An investigation of uranium sorption on minerals of the surrounding rock

Present technological methods do not ensure complete extraction of uranium from ores, because a definite amount of it remains in the solid part of the residue, even under conditions of severe leaching. The bulk of the over-all uranium losses during hydrometallurgical processing of ores is due to this fact.This paper gives experimental data confirming that one of the causes of uranium losses with the solid part of the residue is its sorption on minerals of the surrounding rock. The sorption capacity of such minerals as montmorillonite, kaolinite, bauxite, and albite, and the possibility of uranium desorption by solutiom of various salts and acids, were investigated.Translated from Atomnaya Énergiya, Vol. 16, No. 1, pp. 51–55, January, 1964     

Nuclide release from an HLW repository: Development of a compartment model

A compartment model, ACGEO, has been developed for the Korean reference Repository System (KRS) for an HLW disposal by utilizing AMBER, a general compartment modeling tool. ACGEO is a flexible and adaptable tool, by which a transient calculation of both a nuclide release and transport in the near- and the far-field of a repository system with various and complex shapes can be carried out. It is expected to be used for the treatment of nuclide transport of a decay chain for safety assessment, as well as for design feedback on the KRS both on a deterministic and a probabilistic bases. Some illustrative cases are also investigated in which the influence of varying the near-field system features on the release of selected nuclides for an HLW repository are examined, in order to demonstrate the usability of ACGEO for feedback to a refined repository concept and for establishing an appropriate safety margin for an HLW repository.     

Modelling the effect of diffusion into the rock matrix on radionuclide migration

Diffusion into the rock matrix is potentially an important retardation mechanism for nuclides leached from an underground radioactive waste repository in a fractured hard rock. Models of this diffusion process are discussed and incorporated into three-dimensional radionuclide migration models. Simple solutions to these models are derived for two regions: the region near to the repository where the nuclide is diffusing into effectively infinite rock, and that much further downstream where the concentrations in the rock and fractures are almost in equilibrium. These solutions are used to evaluate the possible impact of migration. It is shown that retardation factors in excess of 100 and reductions in the peak concentration at a given point on the flow path by three or four orders of magnitude are possible for non-sorbed ions, which would otherwise be carried by the flow and not retarded at all.     

Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations

One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel.Therefore, our attention is focused on understanding and modelling the barrier performance of the different parts of an HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand this behaviour, it is necessary to start with investigations of unirradiated material, and to proceed with experiments with external gamma irradiation to determine the effects of oxidising radiolysis species. Further experiments with irradiated material have to be performed to investigate the influence of the irradiation damage, and finally an investigation has to be made of the irradiated material plus additional gamma irradiation. Experimental data are now available for the diffusive transport of radionuclides in the water-saturated graphite pore system, the corrosion rates of unirradiated graphite with and without external gamma irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO₂ and (Th,U)O₂ fuel kernels with and without external gamma irradiation. All investigations were performed in aquatic phases from salt, granite, and clay host rock.     

Colloid-facilitated effects on migration of radionuclides in fractured rock with a kinetic solubility-limited dissolution model

<正>Nuclides can move with groundwater either as solutes or colloids,where the latter mechanism generally results in much shorter traveling time as the nuclides interact strongly with solid phases,such as actinides.In the performance assessment,it is therefore essential to assess the relative importance of these two transport mechanisms for different nuclides.The relative importance of colloids depends on the nature and concentration of the colloids in groundwater.Plutonium(Pu),neptunium(Np),uranium(U) and americium(Am) are four nuclides of concern for the long-term emplacement of nuclear wastes at potential repository sites.These four actinides have a high potential for migrating if attached to iron oxide,clay or silica colloids in the groundwater.Strong sorption of the actinides by colloids in the groundwater may facilitate the transport of these nuclides along potential flow paths.The solubility-limited dissolution model can be used to assess the safety of the release of nuclear waste in geological disposal sites.Usually,it has been assumed that the solubility of the waste form is constant.If a nuclide reaches its solubility limit at an inner location near the waste form,it is unlikely that the same nuclide will reach its solubility limit at an outer location unless this nuclide has a parent nuclide.It is unlikely that the daughter nuclides will exceed their solubility limit due to decay of their parent nuclide.The present study investigates the effect of colloids on the transport of solubility-limited nuclides under the kinetic solubility-limited dissolution(KSLD) boundary condition in fractured media.The release rate of the nuclides is proportional to the difference between the saturation concentration and the inlet aqueous concentration of the nuclides.The presence of colloids decreases the aqueous concentration of nuclides and,thus,increases the release flux of nuclides from the waste form.     

Model of nanotube formation in solution under molecular sorption conditions

Reliable disposal of radioactive wastes in nanosize containers which ensure that the wastes are isolated and strongly contained is investigated. A model of the process of incorporating impurity components into solutions where nanotubes are formed is presented to substantiate the possibility considered. An attempt is made to describe the rolling up of nanotubes in a liquid medium from the standpoint of the energy balance in a system that consists of a nanometer-size thin film and molecules which are sorbed on its surface from solution. The results show that when molecules are sorbed from solution onto a nanolayer surface the layer rolls up. The possibility that a roll or nanotube loses stability when sorbed molecules are present in them must be taken into account when studying the final state of the nanolayer.     

Progress of a cost optimization for an HLW repository in Korea

This paper provides the results of a cost optimization for a CANDU spent fuel canister as well as the operational duration of an HLW repository. From the design change of an advanced-CANDU spent fuel canister, the overall costs were expected to be reduced by 124 MEUR in the case of disposing of 36,000 tU in an HLW repository, and it was also found that the optimal operational duration for an HLW repository was 83 years, to minimize the total cost. But this operational duration was only calculated from the aspect of cost benefits with economics' perspectives.We confirmed that the canister and operational duration are the dominant cost drivers for surface facilities and underground facilities for a cost optimization, respectively. Especially, the manufacturing method of an outer canister using the cold spray coating technique which was developed through collaboration with a domestic company is suggested to minimize the overall costs.     

Copper canister lifetime limited by a sulfide intrusion in a deep geologic repository

Korea has continuously implemented an ambitious nuclear energy deployment program since 1978. Korea currently operates 20 units, 16 PWRs and four CANDUs and constructs four and reviews license application of two more units. Also, Korea plans to build two more units by 2016. In addition, according to the new “Green Growth Plan while reducing the emission of carbon dioxide” Korea will introduce 10–12 units by 2030. This will inevitably result in more burdens on the safe management of spent nuclear fuels. Korea Atomic Energy Research Institute has developed a final disposal concept for Spent Nuclear Fuel (SNF) named KRS. KRS proposes to emplace SNF in a deep geologic formation such as a crystalline rock. Two key engineered barriers are applied to retard the potential release of a radionuclide from an embedded SNF; a waste container and an engineered barrier. Such an engineered barrier is composed of domestic calcium bentonite and the waste container is composed of an outer copper layer and an inner steel layer. The outer layer, a copper layer is dedicated to protect a waste container against corrosion. The main corrosion mechanism to corrode a copper waste container is a pitting whose speed of corrosion is 5–25 times higher than that of a uniform corrosion. In this paper, a special mass transfer resistance model is developed to predict the migration of sulfide from a fracture to a waste container surface via a bentonite layer. Based on it the lifetime of a copper canister layer limited by a pitting corrosion is estimated. Results show that under normal conditions, a copper layer can sustain its integrity for up to more than millions of years.     

Analytic solutions of radionuclide transport with the limited diffusion from the fracture into a porous rock matrix

To identify the effect of the limiting zone for the matrix diffusion from an open fracture into surrounding medium recently found from field studies, full analytic solutions are derived. The solutions are then verified with existing solutions and a numerical inversion of the Laplace transform approach. The relevant field data are collected and then numerical evaluations are pursued for these complicated analytic solutions with multiple integrals. Numerical results illustrate that for given data sets the effect of limiting zone for the rock matrix diffusion can be negligible so that the current infinite rock model may underestimate the radionuclide transport through fractured porous rock.     

A collisional–radiative cooling model for light impurity elements in hot plasmas under non-equilibrium conditions

The power radiated by low-temperature high-density impurity inclusions in plasmas, such as the clouds surrounding ablating impurity pellets, or the vapor layers evolving over vaporizing surfaces subjected to high-temperature plasmas, is calculated by means of a collisional–radiative (CR) model without the usual assumptions of equilibrium conditions. The populations of the ionization levels are determined by finite-rate calculations and, due to the much shorter characteristic times involved, instantaneous relaxation is assumed for the intermediate excitation levels considered. Data obtained with this model in the low-density limit are compared with those of Post–Jensen corona model [D.E. Post, R.V. Jensen et al., At. Data and Nucl. Data Tables 20 (1977) 397; R. Clark, J. Abdallah, D.E. Post, J. Nucl. Mater. 220–222 (1995) 1028; D.E. Post, J. Nucl. Mater. 220–222 (1995) 143]. Results of representative scenario calculations pertaining to the ablation of carbon and neon pellets are presented.     

单轴载荷下岩石核磁共振特征的实验研究

对岩石在单轴载荷下的核磁共振(NMR)特征进行了实验研究.选取两组砂岩和一组人工烧结陶瓷实验样品,在轴向施加压力作用改变岩石的孔隙结构,然后进行核磁共振谱测量,分析岩石在加压前后核磁共振谱和谱面积的变化特征.实验结果表明,由于外加载荷改变了岩石的孔隙结构,岩石的核磁共振谱产生明显的变化,其中横向弛豫时间T2谱面积是反映岩石孔隙结构变化的一个重要参数.     

Migration processes in a model deep repository for the disposal of intermediate-level nuclear waste

It is difficult to conceive of radionuclides escaping from a repository by any means other than migration in groundwater. Simple models of the repository are constructed and various migration processes are identified and assessed, according to the flow speed of water through the repository. Diffusion in static water and advection in fast flows are considered separately initially, but later we examine the effect of slow flows in which both these processes contribute to the removal of radionuclides. Concentration profiles across the repository, fluxes of nuclides and total losses are obtained from the analysis. We investigate the time scales necessary for the steady state to be achieved in the repository and conclude that flow speed is roughly inversely proportional to this time scale, i.e. faster flows establish a steady state sooner than slow ones. We also assess the sensitivity of the results to the physical properties of the components of the repository.     

Analysis of thermo-hydro-mechanical process in the engineered barrier system of a high-level waste repository

This paper presents a simulation of the KAERI's engineering scale experiment, KENTEX which is to study the thermal, hydrological and mechanical (THM) processes occurring in the engineered barrier system of a high-level waste repository. The simulation was performed by using the computer code, TOUGH2, which analyzes the multi-dimensional fluid and heat flows of multiphase, multicomponent fluid mixture in unsaturated medium. The detailed geometry of KENTEX was incorporated into the model, and the laboratory experiments were carried out to determine the thermal, hydrological and mechanical properties of bentonite, which were used as input parameters for the simulation. The calculated results on the temperature, water content, and total pressure distribution throughout the bentonite buffer were compared with the experimental ones.     

Calculation of neutron dose rates including backscattering of a POLLUX cask in drifts of a repository mine

The Active Handling Experiment with Neutron Sources (AHE) is intended to study the effect of neutron backscattering from a salt environment during handling of highly active material (spent fuel and high level waste) in a repository located in a salt dome.The AHE project is planned to provide the calculated dose rates of a POLLUX cask with and without salt environment by comparison with the measured dose rates at a smaller experimental shielding cask loaded with Cf-252 neutron sources.     

Effect of helium on the microstructure of spent fuel in a repository: An operational approach

In a repository, the release of radionuclides from spent fuel rods will strongly depend on the pellet microstructure existing when water comes into contact with the spent fuel surface, i.e. after 10,000 years of disposal. During this period, a large quantity of He atoms is produced by α-disintegrations of actinides in the spent fuel. A conservative model is proposed here to evaluate the consequences of He on the spent fuel microstructure. According to the solubility and diffusion properties of He under repository conditions, two scenarios are considered: He atoms can be trapped in fission gas bubbles or form new bubbles. In spite of the conservative assumptions of the model, the calculated values of bubble or pore pressure are much lower than critical values derived from rupture criteria. No evolution of the microstructure of the spent UO₂ fuel is thus expected before the breaching of the canister.     

高压氘氚气氛下不锈钢吸附氚的初步研究

本工作对于氚在不锈钢表面的吸附和解吸行为进行了初步研究.样品在n(D)∶n(T)=1∶1,230℃时,15 MPa下恒温8 h后,接着在27 MPa下恒温6 h的情况下进行了氚的吸附,测量了室温下和加热到1173 K时的解吸氚量和总吸附量.其结果如下:不锈钢的总吸附氚量是857.4 MBq·cm-2,不锈钢的解吸氚量是722.2 MBq·cm-2;在本实验的条件下,在室温和加热条件下,不锈钢所释放的氚中,化学成分主要是HTO和HT两部分,大部分以HT形式存在;不锈钢的自由氚占吸附总氚量的7.34%;不锈钢的热解吸谱存在三个解吸峰,其解吸温度分别为439、530和843K.     

极低放射性废物填埋场中同位素迁移与屏障研究

在查明场址地球化学特征的基础上,研究了238U、90Sr在场址包气带土壤中的吸附与迁移特征。设计了一系列非均匀介质实验模型,研究了在瞬时源和连续源条件下238U、90Sr在非均匀介质中的迁移特征。同位素迁移屏障研究在研究场址围岩对核素的吸附特征的基础上,筛选出合适的矿物添加剂以提高场址岩土对90Sr迁移的屏障能力。结果表明,238U和90Sr在场址土壤中迁移速度分别为0.365和0.385mm/a。非均匀介质对同位素迁移影响首先是减弱了主流体的污染强度,同时延长了污染作用时间;添加剂人造沸石可作为90Sr迁移屏障材料。