The French experience in low level radioactive waste management

The Radioactive Waste Management Agency (ANDRA-FRANCE) is now operating a new facility in the eastern part of the Paris basin which is designed to dispose of one million cubic meters of waste.

The safety of the waste disposal is based on a multibarrier concept including waste packages, concrete disposal modules, site and closure operations.

Under normal conditions, confinement is guaranteed by the waste packages and the disposal modules, as they prevent the waste from being leached by rainfall or underground water over a certain period of time.

The site must bring an additional guarantee concerning the isolation of waste from water. Consequently, the chosen site must be located in an area where no natural disasters (landslides, earthquakes, etc.) can harm the isolating barriers. The geological, hydrogeological and chemical characteristics must allow us to minimize and control the transport of radionuclides within the ground. Finally, the chosen site must be in an area where it is easy to implement a system to monitor the environment.

A set of criteria guides the choice of site. The criteria include such factors as low seismicity, geotechnical stability, a hydrogeology that is simple to model, a location sufficiently above the water table and safe from the threat of flooding, good radionuclide sorption and the absence of any mineral or other natural resources of economic interest.

At the time of the closure of the disposal facility, the entire collection of modules will be covered by an impervious cap composed of clayey layers interbedded by sandy layers and overlain by humus to promote the growth of grass. The facility will then look like a succession of undulating green mounds.

A 300-year monitoring period will follow the closure. During this period, the water collecting networks and cap will be maintained and radioactivity in underground and stream water will be controlled.

We have selected the AUBE site as a case study to illustrate the French waste management experience. We will report on how the site characterization program has been calTied out, including the hydrogeological modelling which is being applied to both the operating and post-closure periods.     

Aqueous geochemistry of radioactive waste disposal

Present plans for disposing of high-level radioactive waste include converting it to a stable solid form, enclosing it in metal canisters, placing the canisters in mined cavities several hundred meters below the surface, and filling and sealing all entries to the cavities. The chief hazard in such disposal is dissolution of radionuclides from the waste in the ground water that will eventually fill the cavity and may carry the dissolved material to surface environments. To prevent or delay release of radionuclides in this manner, the form of the waste must be made as insoluble as possible, the canister metal must be resistant to corrosion, both the cavity filling and surrounding rock must be impermeable and highly sorbent, and the repository must be located at a site where ground water is scanty and slow-moving and its path to the ground surface is long. The effectiveness of these barriers in isolating the waste can be predicted for long times in the future by models based on experimentally determined solubilities and sorption coefficients and on measured rates of ground-water movement. Much uncertainty is involved in such predictions, and additional research is needed to lower the uncertainty. But there seems little question that a well-located and well-engineered repository of this sort will provide adequate protection against harm from radioactive releases far into the future.     

ArcGIS在北山高放废物处置库预选区地学数据管理中的应用

以甘肃北山高放废物处置预选区所积累的数据为基础,研究地学信息数据管理平台的建立,实现对各种空间数据及其他非空间数据统一管理与应用,并提出了基于ArcGIS产品的预选区数据管理与共享的解决方案。     

Confinement matrices for low- and intermediate-level radioactive waste

Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW); 95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000 reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of shallow-seated repositories in various geological media, and the use of inexpensive borosilicate glass.     

Subterranean microorganisms and radioactive waste disposal in Sweden

In 1987, microbiology became a part of the Swedish scientific program for the safe disposal of high level nuclear waste (HLW). The goal of the microbiology program is to understand how subterranean microorganisms will interact with the performance of a future HLW repository. The Swedish research program on subterranean microbiology has mainly been performed at two sites in granitic rock aquifers at depths ranging from 70 m down to 1240 m, the Stripa research mine in the middle of Sweden and the Äspö Hard Rock Laboratory (HRL) situated on the south eastern coast of Sweden. Some work has also been performed in cooperation with other national or international research groups in Sweden, Canada and at the natural analogue sites Oklo in Gabon and Maqarin in Jordan. The following conclusions are drawn. There is a very high probability of the existence of a deep subterranean biosphere in granitic rock. The documented presence of a deep biosphere implies that relevant microbial reactions should be included in the performance assessment for a HLW repository. A HLW repository will be situated in a subterranean biosphere that is independent of solar energy and photosynthetically produced organic carbon. The ultimate limitation for an active microbial life will be the availability of hydrogen as energy source over time, and hydrogen has indeed been found in most deep groundwaters. Sulphide producing microorganisms are active in environments typical for a Swedish HLW repository, and the potential for microbial corrosion of the copper canisters must be considered. The bentonite buffer around the copper canisters will be a hostile environment for most microbes due to the combination of radiation, heat and low water availability. Discrete microbial species can cope with each of these constraints, and it is theoretically possible that sulphide producing microbes may be active inside a buffer, although the experiments conducted thus far have shown the opposite. Microorganisms have the capability to enzymatically recombine radiolysis oxidants formed by radiation of water. It has earlier been concluded that the migration of radionuclides due to sorption on microorganisms can be neglected. The influence of microbially produced complexing agents remains to be studied at realistic conditions in deep groundwater. Microorganisms have been found in natural alkaline groundwaters, but it could not be conclusively demonstrated that they were in situ viable and growing, rather than just transported there from neutral groundwater. A possible hypothesis based on the obtained results from investigations of natural alkaline groundwaters is that fresh concrete may be a bit too extreme for active life even for the most adaptable microbe – but this remains to be demonstrated.     

Glasses for immobilization of low- and intermediate-level radioactive waste

Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m³ liquid LILW with a salt concentration of ～300 kg/m³ leaves behind only 0.2 m³ waste, that is, 4–6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing.     

Application of the coconut fiber in radioactive liquid waste treatment

The treatment of radioactive liquid waste containing organic compounds was always a cause for concern to radioactive waste management facilities because the processes available are expensive and difficult to manage. Biosorption has been studied as a new process in simulated wastes as an alternative to treating them. Among the potential biomass, the coconut fiber has very attractive features that allow the removal of radionuclides using a low-cost biosorbent. The aim of this study was to evaluate the capacity of coconut fiber to remove uranium, americium, and cesium from real radioactive liquid organic waste. Experiments with the biosorption of these radionuclides in coconut fiber were made including (1) preparation, activation, and characterization of biomass and (2) biosorption assays. The biomass was tested in raw and activated form. Biosorption assays were performed, adding the biomass to real waste solutions. The solutions contain natural uranium, americium-241, and cesium-137. The contact times and the concentrations range were varied. The radioisotopes remaining concentration in the solutions was determined by inductively coupled plasma optical emission spectrometry and gamma spectrometry. The results were evaluated by maximum experimental sorption capacity and isotherm and kinetics ternary models. The highest sorption capacity was observed with the activated coconut fiber, with values of 2 mg/g of U (total), 70E?06 mg/g of Am-241 and 40E?09 mg/g of Cs-137. These results suggest that biosorption with activated coconut fiber can be applied in the treatment of radioactive liquid organic wastes containing uranium, americium-241, and cesium-137.     

Evaluation of uranium and arsenic retention by soil from a low level radioactive waste management site using sequential extraction

The European Communities Bureau of Reference (BCR) and Chunguo sequential extraction procedures were employed to evaluate the retention of U and As by a soil contaminated with low level radioactive waste. Modifications were made to both procedures to optimize the measurement of soil and extractant samples using epithermal neutron activation analysis. Based on the BCR procedure, approximately 20% of the U appeared to be bound to the carbonate fraction, 10% to the mineral oxide fraction and 20% to the organic fraction. In the case of As, the majority was strongly bound in the residue fraction. The results obtained with the Chunguo procedure supported these conclusions to some extent, in that the majority of the U and As was found to be strongly bound to the soil in a manner consistent with its presence in the residue fraction.     

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。     

Migration of transuranic elements in groundwater from the near-surface radioactive waste site

Field experiments and laboratory studies were performed to investigate migration processes of plutonium isotopes from a near-surface radioactive waste trench to the underlying sandy aquifer at the Red Forest waste dump in the Chernobyl zone. The objectives of these experiments were to characterize the spatial distribution and possible migration mechanisms of plutonium in the aquifer. During 2002–2007 experimental investigations were carried out and spatial distributions of plutonium isotopes (^239,240Pu, ²³⁸Pu), ⁹⁰Sr and major ions in the aquifer in the direction of the groundwater flow were obtained. Specific activities of radionuclides in groundwater depended on the location of the piezometer and varied in the range of 1–360 mBq kg⁻¹ for ^239,240Pu, 0.5–180 mBq kg⁻¹ for ²³⁸Pu and n–n·10⁴ Bq kg⁻¹ for ⁹⁰Sr. It was found that the spatial features of the distributions of plutonium and strontium specific activities in the upper eolian aquifer were similar, i.e. there was a correlation between the positions of the activity maxima of the radionuclides. The Pu isotopes plume in the aquifer spreads about 15 m downstream of the radionuclides source. Characterization of the initial radionuclide composition of the waste showed that all plutonium in the aquifer originated from the trench. The ratio of plutonium isotopes (^239,240Pu/²³⁸Pu) at the sampling time was the same in waste material and in groundwater samples. In situ ultrafiltration of several groundwater samples was carried out. The size fractionation data obtained suggest that a significant part of plutonium (50–98%) in the groundwater sampled close to the source from the upper part of the aquifer is associated with a very low molecular weight fraction (<1 kDa).     

高放废物地质处置系统核素迁移模型研究

高放废物地质处置系统核素迁移研究是高放废物安全处理和处置的重要内容.从整个系统的角度出发,把高放废物地质处置系统核素迁移过程分为三种模式:工程屏障中的释放-扩散模型;地质屏障中的对流-弥散模型和在生物圈中的分区传递模型.在分区传递模型中利用转移系数描述核素在各分区之间的迁移.通过对每种模型核素迁移机理的分析得出概化模型,并给出了相应的数学描述.     

System reliability analysis for near-surface radioactive waste disposal facilities

The low- and intermediate-level radioactive wastes are generally disposed in near-surface disposal facility (NSDF). The NSDF is composed of engineered barriers. The probabilistic safety assessment model has been developed to analyse the performance of NSDF. The endpoints of the assessment are the concentration of the radionuclide in the groundwater and the corresponding dose rate of the radionuclide. The barrier system can have multiple failure modes but practically the possible failure modes are failure of top cover, failure of waste container, degradation of waste form and failure of bottom cover, which are usually considered as independent failure events. Through a sensitivity analysis, the most critical parameters affecting the design reliability for failure criteria are identified as the groundwater velocity and distribution coefficient. The study shows that for the NSDF considered, there is a high degree of dependence between the failure modes, and demonstrates the probability of the simultaneous occurrence of failures. Thus, the need to consider the system reliability in the NSDF is highlighted. The study also advocates the use of optimisation techniques to evaluate the probability of failure, which provides a better estimate of the probability of failure, as validated from the results obtained from Monte Carlo simulations.     

Long-term prediction for seismic hazard for radioactive waste disposal

We consider possible approaches to the long-term prediction for seismic hazard in relation to the practical need for the safety of geological disposal of long-lived radioactive waste. The required period of prediction significantly exceeds the one reflected in the set of maps of General Seismic Zoning of the territory of the Russian Federation (GSZ-97). The first geological repository in Russia is planned to be set up in the Nizhnii Kan granite massif in the Krasnoyarsk Krai. This region is an intraplate territory with a relatively high seismic activity. We summarize the analysis of the known empirical generalizations and theoretical principles underlying the seismic hazard prediction. Real seismic events constantly violate forward-looking statements even for relatively short periods of time. These and other arguments suggest that the hypothesis of stationarity of the seismic regime, which is the basis of long-term prediction today, has limited and uncertain applicability in time. Intraplate earthquake prediction is especially uncertain because of the uncertainty in the factor responsible for generating tectonic stresses in these regions. The short horizon of the prediction, based on statistical methods, can be attributed to the nonlinearity of seismic geodynamic processes. Fundamental laws of tectonic processes should be used as the scientific basis for long-term predictions for seismic hazard at the sites chosen for geological disposal of long-lived radioactive waste. These processes can be reflected in models for the migration of the seismically active boundaries of lithospheric plates and the occurrence of seismic activity in intraplate regions.     

A model approach to radioactive waste disposal at Sellafield

One of the great environmental problems of our age is the safe disposal of radioactive waste for geological time periods. Britain is currently investigating a potential site for underground burial of waste, near the Sellafield nuclear plant. Future leakage of radionuclides depends greatly on subsurface water flows; these must be understood from the past, to predict hydrogeology 10⁴–10⁵ years into the future. We have taken information from the present-day, published by the government company Nirex, and used a finite-element steady-state fluid flow computer code to examine water flows in the subsurface. We find that flow directions at the planned Repository are persistently upwards, and that geologically significant flow rates could occur. How rates are particularly sensitive to uncertainties of rock permeability (conductivity) measurements made from site investigation boreholes. The hydrogeology at this site needs longer term investigation before a confident and credible prediction can be made.     

极低放射性废物地质处置中的几个问题

极低放废物是国际上近年来新提出的一种放射性废物,目前尚未形成公认的定义和处置方案。针对地质填埋中几个关键科学问题进行了初步探讨,涉及到的内容主要包括区域地壳的稳定性、基础地质条件、水文地质条件、地球化学条件、地质灾害、安全评价等方面,最后提出了今后几个重要的研究方向。     

极低放射性废物地质处置中的几个问题

极低放废物是国际上近年来新提出的一种放射性废物,目前尚未形成公认的定义和处置方案。针对地质填埋中几个关键科学问题进行了初步探讨,涉及到的内容主要包括区域地壳的稳定性、基础地质条件、水文地质条件、地球化学条件、地质灾害、安全评价等方面,最后提出了今后几个重要的研究方向。     

高放废物地质处置元数据设计与编辑模块开发

在对元数据的含义、作用、组成及对高放领域研究的数据内容和特点分析的基础上,对元数据编辑模块的设计进行了详细描述,并在VS2005平台下应用C#语言进行了模块的实际开发,最终实现了便于进行数据共享的XML格式存储功能。元数据编辑模块的应用将促进我国高放废物地质处置信息化工作的进一步深入。     

包气带黄土环境研究与放射性废物处置

黄土中的粘粒含量较高、比表面积较大、吸附性较强、阳离子交换容量也较高,这些因素对迟滞放射性核素的迁移有利。而且黄土的非饱和渗透系数K(θ)远远低于它的饱水渗透系数K_(?)。使得放射性核素在黄土中的迁移很慢。因此,包气带黄土作为中、低放射性废物处置库的环境屏障是可行的。