RADIOCHEMICAL TECHNOLOGIES FOR FUEL CYCLE OF FAST REACTORS

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Опубликовано в библиотеке: 2021-11-10
Источник: Science in Russia, №6, 2013, C.4-9

by Andrei SHADRIN, Dr. Sc. (Chem.), Deputy Director for Science of the Center for Radioactive Wastes and Spent Nuclear Fuel Management, Bochvar High-Technology Research Institute of Inorganic Materials (Moscow)

 

Attainment of energy and economic leadership of our state is based on the creation of ecologically clean, safe and cheap nuclear power engineering. These parameters can be achieved by transition to a closed fuel cycle when nuclear "wastes" at the expense of extraction of uranium and plutonium from them, become a new fuel suitable for using in nuclear power stations. But modern industrial technologies for spent (irradiated) nuclear fuel processing and fabrication of a mixed uranium-plutonium fuel prevent from solving in full the pressing problems of a closed fuel cycle. Therefore, the present-day efforts of metallurgists, process engineers and designers should be directed to their profound updating.

 
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Schematic diagram of main operations of combined technologies for reprocessing of low-cooled irradiated nuclear fuel.

 

POWER ENGINEERING OF A NEW GENERATION

 

Specialists have long admitted that the technology based on thermal nuclear reactors with water- or graphite neutron decelators, which was widespread in our country and worldwide in the 1970s, cannot provide a full-scale and safe functioning of the industry. It is attributed to the low efficiency of a fissile material, namely uranium-235, which is equal only to ~0.7 percent in natural uranium (its remaining part is a "ballast") in such systems. Therefore, the long-term strategy of development of the large-scale nuclear power engineering involves transition to an advanced technology of a closed fuel cycle based on fast neutron reactors. They allow fuel breeding by conversion of uranium-238 going to wastes to plutonium-239 for its consequent return to a power cycle, and thus provide safety of nuclear power engineering and its practically infinite, in historical terms, resource self-sufficiency.

 

Fast neutron reactors have been long developed in many countries but are not widely used yet. Today in Russia there is in operation a primary sodium reactor BN-600 of 600 MW electric power output, the only such type in the world. It was commissioned in 1980 at the Beloyarsk nuclear power station near the city of Zarechny, Sverdlovsk Region. Another power unit is under construction there with 880 MW reactor output, whose starting is planned for 2014. But the knowledge stored in this field can be realized to the full in 1,200 MW fast neutron reactors with a sodium (BN-1200) and lead (BREST-1200) coolants developed at Afrikantov Experimental Design Bureau of Machine-Building (Nizhni Novgorod) and Dollezhal Research and Development Institute of Power Equipment (Moscow). It is planned to put into operation a prototype of the commercial fast neutron pilot demonstration reactor BREST-300-OD with a nuclear fuel cycle plant (plant designed for reprocessing of spent nuclear fuel and production of a new nuclear fuel from regenerated materials) in the 2020s in the territory of the Siberian Integrated Chemical Plant in the city of Seversk of the Tomsk Region.

 

But the mass transition to a closed fuel cycle based on fast neutron reactors is impossible without solving a number of chemical-engineering problems. They, in particular, include decrease in duration of the so-called external fuel cycle (phase of fissile material processing outside the reactor) and increase of its burnup; reduction of expenses for fuel processing and fabrication at the expense of cutback in the production method and also in volumes of secondary nontechnological wastes and increase in equipment service life; reduction of expenses for radioactive wastes at the expense of transmutation (transformation) of long-living actinides, recycling (return) of structural materials and utilization of reagents decomposing during reprocessing. Besides,

 
стр. 5

 

 

Schematic diagram of a combined (pyro + hydro) technology for reprocessing of spent nuclear fuel of fast neutron reactors.

 

transition to a closed cycle raises questions connected with a high content of fissile materials in an irradiated and refabricated (newly produced) fuel.

 

HIGH BURNUP AND LOW COOLING TIME

 

As of today the average burnup of processed uranium or mixed oxide spent nuclear fuel of thermal reactors makes ~55 and 45 hW-day/t respectively, while for uranium-plutonium fuel of fast neutron reactors this parameter is expected to be 100 hW-day/t and more. Near to double increase of this parameter will cause almost proportional increase of fission products in irradiated fuel, and many of them present a serious problem in terms of precipitation formation in processing by conventional hydrometallurgical methods based on the PUREX (from the English Plutonium-Uranium Recovery by Extraction) process.

 

Besides, growth of fission products will inevitably provoke occurrence of molybdenum and zirconium sediments in case of a low acid intensity of solutions or strontium and barium sediments in case of high acid intensity, but increased concentration of plutonium will provoke its capture by sediments and necessity of special operations for their washing off. All that will adversely tell on the economic efficiency of spent nuclear fuel processing, which is already a rather expensive process.

 

Low cooling time of the irradiated fuel after its unloading from the reactor will allow to decrease storehouse volumes and the amount of the stored nuclear materials, but will require operations with spent nuclear fuel, possessing high heat generation and temperature. Surely, all mentioned risks can be reduced at the expense of dilution of solutions, but it will essentially increase the volume of the processed high- and average-activity wastes. Such decision will hardly be cost-effective even in case of using the oxidation technology for tritium predistillation or crystallization for giving up organic extractive agent (solvent). In other words, the attractiveness of conventional hydrometallurgical processes decreases with the growth of irradiated fuel burnup.

 

Let us point out here that the "dry" (waterless, pyroelectrochemical and gas-fluoride) technologies of spent nuclear fuel reprocessing have been developing for long, including in relation to dense (metallic) fuel of fast neutron reactors. Pyroelectrochemical processes running in molten salts allow extraction of uranium,

 
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plutonium and neptunium in the form of metal. Advantages of a gas-fluoride or dry technology are evident in the reprocessing of spent nuclear fuel of thermal reactors, which results in the production of uranium in the form of hexafluoride suitable for its enrichment, but are not evident at all in case of fast neutron reactors when uranium and plutonium are not separated from each other in accordance with the principles of non-proliferation of nuclear arms.

 

Reduction of costs for reprocessing of highly burnt down fuel with a low cooling time can be promoted also by different mixed technologies based on "dry" methods and also in combination with "water" processes. Each of them implies operations which separate actinides (group of radioactive chemical elements with atomic numbers 90-103) from fission products without use of water solutions, which provides an opportunity to work with a fuel of a low cooling time. In a word, a low cooling time and a high burnup compel to use non-aqueous processes.

 

HIGH CONTENT OF FISSION MATERIAL

 

Another but similar peculiarity of reprocessing of spent nuclear fuel of fast neutron reactors is a high content of intrinsic fission materials. At present only France and Russia have production experience in handling such fuel. Probably the problems connected with nuclear safety force again to refer to "dry" methods. It should be noted that the hydrometallurgical methods are also suitable for operation with large quantities of fission materials but they demand dilution of solutions or use of circular devices, which affects adversely the reprocessing economics due to the increased volumes of wastes and devices.

 

REDUCTION IN THE NUMBER OF OPERATIONS AND WASTE VOLUME

 

The attempts to achieve cost cutting at the expense of reduction in the number of operations are made mainly by designers of hydrometallurgical technologies. Probably it is attributed to the experience accumulated by them in the commercial operation of radiochemical production plants. Besides, it is just within the limits of these technologies and with a view to prevent formation of large quantities of water and organic wastes that they suggest transition to precipitation processes and reorientation to inorganic vehicles for sorption (absorption) and also application of direct thermal

 
стр. 7

 

denitration of actinide nitrates (decomposition of acid salts) for production of uranium and plutonium powders. For hydrometallurgical mixed technologies they study also corrosive activity of media for selection of more durable structural materials and also possible application of decomposable reagents within the framework of the CHON concept well-known in Europe. Accepted in France it implies using as a solvent and extractive agent completely burning substances, which consist only of carbon (C), hydrogen (H), oxygen (O) and nitrogen (N), hence the name CHON.

 

The transfer of the concept of cost reduction (including at the expense of combining dissimilar processes, increase of the service life of the equipment, introduction of more durable vehicles for sorption and adsorption, i.e. introduction of the principle "save everywhere and in everything") to development of "dry" technologies is a major task of researchers, production engineers and designers.

 

HANDLING OF SMALL ACTINIDES

 

It is a key problem of not only reprocessing of spent nuclear fuel of fast neutron reactors but also of the whole closed fuel cycle, which provokes no doubts among specialists. Afterburning in fast neutron reactors is the most promising handling method of small actinides. But the problem of whether this method shall be homogeneous (small actinides are introduced directly to regenerated nuclear fuel) or heterogeneous (small actinides are burnt up in special targets) still has no single-valued solution as well as the problem of the final fate of transuranic element curium. Meanwhile, their solution affects not only the structure of a fuel cycle, but also selection of reprocessing technologies of spent nuclear fuel and also requirements to the procedures of solidification and burial of radioactive wastes.

 

Today large-scale studies of fractionation for hydrometallurgical and pyroelectrochemical technologies is under way in Europe. Unfortunately, far from all processes being developed have been tested on real products, and the tests were not always successful. Still, EXAM and SETFICS processes connected with extraction of transuranic elements are considered the most promising. But they cannot be regarded as ready for industrial application. Therefore, development of fractionation techniques and their introduction to industry remain a major task of the present-day radiochemistry.

 

COMBINED TECHNOLOGY OF SPENT NUCLEAR FUEL REPROCESSING

 

Let us emphasize that the "dry" technologies (pyroelectrochemical and gas-fluoride) being developed nowadays are suitable in principle for reprocessing of "fast" neutron reactor fuel with the lag less than a year, but still none of them guarantees formation of a final

 
стр. 8

 

uranium-plutonium product suitable for production of nitride or carbide pellets and return of 99.9 percent of actinides to a fuel cycle. For the time being this is possible only in case of application of a combined technology (pyro + hydro) based on a combination of pyroelectrochemical fractionation of uranium-plutonium-neptunium and its hydrometallurgical refining, i. e. decontamination. It is applicable to oxide, nitride, carbonitride and metallic irradiated fuel of "fast" neutron reactors and implies use of pyroelectrochemical techniques capable of processing spent nuclear fuel with a small cooling time (up to half a year) and separating the main portion (up to 99 percent) of highly active fission products from uranium, plutonium and neptunium for fuel refabrication. Hydrometallurgical processes are designed also for refining of recyclable components, isolation and separation of radioactive elements of americium and curium and additional extraction of actinides from wastes.

 

Combination of two methods (pyro + hydro) provides a positive synergetic (joint) effect and allows reprocessing of spent nuclear fuel with the high burnup and low cooling time, which results eventually in the reduction of the volume of stored irradiated fuel and the quantity of plutonium in a closed cycle. It should be stressed that the combined technology allows recycling of any type of spent nuclear fuel of "fast" reactors, formation of a uranium-plutonium-neptunium product of a high degree of purification and application of the pellet technology for fuel refabrication.

 

Evidently for cutting of the external fuel cycle time one ought to apply non-aqueous technologies, which allow to work with a highly depleted and low-cooled fuel of "fast" neutron reactors. But without reduction of expenses on handling of spent nuclear fuel and radioactive wastes, most likely, it will not be possible to achieve an efficient fuel cycle.

 

It should be admitted that specialists have not found yet a global solution of all mentioned problems. Moreover, there are also different variations as regards particular matters. But one thing is evident: for industrial introduction of a closed cycle with fast neutron reactors it is necessary to master at least the technologies of removing of shells of spent nuclear fuel, handle volatile fission products (14C, Kr, Xe, etc.), extract transplutonium elements and separate americium and curium, provide mould solidification of radioactive wastes suitable for final isolation, handle structural materials and regenerate the medium.

 

Illustrations supplied by the author


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© Andrei SHADRIN () Источник: Science in Russia, №6, 2013, C.4-9

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