Physical plasma method of separation by mass of the constituent elements together with chemical methods could find application in the fuel cycle of fast recators. These methods could also be useful for separating decay products during reprocessing of spent fuel of existing reactors and in the solution of questions concerning safety and nonproliferation. A possible plasma separation scheme is presented: solid-state material is transferred in a magnetic field into a low-temperature plasma flow and ions in a prescribed mass range are selectively accelerated and then spatially separated in a toroidal magnetic field.The problems of nuclear materials handling arise beause it is necessary to separate the initial material with a complicated composition into groups of elements in order to obtain products which are suitable for subsequent use or disposal. Ordinarily, this is accomplished by dissolving solid nuclear material and then separating the components. Radiochemical methods have their undoubted advantages and disadvantages, specifically, due to the formation of a large quantity of radioactive wastes when the spent fuel is reprocessed.The types of nuclear materials are diverse. They can be spent nuclear fuel or radioactive wastes. They can differ according to the degree of burnup, storage time, and technical state. Methods for handling these materials must take account of all their characteristics, be ecologically clean, and meet the requirements of the nonproliferation principle.The idea of using the plasma state of a material to develop a universal method for reprocessing wastes was advanced in the middle of the twentieth century. It was suggested that a thermal nuclear reactor be used for this [1]. Now, it is suggested that a low-temperature plasma be used for these purposes. Commercial reprocessing of toxic and other wastes by plasma methods is being considered [2]. Full-scale investigations of isotope separation by plasma methods have been performed [3]. The technological processes using low-temperature plasma have been developed and are used in the nuclear fuel cycle [4]. A plasma technology is proposed and used for reprocessing radioactive wastes [4,5].A long-term program for developing nuclear power requires the development of fast reactors with a solution of the problems of fuel balance, economics, and safety. Elimination of the isotopic separation of uranium and separation of plutonium from the uranium is a necessary condition for nuclear power to be proliferation resistant. In contrast to chemical methods, physical plasma separation methods without high resolution with respect to atomic mass can meet this condition.

The results of calculations of the neutron-physical characteristics of three variants of the fuel load in a VVÉR-1000 core are presented: a load consisting completely of enriched natural uranium (standard fuel) or reprocessed uranium-plutonium fuel from the first and second recycles. The calculations were performed for a stationary load of the core with a four-year fuel run. The difference between the neutronphysical characteristics of a core with a full load of uranium and reprocessed uranium-plutonium fuel is discussed. An analysis of the neutron-physical characteristics did not show any fundamental limitations for a possible 100% load of reprocessed uranium-plutonium fuel in a VVÉR-1000 core.Recirculating reprocessed uranium and plutonium in thermal reactors could increase the utilization efficiency of nuclear fuel and expand the resource base of nuclear power [1]. At the present time, experience has been gained in using reprocessed uranium and plutonium separately in thermal reactors. In our country, the uranium separated from spent VVÉR-440 fuel is mixed with the uranium extracted from spent BN-600 fuel and then usud for fabricating RBMK-1000 fuel [2]. The same scheme is used to fabricate fuel for experimental-commercial operation of fuel elements with reprocessed uranium in VVÉR-440 and -1000 reactors [3]. Plutonium separated from spent PWR fuel is used abroad, mainly in France, as a component of mixed fuel (a mixture of reprocessed plutonium and depleted or natural uranium) which is loaded into 30% of the PWR core [4,5].It has been proposed that fuel made from uranium and plutonium which are separated from the spent fuel of thermal reactors be used in these reactors as fuel after other actinides and fission products are removed and enriched natural uranium is added, taking account of the compensation of the even isotopes of uranium and plutonium [6,7]. It was supposed that because the content of plutonium in the reprocessed uranium-plutonium fuel is relatively low such fuel can make up 100% of the load of a VVÉR-1000 core.To check this assumption, the neutron-physical characteristics of three variants of a stationary load of a four-year run of fuel in a VVÉR-1000 reactor were analyzed. In the first variant, the load consisted entirely of fuel enriched with natural ura-UDC 621.039.516

A possible version of the VVÉR-1000 fuel cycle without separation of uranium and plutonium during reprocessing of spent fuel is examined. In this fuel cycle, the uranium-plutonium regenerate obtained, from which other actinides and fission products have been removed, is used after enriched natural uranium is added for preparing VVÉR fuel. The results of a calculation of the content of uranium and plutonium isotopes in the spent uranium-plutonium fuel after one and two recycles in VVÉR-1000 are presented. The main advantages of the fuel cycle are discussed: lower risk of plutonium proliferation, savings of natural uranium, and less spent fuel as compared with an open uranium fuel cycle.In currently existing technology for reprocessing spent fuel, uranium and plutonium are divided into two fractions, which presupposes that the fractions have different uses. Thus, at the RT-1 plant the uranium regenerate separated during reprocessing of spent VVÉR-440 fuel is mixed with uranium obtained from spent BN-600 fuel, and used for fabricating RBMK-1000 fuel, which contains up to 2.6% 235 U. The same scheme is used to fabricate fuel for the experimental-industrial operation of fuel assemblies with regenerated uranium in VVÉR-440 and -1000 reactors [1]. The power plutonium, whose quantity has reached 30 tons [2, 3], separated during regeneration of uranium is accumulating. France has achieved substantial success in using regenerated plutonium in thermal reactors. Of the 11-12 tons of plutonium obtained by reprocessing spent PWR and BWR fuel, approximately 8.5 tons goes to fabricating mixed fuel, which is loaded into 30% of the core of 20 PWR [40].Other forms of fuel for burning regenerated plutonium are also being investigated [5]. One is plutonium dioxide, incorporated into an inert matrix, for example, zirconium oxide [6]. After irradiation, such fuel can be immediately sent into storage. A mixture of regenerated plutonium with enriched natural uranium with plutonium content ~2% has also been proposed. Such fuel can be loaded up to 100% into a PWR core [7,8].In France, regenerated uranium is partially loaded into PWR and BWR on an experimental-commercial scale. For additional enrichment, it is mixed with highly enriched weapons uranium [9] or enriched at an isotope-separation plant [2]. However, most of the regenerated uranium is stored.To simplify the reprocessing of spent fuel from the fast reactors which are being designed and to decrease the risk of plutonium proliferation, it has been proposed that uranium and plutonium be separated together [10]. The same approach is also possible for reprocessing fuel from thermal reactors. This will make it possible to return the regenerated uranium and plutonium into the fuel cycle. In the present article, one possible way of using regenerated uranium and plutonium in VVÉR-1000 reactors is examined.

Deep burial of liquid radioactive wastes in porous rocks is one of the methods of dealing with waste used in Russia [1]. Reliability in localizing wastes in such stores is determined primarily by the geological parameters, which should guarantee isolation from the surface and aquifers. The wastes represent a complicated multicomponent system, which may influence geochemical equilibria and alter the conditions in an underground store [2, 3]. Therefore, long-time forecasting for the state of such a store is impossible unless one knows the main transformations occurring in the waste-groundwater-rock system [4, 5].There is evidence on the main parameters governing the trends and extents of physicochemical processes in the thermal and radiation fields from the behavior of major components of the wastes and the radionuclides, including sorption on rocks, coprecipitation on solids, and so on [6][7][8][9]. The stratal temperature can be monitored periodically in injection and observation boreholes. These data characterize individual points but do not give a general picture of the temperature pattern and do not define zones of maximum heating or the temperatures there. To forecast component states at various times after deposition, one needs to know the distributions of the heat and dose levels throughout the store.Descriptions have been given [10] of ways of determining energy release and radiation doses in deep storage. Methods have been given [11] for calculating temperature patterns in storing liquids with the addition of cement involving hydraulic stratal fracturing. That form differs considerably from the storage of liquid wastes because the cement converts them to the solid state, which radically alters heat transfer. Thermal calculations on liquid waste storage [4, 7, 12] have shown that agreement is obtained with experiment when one considers the detailed technology, which includes not only depositing the wastes but also the injection of preparatory and displacing solutions. That is fairly obvious because the supply of large amounts of inactive solutions substantially reduces the radionuclide concentrations, as the radionuclides are the sources of heat and affect the heat-transfer conditions. Unfortunately, those papers give no details of the models, and the software used remains unknown, so one cannot perform calculations for other storage conditions.The data show that one can characterize the state of an underground waste store from a model that includes the following: 1) description of deposition in the storage rock; 2) calculation of energy production and radiation dosage; 3) calculation of temperature pattern at various times; 4) a physicochemical model for the state of the components that includes sedimentation, sorption, coprecipitation, and so on; and 5) calculations on component migration underground.

The results of a calculation of the neutron-physical characteristics of different variants of the fuel load of a VVER-1000 core are presented. The load completely consists of either enriched natural uranium (standard fuel) or recovered fuel, consisting of a uranium-plutonium mixture of recovered material, recovered plutonium, and highly enriched uranium. The mass fraction of plutonium isotopes in the recovered fuel ranged from 2 to 5%. All load variants examined have the same energy potential. The neutron-physical characteristics of a core with a full load of uranium are compared for a core with recovered uranium-plutonium fuel. Analysis of the neutron-physical characteristics did not show any fundamental limitations on the possibility of a VVER-1000 core load consisting entirely of recovered uranium-plutonium fuel.It is proposed in [1] that fuel based on recovered uranium-plutonium (an unseparated mixture of isotopes) which is separated from the spent fuel from thermal reactors and from which other actinides and fission products are removed and is mixed (taking account of the compensation and even isotopes of uranium and plutonium) with enriched natural uranium be used in thermal reactors. The specific consumption of natural uranium in the production of such fuel is approximately 20% lower than for fabrication of uranium fuel, and the specific separation work required to enrich the uranium part of the mixture remains almost unchanged [2]. The neutron-physical characteristics of a core with a 100% load of uranium and recovered uranium-plutonium fuel which affect the VVER-1000 operating safety differed negligibly, since the mass fraction of plutonium in the recovered fuel is approximately 1% [2].The required content of fissile nuclides in the fuel can be obtained by adding to the reprocessed uranium-plutonium and mixture, together with enriched natural uranium, the plutonium separated from spent VVER-440 fuel. Many years of European experience has validated the possibility of using in PWR plutonium mixed with uranium-plutonium fuel (mixture of recovered plutonium and depleted uranium) with mass fraction of plutonium fissile isotopes 4-6% [3].