results of cyclic polarization tests, it was found that both types of steel, as well as the corresponding welds, had a high pitting potential and a high protective potential, which means that they have a strong tendency to form a compact and corrosion-resistant passive film. The repassivation capability of both types of steel prevents the occurrence of stress corrosion cracking at the level of concentration of chloride ions which corresponds to the described type of waste, whereas absorbed atomic hydrogen does not reduce toughness or cause hydrogen embrittlement. The results of the research work confirmed that it is possible to use AISI 304L structural steel for the construction of containers for the temporary, 30-year storage of low and medium level radioactive waste.     

Water hyacinth for phytoremediation of radioactive waste simulate contaminated with cesium and cobalt radionuclides

Phytoremediation is based on the capability of plants to remove hazardous contaminants present in the environment. This study aimed to demonstrate some factors controlling the phytoremediation efficiency of live floating plant, water hyacinth (Eichhornia crassipes), towards the effluents contaminated with ¹³⁷Cs and/or ⁶⁰Co. Cesium has unknown vital biological role for plant while cobalt is one of the essential trace elements required for plant. The main idea of this work i.e. using undesirable species, water hyacinth, in purification of radiocontaminated aqueous solutions has been receiving much attention. The controlling factors such as radioactivity concentration, pH values, the amount of biomass and the light were studied. The uptake rate of radiocesium from the simulated waste solution is inversely proportional to the initial activity content and directly proportional to the increase in mass of plant and sunlight exposure. A spiked solution of pH ≈ 4.9 was found to be the suitable medium for the treatment process. The uptake efficiency of ¹³⁷Cs present with ⁶⁰Co in mixed solution was higher than if it was present separately. On the contrary, uptake of ⁶⁰Co is affected negatively by the presence of ¹³⁷Cs in their mixed solution. Sunlight is the most required factor for the plant vitality and radiation resistance. The results of the present study indicated that water hyacinth may be a potential candidate plant of high concentration ratios (CR) for phytoremediation of radionuclides such as ¹³⁷Cs and ⁶⁰Co.     

Vitrification of high-level radioactive waste considering the behavior of platinum group metals

To demonstrate a method using liquid metal for the removal of PGMs during the vitrification process of high-level radioactive waste, removal of Pd was performed using Cu from molten glass containing fission products such as Nd, Sr, Zr, Mo, Te and Ni. Almost all the Pd, 93%, was extracted into liquid Cu and removed as a separable Cu–Pd metal button from the glass. Tellurium and Ni were also extracted 42 and 5.6%, respectively. Nearly 100% of the other elements, especially the heat generating elements such as Sr and Cs, for which Na was used as a substitute of Cs remained in the glass.     

A new method for monitoring the radiation damage in nuclear waste containers using ion channelling

Immobilizing nuclear wastes has been one of the most important challenges in nuclear technology. A method to quantify and monitor the radiation damage to waste immobilizing crystalline materials like zircon is proposed. This method will make use of proton/ion channelling measurements of the crystalline containment sample or test crystalline sample placed in the crystalline or amorphous containment of nuclear waste for a long time from years to a few decades and the mathematical method to determine the structure collapse rate of the containment material using channelling measurements. Implementation procedure of this method/technique for radiation damage measurement in nuclear waste container materials is described.     

The EIX process for radioactive waste treatment

The use of an electric potential in place of chemical reagents to elute ion-exchange media by ion migration, as well as perform redox and pH changing reactions is of considerable importance in simplifying subsequent processing stages, such as, the recovery or disposal of the concentrate. The process has become known as Electrochemical Ion Exchange (EIX). By making multiple and complete use of ion exchange capacity in this way, large volume reduction factors can be achieved for only <5% (depending on the feed concentration) of the energy required for evaporation.

AEA Technology has developed EIX over the last decade from concept to pilot demonstration, and during the last five years has gained vast experience in the design, build and operation of pilot systems (often as one stage integrated into a larger processing plant) for a variety of applications. The principal application areas for EIX relate to the removal/recovery of metals from aqueous streams[1]. These include: the decontamination of streams produced by Pressurised Water Reactors (PWRs) and the separation of radioactivity from tank waste storage facilities, the selective removal and recovery of precious metals from metal refining operations and base metals from electroplating rinse waters, the treatment of hospital waste (active) streams and the removal of silver from photographic wash waters[2]. The process has also been used successfully to remove toxic heavy metals, such as arsenic, from drinking water supplies.

The technology has proven to be both robust and reliable. Direct comparison with ion exchange and evaporation has demonstrated significant cost savings for the majority of applications listed above.     

Safety requirements for the design of large containers for decommissioning waste

About 50% of the total volume of conditioned radioactive waste from nuclear power generation will finally result from the decommissioning of nuclear power plants (NPPs). Higher activated metallic waste from the core region of the reactors would, according to the International Transport Regulations (IAEA), require a type B container. This would, however, bring a significant increase in the costs of the management of such waste. A considerably cheaper solution of this problem can be achieved by separating the protection requirements for type B packaging (i.e. mechanical integrity and tightness). By using industrial packaging (IP) designed for a higher mechanical integrity, it is possible to cope with higher IAEA protection goals for the safe transport of dismantling waste without strictly following the extremely high and also expensive tightness requirements for type B packaging. This is because the radioactivity is bound in the metallic lattice of such activated decommissioning waste.The above mentioned strong IP for decommissioning waste can also be used for other highly activated waste from the core region (e.g. fuel element boxes, control rods or other activated equipment) during the operation of the NPPs.     

A new approach to radioactive waste self-burial using high penetrating radiation

The method of self-burial of radioactive waste in geological formations using direct heating of rocks by radiation is proposed in this paper. In the currently known studies, thermal conductivity is considered as a main heat transfer mechanism. Application of high penetrating gamma radiation for direct melting of surrounding rocks will reduce the energy absorption inside the sinking device and will lower maximum temperature and temperature gradients in the elements of the device. In this paper, conditions of realization of the direct heating by radiation mechanism are presented and requirements to heat-generating radionuclides have been derived. Assessments of the spatial distribution of energy release in the surrounding rocks for the point and plane sources with the radionuclide ⁶⁰Co have been performed. Based on these data, the temperature distributions in the surrounding rocks and the expression for determining the descent velocity as a function of ⁶⁰Co surface activity in the sinking device have been obtained. Estimations of energy absorption fraction inside the spherical heat-generating elements filled with ⁶⁰Co and surface activity of ⁶⁰Co, necessary to achieve velocity of about 1 km per year, have been made. The results are given for granite and salt rocks.     

高放废物地质处置中的工程材料

凡人类从事于与核材料有关的许多生产、生活活动均可能产生不同活度的放射性废物.高放废物由于具有放射性水平高、发热量大、核素寿命长等特点,其安全处置倍受全球科学家和广大公众所重视.目前深地质处置被国际上公认为处置高放废物的最有效可行的方法.借鉴已有研究成果,我国采用多重工程屏障系统(包括废物固化体、废物罐及其外包装和缓冲/回填材料)和适宜的地质围岩地质体共同作用来确保高放废物与生物圈的安全隔离.参照国际上该领域的研究成果,结合我国处置概念,本文就高放废物地质处置中的工程材料(废物固化体、废物罐、外包装、缓冲材料、回填材料),以及其材料选择、设计要求和研究重点等进行了总结.     

Regional system for registration and monitoring radioactive waste

A regional structure is proposed for the State registration and monitoring of radioactive waste for the Moscow region. Consideration is given to the problems, the information communications, the role of an interregional center for information analysis, and to solving the problem of the registration of radioactive waste as part of this system. An approach is proposed to monitoring the formation of waste which is based on ensuring conditions for not exceeding the temporary storage norms and for the correct time for removing radioactive waste. An algorithm is given for the efficient planning of its removal taking account of the requirements for radiation safety. Scientific-Production Association MosNPO “Radon”, Moscow. Translated from Atomnaya énergiya, Vol. 86, No. 6, pp. 465–470, June, 1999.     

An improved segmented gamma scanning for radioactive waste drums

In this paper, the equivalent radius of radioactive sources in each segment is determined by analyzing the different responses of the two identical detectors, and an improved segmented gamma scanning is used to assay waste drums containing mainly organic materials, and proved by an established simulation model. The simulated radioactivity distributions in homogenous waste drum and an experimental heterogeneous waste drum were compared with those of traditional segmented gamma scanning. The results show that our method is good in performance and can be used for analyzing the waste drums.     

The matrix method for radiological characterization of radioactive waste

Beam losses are responsible for material activation in some of the components of particle accelerators. The activation is caused by several nuclear processes and varies with the irradiation history and the characteristics of the material (namely chemical composition and size). Once at the end of their operational lifetime, these materials require radiological characterization.The radionuclide inventory depends on the particle spectrum, the irradiation history and the chemical composition of the material. As long as these factors are known and the material cross-sections are available, the induced radioactivity can be calculated analytically. However, these factors vary widely among different items of waste and sometimes they are only partially known.The European Laboratory for Particle Physics (CERN, Geneva) has been operating accelerators for high-energy physics for 50 years. Different methods for the evaluation of the radionuclide inventory are currently under investigation at CERN, including the so-called “matrix method”. This paper provides a mathematical formulation of the matrix method highlighting its advantages and limits of validity.     

Electrochemical method of reprocessing radioactive waste

A. A. Bochvar All-Union Scientific-Research Institute of Nuclear Materials. Atomflot Radiotechnical Industry. Translated from Atomnaya énergiya, Vol. 78, No. 1, pp. 37–41, January, 1995.