Physicochemical and biological monitoring of deep repositories for liquid radioactive wastes

The objective of this work was to study the physicochemical conditions and the microbiological composition of underground water in the second level of the Severnyi site at the Mining and Chemical Works and to clarify the possibility of biogenic gas production by formation microflora from macrocomponents of the wastes (nitrates and sulfate ions). Chemical analysis of samples of formation liquid from wells located in the dispersion zone of the wastes showed an increase in the content of dissolved carbonic acid, sodium nitrate, and nitrates of radionuclides and an increase of their concentrations in the sampling part. In all other samples, the contents of the main cations and anions were close to the background values. Microbiological investigations showed an increase in the number of microorganisms capable of forming gases from possible macrocomponents of the wastes and in the rate of sulfate reduction and methane generation processes within the propagation boundary of the wastes. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 106–112, August, 2007.     

An extreame disposition method for low-level radioactive wastes using supercritical water (3)

Disposition of low-level radioactive wastes has been performed in supercritical water with RuO₂ as a catalyst without the addition of any oxidizing materials. These wastes arose from nuclear power plants etc., constituting of nonflammable organic materials, such as fire resistant sheeting and rubber gloves. We investigated the distribution behavior of iron and cobalt attached to nonflammable organic materials, in solid, liquid and gas phases during the decomposition of this method. The distributions of these elements under various conditions (initial amounts) were determined by using their radioisotopes as simulated low-level radioactive wastes in order to ease the detection of trace amounts of elements even in solid and gas phases. Iron and cobalt were found only in the solid phase when non-radioactive iron hydroxide was added as a precipitation reagent before the supercritical-water reaction.     

Handling of low-level wastes at the Siberian Chemical Works

Safety codes require that open surface disposal sites for liquid low-level wastes at radiochemical works be closed down. At the Siberian Chemical Works, the first operation was to stop the flow of wastes into the pulp repository. The presence of burial sites predetermined the choice of precisely this method as an alternative for dumping wastes into open disposal sites. The results of a study of the physicochemical characteristics of the wastes flowing into the pulp repository PKh-1,2 from four plants at the Works made it possible to develop a unified scheme for preparing them for disposal — wastes are put into a deep repository and maximum use is made of the individual properties of reprocessed wastes, which decreases the consumption of additional reagents substantially, i.e., it decreases the cost of the preparation process. The scheme developed has successfully passed commercial prototype tests and is now in the design stage. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 112–115, August, 2007.     

Safety assessment of liquid radioactive wastes buried for a long time in deep repositories

The concepts of specific and absolute toxicity permit distinguishing in radioactive wastes belonging to different categories the particular components that present the greatest ecological danger at different stages of their disposal in repositories. It is shown that transuranium elements determine the toxicity parameter of high-and medium-level wastes only 350–400 years after initial disposal. At the present stage of operation of deep repositories, the behavior of the long-lived fission products of ⁹⁰Sr and ¹³⁷Cs is most important. For low-level wastes, the chemical and not the radioactive component is most dangerous. __________ Translated from Atomnaya énergiya,Vol. 100, No. 2, pp. 86–92, February, 2006.     

Substantiation of the Radiochemical Characteristics of Liquid Radioactive Wastes That Determine Safe Operation of Deep Waste Disposal Sites

The qualitative and quantitative radiochemical composition of liquid wastes is substantiated. This makes possible further safe operation of deep waste disposal sites at the Mining-Chemical Plant and the Siberian Chemical Plant. The toxicity and temperature in a formation with the wastes removed are used as the assessment criteria, satisfaction of which guarantees that the wastes will remain localized within the assigned boundaries of the waste-disposal formation site. It is concluded that the standard limit should be imposed on the specific activity of the long-lived group of radionuclides – ⁹⁰Sr and ¹³⁷Cs – rather than on the total specific activity of the wastes placed in deep disposal sites. For maximum specific activity of ⁹⁰Sr and ¹³⁷Cs of 37 GBq/dm³ and total specific activity not exceeding 185 GBq/dm³ for buried wastes with radionuclide composition characteristic of modern radiochemical production operations, it is impossible for potentially dangerous radiation and thermochemical processes to occur in the waste-disposal formation site. The recommended limit permits reducing substantially the volume of buried wastes and therefore the region over which the wastes propagate in the deep disposal site.     

Development of a γ-ray Measurement System of Radioactive Wastes

According to the criteria of radioactive wastes, intermediate and low-level radioactive solid waste produced in nuclear fuel recycle must be characterized for near surface disposed. For this purpose, a special y-ray measuring system used for classifying low or medium-density drummed radioactive wastes was developed. The system consists of three sets of high purity germanium detectors with multi-channel analyzers. When the system works, waste drums rotate on axial and divided into three segments on radial.     

The verification tests of the melting conditions for homogenization of metallic LLW at the JAEA

To clarify the operating conditions for the homogenization of metallic low-level radioactive wastes (LLW), melting tests were conducted in the metal melting system of the Advanced Volume Reduction Facilities at the Japan Atomic Energy Agency (JAEA). Simulated metallic LLW with a non-radioactive tracer were melted in the induction furnace of the metal melting system. Regardless of the chemical composition or amounts of the simulated wastes, the tracer distribution in the solidified products was nearly homogeneous when the simulated wastes were completely melted in the range from 1550 °C to 1638 °C.     

Immobilization of Radioactive Wastes by Embedding in Phosphate Ceramic

This work is devoted to the development of new composite materials for immobilizing medium- and low-level radioactive wastes. The most promising existing methods of solidification are low-temperature processes because they do not require complicated equipment and are cost-effective. As a result of the higher cost-effectiveness, less inert binder is needed in the overall volume of the solidified compound. Thus, using a complex phosphate, making it possible to increase the concentration of toxic products while maintaining reliable immobilization of the latter, is certainly a topical problem.Two advantages of phosphate ceramic are that high concentrations of anions which adversely affect the longevity of cement monoliths can be included and the spectrum of types of wastes that can be handled, including sol residues of combustion and mercury-containing wastes, is wide.     

美国低放废物处置概况及其经验和教训

本文概述了美国低放废物的产生和处置情况,介绍了1986年“低放废物政策法令的1985年修正法案”颁布后对美国低放废物处置的影响及其变化,指出我国低中放废物处置应从中借鉴的经验和教训。     

Handling spent nuclear fuel and radioactive wastes on the shore servicing base in Gremikha

The situation which has developed at the shore base in Gremikha involving fuel assemblies in the removed cores of water-moderated water-cooled reactors (first-generation submarines), which are located on an open site in containers and receiving chambers, and involving solid and liquid radioactive wastes present at the base is examined. Data are presented on the number of fuel assemblies and their technical state and on the state and amount of solid and liquid radioactive wastes. Suggestions on what should be done with the fuel assemblies and wastes are discussed. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 61–65, July, 2006.     

Volume reduction, treatment and recycling of radioactive waste

Both solid and liquid wastes are generated in a nuclear power plant. Liquid wastes are better treated on site owing to the large quantities generated. For solid wastes off-site treatment in central facilities makes better sense. The largest volume reduction factors are obtained for solid, combustible wastes through incineration and subsequent supercompaction. Contaminated metal components, such as used heat exchangers, vessels, tanks, valves, etc. are recycled; that is, they are remelted and then manufactured into new components for use in nuclear facilities. Evaporation of liquid radwaste provides initial volume reduction, which is followed by drying to reduce the waste to a dry, solid product.     

我国低放废物热解焚烧技术的应用及改进

根据国际上焚烧技术发展趋势,针对我国放射性废物管理的特点,中国辐射防护研究院自主开发出多用途放射性废物热解焚烧技术,在国内应用近20年,建成3座焚烧设施。这些焚烧设施运行、处理了大量低放废物,具有净化效率高、减容效果好、核素控制好等特点,验证了热解焚烧系统的稳定性、可靠性与先进性。中国辐射防护研究院针对早期焚烧设施运行中出现的问题,不断进行完善与改进,降低了设备腐蚀问题;提高了系统运行的安全性、稳定性;减少了二次废物的产生量;并进一步提高了焚烧系统的自动化程度和对高塑料橡胶含量废物的兼容处理能力。经过这些技术改进,国内多用途放射性废物热解焚烧技术逐渐完善。     

油气工业中放射性废物的处理和处置

油气的开采会产生大量的放射性废物,对放射性废物的处理与处置事关公众和环境的辐射安全问题。本文从油气工业放射性废物的源项和存在形式等特征出发,对放射性废物的临时贮存、污染设备的去污及放射性废物的处理、处置过程中涉及的技术方法和相关管理要求进行了介绍,为实践中选择合理可行的处理、处置方案以及建立油气工业放射性废物监管体系提供借鉴和参考。     

Packaging Design and Qualification: The Experience of the CDTN/CNEN

Abstract

Since 1982 the CDTN, the Nuclear Technology Development Centre, has been designing, testing and qualifying packaging for radioactive materials. These packagings are used for the transport of radioisotopes and disposal of spent sealed sources, wastes generated in the nuclear fuel cycle and the wastes produced in the radiological accident that occurred in the city of Goiânia. For radioactive tracers and medical/industrial radioisotopes, the packagings used are cardboard and wood boxes, while the spent sealed sources are preferably conditioned in metal drums containing lead shielding and a gas absorber material. To condition and transport the wastes from the various nuclear cycle activities, metal drums and boxes are used in Brazil. For the higher active wastes from the nuclear power plant Angra I, a metallic drum in a concrete overpack is used. The wastes generated in the accident were first conditioned in the readily available packaging. Later on, more appropriate packaging was designed by the CDTN staff. CDTN has carried out a programme since 1983 to evaluate the durability of commercial drums used for waste conditioning.     

Handling of the Bottom Residues of a Nuclear Power Plant

The following technological operations are suggested for reprocessing bottom residues from nuclear power plants: separation of radionuclides using oxidation, filtration, and selective absorption, solidification and long-term storage of secondary radioactive wastes (cement compound from filtration stage and spent sorbent in filters); concentration and obtaining dry salts from bottom residues from which radionuclides have been removed. Laboratory and stand tests have been performed, showing that radionuclides can be removed from the bottom residues of nuclear power plants to a level below the ASA_sat according to NRB-96. This treatment decreases the volume of radioactive wastes by approximately a factor of 100. The dried purified bottom residues, which are commercial danger class III wastes, are shipped to storage sites used for industrial wastes. Calculations of the material flows are performed and the site arrangement of the wastes is given for the bottom residues from the Kursk nuclear power plant. 1 figure, 2 tables, 12 references.     

LiNO3 Effect on Corrosion Prevention of Aluminum with Complex Shapes

The LiNO₃ effect on aluminum corrosion prevention after land disposal of cement-solidified dry active wastes was examined quantitatively, in the event that the LiH (AlO₂)₂·5H₂O (Li-Al) preservation film was not formed on aluminum surfaces during the solidification process. It is especially probable for these bare surfaces to be left when the wastes include components of complex shapes. LiNO₃ dissolves from the waste forms into underground water to form the Li-Al preservation film. So, we thought that the LiNO₃ addition would prevent the corrosion. We measured the volume of hydrogen gas generation in mortar-soaked water during the Li-Al preservation film formation, as functions of LiNO₃ addition amount, the weight ratio of water to mortar when the mortar-soaked water was produced, and the aluminum surface area, to quantify the effect.

We found that aluminum corrosion was inversely proportional to the LiNO₃ addition. For the corrosion to be less than 10^?5m in 10³h, the initially added amount of LiNO₃ must be 1.5wt% of the sum of cement and sand. Regardless of the weight ratio of water to mortar when the mortar-soaked water was produced, hydrogen gas generation with LiNO₃ was 10% as much as that without it, in 5 x 10³h. Because of the Li-Al preservation film formation reaction, hydrogen gas generation was proportional to the cubic root of the aluminum surface area.     

镎、钚在处置区的存在形式及影响因素

镎、钚是某些放射性废物中的关键核素，无论是中低放废物泥浆的水力压裂法处置还是高放废物的深地质处置，都必须首先了解处置区和处置条件下水介质中核素镎、钚的存在形式。本文选用地球化学模式程序EQ3/6计算了低中放和高放废物中关键核素镎、钚在处置区水介质中的形式，并讨论了水介质和pH条件对核素存在形式的影响。结论是核素镎、钚的存在形式主要受水介质条件的影响，但pH条件也是影响核素镎、钚存在形式的重要因素。     

Management of radioactive wastes from the operation of nuclear power plants

A prerequisite for the acceptance of the nuclear energy system is the effective management of the rad-wastes. Among the wastes to be considered, there are the wastes from the operation and decommissioning of nuclear power plants, as well as those from the nuclear fuel cycle. For the management of operating wastes, processes and facilities optimized in the course of several decades, are available, with which the raw solid and liquid wastes can be reduced in volume and turned into products which are physically and chemically stable and thus suitable for final disposal. The management of spent fuel can be done either by direct final disposal or reprocessing. The required interim storage facilities are ready for operation. The methods and a facility for packaging spent fuel for direct final disposal are in an advanced stage of development and construction. If fuel assemblies are to be reprocessed abroad, the wastes generated from the process must be taken back. Decommissioning wastes have technical properties which correspond essentially to the various groups of operating wastes and can thus be processed with similar methods; however since large quantities of them are generated in relatively short times, they present particular logistic problems. All waste types end up in final disposal sites to be built under the responsibility of the federal government. A final disposal site for low level wastes is in operation. In addition, two final disposal projects for accommodating higher level wastes including spent fuel for direct disposal and vitrified wastes from reprocessing, are being pursued.     

Effect of the salt melt composition,temperature, and interaction time on the contact exchange reaction in the MCl-PbCl<Subscript>2</Subscript>-MeN systems

The interaction of titanium, zirconium, and aluminum nitrides with a mixture of fused alkali-metal chlorides with lead chloride at 870–1070 K is studied. It is shown that as a result of the contact exchange interaction, which occurs in no more than 5 h, a large fraction of the powdered nitrides transforms into a soluble state, and lead precipitates in a metallic form. The same behavior is also characteristic for compact hot-pressed samples of nitrides. The method proposed for transferring insoluble nitrides into a state which is soluble in salt systems can be used to recover mixed uranium-plutonium mononitride fuel and to fractionate fission products into high-, medium-, and low-level radioactive wastes. __________ Translated from Atomnaya énergiya, Vol. 104, No. 6, pp. 343–348, June, 2008.     

Development of calculation methodology for estimation of radionuclide composition in wastes generated at Fukushima Daiichi nuclear power station

ABSTRACT

A calculation methodology for estimating the radionuclide composition in the wastes generated at the Fukushima Daiichi nuclear power station has been developed by constructing a skeleton overview of the distribution of radionuclides considering the material balance in the system and calculation flowcharts of the transportation of radionuclides into the wastes. The wastes have a distinctive feature that their level of contamination includes considerable uncertainties because the process behind the contamination with the radionuclides released from the damaged fuel during and after the accident is not yet fully understood. Here, the developed method can explicitly specify the intrinsic uncertainties as a band to be included in the estimated radionuclide composition in the wastes and can quantitatively describe the uncertainties by calibration using analytically measured data on actual waste samples collected at the site. Further studies to improve the quality of the calculation method, the introduction of a stochastic approach to describe uncertainties, and acquiring a quantitative understanding of the spatial distribution of radionuclides inside the reactor building are suggested as important steps toward reasonable and sustainable waste management as an integral part of the decommissioning of the Fukushima Daiichi nuclear power station.