熱中性子炉および高速炉におけるPu

Fabrication of Pu-bearing ceramic fuels for thermal and fast reactors for Indian nuclear power programme

The stepwise conversion of Russian nuclear power industry to a two-component nuclear energy system (NES) with thermal and fast reactors in a single closed NFC allows a number of deferred system problems of modern power industry to be solved, such as accumulation of spent fuel from thermal reactors and spent fuel repatriated from foreign Russian-designed NPPs; inefficient use of highly limited raw uranium inventories; recycling of minor actinides and other long-lived high-level waste accumulated in the course of thermal reactors' operation. The paper examines the fundamental capabilities of two-component NES, which appear due to the availability of commercial fast neutron sodium reactors. Due to the synergistic development of thermal and fast reactors, it is expected to achieve an economic effect associated with introduction of new fuel services in the nuclear energy system, such as production of plutonium and improvement of its isotopic composition for thermal reactors, transmutation of "external" minor actinides, production of isotopes for various purposes, production of hydrogen, etc. The expansion of Russia's export opportunities is also being considered, it can be done through fuel supply for both power units built in compliance with Russian technologies and foreign power units with MOX fuel. In this approach, it is important to analyze the economy of the whole system, where fast reactors, due to their inherent surplus of neutrons, perform the functions required to the entire NES, rather than to analyze the technical and economic indicators of individual reactors. The article makes a preliminary technical and economic assessment of such system functions provided by fast sodium reactors. It is shown that fast and thermal reactors operating together in a two-component nuclear energy system, with the same electrical power and similar technical and economic indicators, can significantly reduce the necessary infrastructure costs and/or generate additional income, thereby reducing the specific levelized cost of electricity production.

While the optimal use of the energy potential of uranium calls for development of the fast breeder reactor, it appears that their introduction could take more time than anticipated a few years ago. During this interim period, the recycle of the plutonium in thermal reactors is considered by several users as an interesting provisional alternative. In western Europe, 20 to 30 tons of irradiated mixed oxide (MOX) fuel exist today and between 100 and 300 tons could have been unloaded by about 1995, depending on the reprocessing capabilities for standard light water reactor (LWR) fuel and the implementation program for fast breeder reactors. The main differences between standard and MOX fuels result from the composition of the core material; however, these differences do not cause major technical difficulties that cannot be solved by established or known means. The reopening of the former Eurochemic plant at Mol, Belgium, for the reprocessing of nonstandard fuels, among which plutonium fuel unloaded from LWRs could occupy a leading position, would offer an opportunity to demonstrate the feasibility of this type of reprocessing and simultaneously cover the needs of the European countries adopting the recycle of plutonium in thermal reactors.

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One of the main reasons for the reprocessing of spent uranium oxide fuel is that of resource conservation. The fuel is termed 'spent' when it has been irradiated in a reactor for up to three years and is reaching the stage of diminishing efficiency in the production of heat through the fission of atoms. If retained in the reactor after this period heat is produced less efficiently and so the fuel is removed. 97 per cent of this irradiated fuel is uranium, up to 1 per cent is plutonium and 2–3 per cent is radioactive waste (Allday, 1977). As the fuel has been enriched — i.e. it has had uranium 235 added — it contains far more U235 after irradiation in a thermal reactor than is found in natural uranium. Reprocessing frees this uranium and plutonium from other actinides and fission products so that they may be reconverted into fuel for re-use in fast reactors and, conceivably, in thermal reactors. The effect of re-using the reprocessed fuel in a thermal reactor is to enable 30–40 per cent more power to be generated from the original material. However, such fuel can only used once. On the other hand recycling this material in a fast reactor would give an energy saving 50 times greater than that elicited from a thermal reactor (Allday, 1977). Hence the uranium and plutonium in spent oxide fuel is self-evidently a valuable source of energy if it can be recovered at a cost which permits the recycled U and Pu to have a value such that their use in thermal or fast reactors would provide power at a competitive price.1KeywordsFission ProductOxide FuelThermal ReactorSpend FuelUranium OxideThese keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

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The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. This approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

The Plutonium Utilization Program was initiated in 1956 with the objectives of developing technology for safe and economical recycle of plutonium fuel in power reactor systems, and demonstrating plutonium recycle in a practical manner. The program is conducted by the Pacific Northwest Laboratory, Battelle Memorial Institute under sponsorship of the AEC Division of Reactor Development and Technology. This report summarizes Plutonium Utilization Program results for fiscal year 1967. During the past year, it has become increasingly more evident that plutonium recycle in thermal reactors will be required in order to attain the most economic fuel cycle. This has come about because of the surge in orders from the utilities for water reactors and the uncertainty in the timing of fast reactor introduction. Based on present reactor startup projections, it appears that excess plutonium will be available for plutonium recycle by 1973. Although technical feasibility of plutonium recycle is now established, many questions remain to be answered to ensure optimum utilization of the plutonium. This will require an integrated effort of the AEC's and industry's programs. Specific steps are being taken by Pacific Northwest Laboratory to transfer to industry the technology developed in the AEC program and to cooperate with industry in developing programs of mutual interest. The present phase of the program is focused on obtaining the data required to assure timely application of plutonium to water reactors. The over-all program is comprised of research and development in fuels development, reactor physics, fuel cycle analysis, reactor engineering and materials with the major emphasis on the first two areas. Fuels development effort is emphasizing the irradiation behavior of PuO₂-UO₂ under high linear power ratings and exposure, transient and defect behavior, evaluation and comparison of PuO₂-UO₂ fuels fabricated using several methods, and investigation of problems and techniques involved in fuel fabrication. Experimental physics efforts have in the past emphasized measurements on uniform lattices of several fuel compositions. Emphasis has shifted to measurement of core nuclear characteristics as a function of exposure and to specific problems associated with loading plutonium in water reactors. The primary objective of the physics efforts is to evaluate and improve where necessary design and analysis methods to a stage where reliable engineering calculations can be made for predicting the nuclear characteristics of plutonium fueled thermal reactors. Highlights of the program are presented followed by detailed discussion of accomplishments in FY 1967.

Fast reactors in France are being developed by cooperation between CEA and EURATOM.This report describes the present status of the development of fast reactors in France. The following points are emphasized:In chapter I: Importance of the fast reactor program in the French electric power program.In chapter II: Organization and management of the French fast reactor project.In chapter III: Process of past development of the RAPSODIE, and the main results of its mock-up test.In chapter IV: Future fast reactor program (“Filière Réacteurs Rapides”); Studies for a prototype reactor (PHENIX) and a 1, 000MWe power reactor; experiments with the zero power reactor MASURCA and the fast neutron source reactor HARMONIE.

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The access to reliable sources of energy is the key to sustainable development of mankind. The major part of the energy consumed by people is generated with a chemical reaction of fossil fuel burning. This leads to quick depletion of natural resources and progressing environmental pollution. The contribution of the renewable energy sources to the general energy production remains insignificant. A modern 1,000 MW coal-fired thermal power plant (TPP) burns 2.5 million tons of coal per year and produces significant amount of solid and gaseous waste. TPPs are the largest consumers of atmospheric oxygen and sources of carbon dioxide. A nuclear power plant (NPP) of the same power consumes less than 50 tons of fuel per year. Environmentally significant NPP’s waste (liquid, solid and gaseous) is carefully collected, reduced in volume (evaporation, filtering, compaction, incineration, etc.) and securely isolated from the environment at the plant. The annual volume of waste for storage is less than 100 m3. The waste is under the control of a special NPP’s service and regulatory authorities. The energy of fission reaction millions of times exceeding the energy of burning has an enormous potential that mankind can receive. Four hundred and thirty-three nuclear power units with a total capacity of about 400 GW exist in the world. The accident at the Fukushima Daiichi NPP in Japan in March 2011 caused anxiety about nuclear safety throughout the world and raised questions about the future of nuclear power. Now, it is clear that the use of nuclear power will continue to grow in the coming decades, although the growth will be slower than was anticipated before the accident. Many countries with existing nuclear power programmes plan to expand them. Many new countries, both developed and developing, plan to introduce nuclear power. Some countries, such as Germany, plan to abandon nuclear energy. The IAEA’s latest projections show a steady rise in the number of NPPs in the world in the next 20 years. They project a growth in nuclear power capacity by 23% by 2030 in the low projection and by 100% in the high projection [1,2]. The basis of modern nuclear power comprises water-cooled nuclear reactors which use the energy potential of natural uranium inefficiently (thermal reactors). The thermal reactors use isotope U-235 in which the content of natural uranium is <1%. Breeder reactors are capable of using the significant part of energy potential, which is unavailable to thermal light water reactors. As a result, the same starting quantity of uranium can produce 50 times more energy. These reactors can transform U-238 into fissile Pu-239 in larger amounts than they consume fissile material. This feature is called ‘breeding’ [3]. The key problem of using the basic benefitsv of nuclear power is to ensure the safety of its use, as well as decommissioning and reliable isolation of process waste from the biosphere. The long-term large-scale nuclear power should possess guaranteed safety, economic stability and competitiveness, absence of the raw material base restrictions for a long period of time and environmental sustainability (low waste). The nuclear power systems with fast neutron reactors and liquid metal coolant can satisfy these conditions. More than 40 years of Russian experience in the field of construction and operation of sodium fast reactors makes it possible to summarize and analyze the ecological features of reactors of this type, the possibility of their use for sustainable energy supply of mankind and solving environmental problems.

Evolution of the CYCLE code for the system analysis of the nuclear fuel cycle

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density.

The Russian Government approved the Energy Strategy of the Russian Federation (Government Decree No.1523-r of June 9, 2020). The Strategy envisages the use of both thermal (TR) and fast (FR) reactors. The Strategy points out that the problems of nuclear power are associated with po-tential high expenses for irradiated fuel and radioactive wastes management. The previously de-signed model of the Russian nuclear energy development suggested that fast reactors only would operate at NPPs after 2010. Radiological equivalence, expressed as the equivalence of lifetime radiation risks to the public from radioactive wastes and from primary uranium ore, was shown to be achieved after 100-year storage. The burnup of 241Am, 237Np и 242Сm in closed nu-clear fuel cycle with fast reactors is a key part in the achievability of radiation risks equivalence. Scenarios of the Russian nuclear energy development through to 2100 with account of uncertain-ty factors in the measurement of contribution of fast and thermal reactors to the electric energy production are considered in the paper. The following three scenarios were developed: uncer-tainty is replaced by FRs; uncertainty is replaced by TRs; 50 per cent of FRs and 50 per cent of TRs replace uncertainty. If the energy is produced by fast reactors only (scenario 1) radiological equivalence was found to be achieved in 412 years. In two other scenarios radiological equiva-lence will be achieved after more than 1000 years. Contribution of main dose-forming radionu-clides and relevant ratios of potential biological hazards is included in models regardless of whether uncertainty in nuclear energy development is taking or not taking into account. Results of the study of conditions for radiological equivalence achievement should be used for amending Strategic plan of Russian nuclear power development through to 2100 that meets requirements of radiation ecology and radiation protection of the public.

This article chronicles the rise and fall of fast-reactor research in the United States. Research on fast reactors began at the end of World War II and represented a large fraction of the total U.S. research effort on civilian nuclear energy until the early 1980s. The goal of most of this research was to develop a plutonium breeder reactor capable of producing more plutonium from U-238 than is consumed. But with the termination of the Clinch River Breeder Reactor project in 1983, fast reactor development in the United States essentially ended. Safety issues played a role in this end to the fast breeder reactor program, but more important reasons were nuclear proliferation concerns and a growing conviction that breeder reactors would not be needed or economically competitive with light water reactors for decades, if ever.