Comprehensive evaluation of the state of spent VVR fuel assemblies from first-generation nuclear submarines

New, unconventional methods developed at the Russian Science Center Kurchatov Institute for evaluating the state of spent VVR fuel assemblies from submarines are described. They were used in combination with other methods during inspections of spent fuel assemblies held in TK-6 and -11 containers at the point of storage of spent nuclear fuel and radioactive wastes in the town of Gremikha in Murmansk Oblast.     

Fabrication of Mixed Urianium–Plutonium Fuel on the Paket Setup and Waste Handling Methods

The results of fabrication of fuel elements with mixed uranium–plutonium oxide fuel are presented. The experimental fuel assemblies assembled from the fuel elements were tested in BN-350 and -600 reactors. Postreactor investigations of the fuel elements showed that there was no substantial difference in the behavior of the fuel cores consisting of the mixed fuel as compared with UO₂ fuel. Solid and liquid radioactive wastes are produced during the fuel fabrication process. A classification of the wastes and methods for handling them is given. It is shown that the off-grade sintered pellets should be pulverized and returned to the beginning of the mixed-fuel fabrication process.     

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.     

Nuclear safety in handling radioactive wastes containing fissile materials

Nuclear safety in disposal and reprocessing of solid and liquid radioactive wastes produced during the production of nuclear reactor fuel elements was investigated. The results of this work made it possible to determine the parameters which must be limited in handling wastes in order to avoid the occurrence of a self-maintaining fission chain reaction. The numerical values of these parameters are obtained for the new and conventional technology for handling wastes. 3 tables, 5 references. State Science Center of the Russian Federation-A. I. Leipunskii Physics and Power-Engineering, Institute. Translated from Atomnaya énergiya, Vol. 88, No. 5, pp. 362–370, May, 2000.     

Handling spent nuclear fuel at a temporary storage site in Guba Andreev

Spent nuclear fuel has been stored in dry-storage units at a shore base of the naval fleet for 35–45 year. The total activity of the spent nuclear fuel is 170 PBq. This article presents data which characterize the state of the fuel (from normal to defective), the radiation conditions, and information on the individual and collective irradiation dose to workers. The results of an inventory check of the cells and jackets which contain fuel assemblies are presented. The corrosion processes are described and ideas for handling the spent fuel at the RT-1 plant of the Mayak Industrial Association, including handling fuel assemblies and jackets in cases, are described. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 56–61, July, 2006.     

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.     

High-density support matrices: Key to the deep borehole disposal of spent nuclear fuel

Deep (4-5 km) boreholes are emerging as a safe, secure, environmentally sound and potentially cost-effective option for disposal of high-level radioactive wastes, including plutonium. One reason this option has not been widely accepted for spent fuel is because stacking the containers in a borehole could create load stresses threatening their integrity with potential for releasing highly mobile radionuclides like ¹²⁹I before the borehole is filled and sealed. This problem can be overcome by using novel high-density support matrices deployed as fine metal shot along with the containers. Temperature distributions in and around the disposal are modelled to show how decay heat from the fuel can melt the shot within weeks of disposal to give a dense liquid in which the containers are almost weightless. Finally, within a few decades, this liquid will cool and solidify, entombing the waste containers in a base metal sarcophagus sealed into the host rock.     

对秦山核电基地低、中水平放射性固体废物减容处理的考虑

废物最少化是放射性废物管理的原则之一.针对秦山核电基地废物处理现状,分析了国家法规、标准对低、中水平放射性固体废物的管理要求和核电厂面临的相关问题,提出了对秦山核电基地低、中水平放射性固体废物进行焚烧和超级压缩处理的考虑.     

Safety of Pyroelectrochemical Reprocessing of Spent BOR-60 Fuel

The results of an investigation of the safety of reprocessing 15.5 kg of spent BOR-60 fuel are presented. The absolute amounts of radioactive aerosols entering the ventilation system of the pyroelectrochemical setup during the reprocessing of spent fuel are determined. The transfer of radionuclides from fuel into aerosols is estimated. This makes it possible to obtain an expression for the expected inflow of radioactive substances from the setup, depending on the amount of materials handled in the technological scheme. The data presented on the possibility of obtaining large amounts of uranium and plutonium in the finished product when their content in the solid wastes is negligible are presented. Estimates are made of the irradiation of workers at all stages of the process.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 280–288, April, 2005.     

Abstract

Abstract

This article is aimed at summarising the activities related to the transport of radioactive materials carried out in Argentina and, especially, with regard to the transport of radioactive wastes. In particular, the legislation applicable within the national territory is described. Additionally, figures are provided on the features and amounts of transported radioactive materials, including radioactive wastes, concerning both the nuclear fuel cycle and activities related to their industrial and medical applications.     

Possible Application of γ-Ray Spectrometers Based on CdZnTe Detectors

New semiconductor -ray detectors based on CdZnTe have been developed in the last few years, and they are now being widely used. Their advantages are small size and possibility of operating without cooling. In the present work, these detectors are used to measure the radiation from spent fuel assemblies in holding ponds and dry storage sites, unirradiated nuclear materials, and radioactive wastes. The results are analyzed and compared with other types of detectors. The possible applications of CdZnTe-type detectors are determined.     

A concept of nitride fuel actinide recycle system based on pyrochemical reprocessing

A fuel cycle system coupled with nitride fuel fast reactors and a pyrochemical reprocessing has been investigated in order to establish an actinide transmutation recycle system with long-lived radioactive nuclides. Core performance of the nitride fuel fast reactor can provide design flexibility of excellent safety characteristics and a new concept of core composed with Na- and He- bonded fuel assemblies is proposed. The effect of ¹⁵N enrichment on nuclear characteristics and the evaluation of toxicity of ¹⁴C generated from ¹⁴N are appeared, and futhermore, excellent performance for the minor actinide (MA) transmutation is shown.

The study of the pyrochemical process shows that the actinides are reasonably separated from fission products in the nitride spent fuels, and that the high level wastes are of nearly actinide-free form.     

放射性废物超级压缩装置研制及性能试验

为实现放射性废物压缩设备国产化,设计、研制了超级压缩装置,并针对核电站多种典型废物进行了试验研究。结果表明:在累计运行200小时后,装置完好无渗漏,生产能力达到7桶/h。研制的超压装置可用于核电站产生的可压缩放射性废物的处理。     

Actual and possible nuclear power long-time development lines

The current state of nuclear power engineering is described. It is shown that this industry is still at the initial stage in general industrial use in spite of experimental evidence accumulated for more than half a century. It has not even attained the stage of establishment, which would enable one to utilize all the power facilities concealed in it. The long-term future of nuclear power involves a closed fuel cycle including the recycling of fuel from extended breeding of fissile materials, which supports the safe treatment of radioactive wastes. Nuclear power has no alternatives over millennia, and any groups opposed to its extension and companies within the country that strive to delay its development by not providing facilities for the latter in essence have started on a road involving marked difficulty for our civilization. Bochvar All-Russia New Materials Research Institute, Russian Federation State Scientific Center. Translated from Atomnaya énergiya, Vol. 86, No. 2, pp. 83–94, February, 1999.     

Development of a unified technological process for cementing solid radwastes

The results of research on cementing finely dispersed, mixed, closely packed, and coarsely fragmented solid radioactive wastes by mixing, flow-through, and impregnation to create a unified complex of technological processes are presented. A new method of impregnation which consists of feeding under pressure through a probe into the bottom of a container with solid wastes a cement solution with a high penetrating power is shown. The drawbacks of flow-through are eliminated, the integrity of the pouring is guaranteed, and the quality is monitored by determining the degree of impregnation and density of the cement solution which has passed through the layer of solid wastes. This method of cementing can be used for all solid wastes considered and the equipment and cement compositions are basic for the works cementing solid radioactive wastes.     

Chapter 3: Type and Quantity of Radioactive Waste Destined for Disposal

Abstract

According to the current planning status, radioactive waste with negligible heat generation is destined for storage in the Konrad waste repository in accordance with the Provisional Waste Acceptance Requirements⁽⁴⁾. Waste of this kind occurs in the nuclear fuel cycle, research, medicine and technology. In its original state, the primary wastes of this type have various forms, such as: liquids, concentrates, sludges; ion exchange resins; compressible and/or combustible solids; incompressible solids, e.g. structural material components; filters, filter candles; other types of waste. The radioactive waste is appropriately conditioned and packaged before transport to the waste disposal site. To meet the basic requirements specified in the Waste Acceptance Requirements, primary waste must be solidified, for which the most common solidifying agents are cement and concrete although bitumen is also used. Waste products in decomposing, fermenting or liquid form or which contain a significant fraction in such states are not accepted for disposal.     

放射性废物管理中值得关注的一些新概念

本文介绍了国际原子能机构及国际放射防护委员会近年来在放射性废物管理领域提出的一些新概念。诸如对放射性废物实行全过程管理；对长寿命固体放射性废物处置的安全评价采用有约束的最优化办法，并且指出最优化是一个判断过程等。本文用意在于引起同行对这些概念的注意和重视，以便对放射性废物的安全管理有所帮助。     

Design and thermal modeling of radially configured nuclear waste packages

A series of radial design configurations for packaging nuclear wastes are described. These radial arrangements for used nuclear fuel assemblies in containers are effective techniques for packaging significantly more radioactive waste in the available internal container volume. The radial package designs can be applied to packaging the nuclear waste for permanent storage at the Yucca Mountain (YM) repository. The radially configured containers will have high degree of structural strength and will be efficient in transferring heat from the waste form to the package surface due to the minimization of internal gaps. Radial configurations are reported for packaging the Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) used fuel assemblies. These configurations can be varied for co-packaging the colder, i.e. vitrified high level waste (HLW), canisters. Details of the geometry and the materials selected are discussed. Thermal analysis of the radial designs was conducted which confirm the feasibility of the designs demonstrating that no over-heating occurs in the contained nuclear waste in spite of the significantly extra amount of waste. The larger amount of packaged waste per container coupled with efficient heat transfer characteristics of these designs favor hotter and drier conditions for container surfaces in the YM emplacement drifts.     

Analysis of Radioactive Wastes Formed as a Result of the Production of 99m Tc on the Basis of Irradiated Molybdenum Trioxide Using a Centralized Generator as a Base

The radioactive wastes formed as a result of the production of ^99m Tc by the zirconium–molybdenum gel technology are studied. It is shown that high-activity isotopes are not formed when a MoO₃ target is irradiated; there are no gaseous and solid wastes. The activity of the liquid wastes is due to spent ⁹⁹Mo and short-lived ^99m Tc; after the ⁹⁹Mo decays, they can be removed as ordinary water solutions. Therefore, this technology for obtaining ^99m Tc is waste-free from the standpoint of radiation. ⁹⁹Tc accumulation in liquid radioactive wastes is estimated. It is shown that this technology for producing ^99m Tc on a centralized zirconium–molybdenum gel-generator is ecologically clean.